Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, 46557-46569 [E8-18429]

Download as PDF 46557 Proposed Rules Federal Register Vol. 73, No. 155 Monday, August 11, 2008 This section of the FEDERAL REGISTER contains notices to the public of the proposed issuance of rules and regulations. The purpose of these notices is to give interested persons an opportunity to participate in the rule making prior to the adoption of the final rules. NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 RIN 3150–AI01 [NRC–2007–0008] Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events Nuclear Regulatory Commission. ACTION: Supplemental Proposed Rule. rwilkins on PROD1PC63 with PROPOSALS AGENCY: SUMMARY: The Nuclear Regulatory Commission (NRC) is considering the adoption of provisions regarding applicability of the rule and new provisions regarding procedures to perform surveillance data checks related to the updated fracture toughness requirements for protection against pressurized thermal shock (PTS) events for pressurized water reactor (PWR) pressure vessels. The NRC is considering these provisions as an alternative to the provisions previously noticed for public comment on October 3, 2007 (72 FR 56275). DATES: Submit comments on this proposed rule by September 10, 2008. Submit comments on the information collection aspects on this proposed rule by September 10, 2008. ADDRESSES: You may submit comments by any one of the following methods. Please include the following number RIN 3150–AI01 in the subject line of your comments. Comments submitted in writing or in electronic form will be made available for public inspection. Because your comments will not be edited to remove any identifying or contact information, the NRC cautions you against including any information in your submission that you do not want to be publicly disclosed. Federal e Rulemaking Portal: Go to http://www.regulations.gov and search for documents filed under Docket ID NRC–2007–0008. Address questions about NRC dockets to Carol Gallagher VerDate Aug<31>2005 17:37 Aug 08, 2008 Jkt 214001 (301) 415–5905; e-mail Carol.Gallager@nrc.gov. Mail comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555–0001, ATTN: Rulemakings and Adjudications Staff. E-mail comments to: Rulemaking.Comments@nrc.gov. If you do not receive a reply e-mail confirming that we have received your comments, contact us directly at (301) 415–1966. Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland 20852, between 7:30 am and 4:15 pm during Federal workdays. (Telephone (301) 415–1966). Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at (301) 415–1101. You can access publicly available documents related to this document using the following methods: NRC’s Public Document Room (PDR): The public may examine publicly available documents at the NRC’s PDR, Public File Area O–F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland. The PDR reproduction contractor will copy documents for a fee. NRC’s Agencywide Document Access and Management System (ADAMS): Publicly available documents created or received at the NRC are available electronically at the NRC’s Electronic Reading Room at http://www.nrc.gov/ reading-rm/adams.html. From this page, the public can gain entry into ADAMS, which provides text and image files of NRC’s public documents. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC’s PDR reference staff at 1–800–397–4209, or (301) 415–4737, or by e-mail to PDR.Resource@nrc.gov. Ms. Veronica M. Rodriguez, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555–0001; telephone (301) 415–3703; e-mail: Veronica.Rodriguez@nrc.gov, Mr. Barry Elliot, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555– 0001; telephone (301) 415–2709; e-mail: Barry.Elliot@nrc.gov, or Mr. Mark Kirk, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555–0001; telephone FOR FURTHER INFORMATION CONTACT: PO 00000 Frm 00001 Fmt 4702 Sfmt 4702 (301) 415–6015; e-mail: Mark.Kirk@nrc.gov. SUPPLEMENTARY INFORMATION: I. Introduction II. Background III. Discussion IV. Responses to Comments on the Proposed Rule V. Section-by-Section Analysis VI. Specific Request for Comments VII. Availability of Documents VIII. Plain Language IX. Voluntary Consensus Standards X. Finding of No Significant Environmental Impact: Availability XI. Paperwork Reduction Act Statement XII. Regulatory Analysis XIII. Regulatory Flexibility Act Certification XIV. Backfit Analysis I. Introduction The NRC published a proposed rule on alternate fracture toughness requirements for protection against Pressurized Thermal Shock (PTS) for public comments in the Federal Register on October 3, 2007 (72 FR 56275). This rule provides new PTS requirements based on updated analysis methods. This action is desirable 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, while maintaining adequate safety. These new requirements would be utilized by any Pressurized Water Reactor (PWR) licensee as an alternative to complying with the existing requirements. During the development of the PTS final rule, the NRC determined that several changes to the proposed rule language may be needed to adequately address issues raised in stakeholder’s comments. The NRC also determined, in response to a stakeholder comment, that the characteristics of advanced PWR designs were not considered in the technical analysis made for the proposed rule. The NRC does not have assurance that reactors that commence commercial power operation after the effective date of this rule will have operating characteristics and materials of fabrication similar to those evaluated as part of the technical basis for the proposed rule. Therefore, the NRC has concluded that it would be prudent to E:\FR\FM\11AUP1.SGM 11AUP1 46558 Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules rwilkins on PROD1PC63 with PROPOSALS limit the applicability and the use of § 50.61a to currently-operating plants only, and proposes to modify the applicability provisions of the proposed rule accordingly. Also, several stakeholders questioned the accuracy and validity of the generic embrittlement curves in the proposed rule. The NRC wants to ensure that the predicted values from the proposed embrittlement trend curves provide an adequate basis for implementation of the rule. Therefore, the NRC has continued to work on statistical procedures to identify deviations from generic embrittlement trends, such as those described in § 50.61a(f)(6) of the proposed rule. Based on this work, the NRC is considering enhancing the procedure described in paragraph § 50.61a(f)(6) to, among other things, detect signs from the plant- and heatspecific surveillance data of embrittlement trends that are not reflected by Equations 5, 6 and 7 of the rule that may emerge at high fluences. Because these proposed modifications may not represent a logical outgrowth from the October 2007 proposed rule’s provisions, the NRC concludes that obtaining stakeholder feedback on the proposed alternative provisions through the use of a supplemental proposed rule is appropriate. As discussed in Section VI of this notice, the NRC will consider comments on §§ 50.61a(b); (f)(6)(i) through (f)(6)(vi); Equations 10, 11, and 12 in § 50.61a(g); and Tables 5, 6, and 7 of this supplemental proposed rule. The NRC is also requesting comments on whether there should be additional language added to § 50.61a(e) to allow licensees to account for the effects of sizing errors. This supplemental proposed rule does not reflect other modifications or editorial and conforming changes that the NRC is considering to incorporate in the final rule as a result of the public comments on the October 2007 proposed rule. II. Background PTS events are system transients in a PWR in which severe overcooling occurs coincident with high pressure. The thermal stresses are caused by rapid cooling of the reactor vessel inside surface, which combine with the stresses caused by high pressure. The aggregate effect of these stresses is an increase in the potential for fracture if a pre-existing 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, can be susceptible to brittle fracture. VerDate Aug<31>2005 17:37 Aug 08, 2008 Jkt 214001 The PTS rule, described in § 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 § 50.61 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 § 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, plant modifications to reduce PTS event probability or severity, and reactor vessel annealing, which are addressed in §§ 50.61(b)(3), (b)(4), and (b)(7); and § 50.66, ‘‘Requirements for Thermal Annealing of the Reactor Pressure Vessel.’’ Currently, no operating PWR reactor vessel is projected to exceed the § 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) developed a technical basis that supports updating the PTS regulations. This technical basis concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. This finding indicated that the screening criteria in § 50.61 are unnecessarily conservative and may impose an unnecessary burden on some licensees. Therefore, the NRC created a new rule, § 50.61a, which provides alternate screening criteria and corresponding embrittlement correlations based on the updated technical basis. 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, in a Staff Requirements Memorandum (SRM) dated June 30, PO 00000 Frm 00002 Fmt 4702 Sfmt 4702 2006, to prepare a rulemaking which would allow current PWR licensees to implement the new requirements of § 50.61a or continue to comply with the current requirements of § 50.61. Alternatively, the NRC could have revised § 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 NRC published the proposed rulemaking on the alternate fracture toughness requirements for protection against PTS for public comment in the Federal Register on October 3, 2007 (72 FR 56275). The proposed rule provided an alternative to the current rule, which a licensee may choose to adopt. This prompted the NRC to keep the current requirements separate from the new alternative requirements. As a result, the proposed rule retained the current requirements in § 50.61 for PWR licensees choosing not to implement the less restrictive screening limits, and presented new requirements in § 50.61a as an alternative relaxation for PWR licensees. III. Discussion The NRC published a proposed new rule, § 50.61a (October 3, 2007, 72 FR 56275), that would provide new PTS requirements based on updated analysis methods because the existing requirements are based on unnecessarily conservative probabilistic fracture mechanics analyses. Stakeholders’ comments raised concerns related to the applicability of the rule and the accuracy and validity of the generic embrittlement curves. The NRC reconsidered the technical and regulatory issues in these areas and is considering adopting the modified provisions regarding the applicability of the rule and new provisions regarding procedures to perform surveillance data checks described in this supplemental proposed rule. The NRC will consider comments on §§ 50.61a(b), (f)(6)(i) through (f)(6)(vi); Equations 10, 11 and 12 in § 50.61a(g); and Tables 5, 6, and 7 of this supplemental proposed rule. As described in Section VI of this notice, the NRC is also requesting comments on whether there should be additional language added to § 50.61a(e) to allow licensees to account for the effects of sizing errors. The NRC will consider the October 2007 proposed rule, the supplemental proposed rule, and the comments received in response E:\FR\FM\11AUP1.SGM 11AUP1 Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules to both, when deciding whether to adopt a final PTS rule. rwilkins on PROD1PC63 with PROPOSALS Applicability of the Proposed Rule, § 50.61a(b) The supplemental proposed rule differs from the proposed rule and from § 50.61 in that it proposes to limit the use of § 50.61a to currently operating plants only. It cannot be demonstrated, a priori, that reactors which commence commercial power operation after the effective date of this rule will have operating characteristics, in particular identified PTS event sequences and thermal-hydraulic responses, which are consistent with the reactors which were evaluated as part of the technical basis for this rule. Other factors, including materials of fabrication and welding methods, could also vary. Hence, the use of § 50.61a would be limited to currently operating PWR facilities which are known to have characteristics consistent with those assumed in the technical basis. The NRC also proposes to allow the holder of the operating license for Watts Bar Unit 2 to adopt the requirements in § 50.61a as this facility has operating characteristics consistent with those assumed in the technical basis. The NRC recognizes that licensees for reactors who commence commercial power operation after the effective date of this rule may, under the provisions of § 50.12, seek an exemption from § 50.61a(b) to apply this rule if a plantspecific basis analyzing their operating characteristics, materials of fabrications, and welding methods is provided. Surveillance Data, § 50.61a(f) Section 50.61a(f) of the proposed rule defines the process for calculating the values for the material properties (i.e. , RTMAX–X) for a particular reactor vessel. These values would be based on the vessel material’s copper, manganese, phosphorus, and nickel weight percentages, reactor cold leg temperature, and fast neutron flux and fluence values, as well as the unirradiated nil-ductility transition reference temperature (i.e., RTNDT). Section 50.61a(f) of the proposed rule included a procedure by which the RTMAX–X values, which are predicted for plant-specific materials using a generic temperature shift (i.e., DT30) embrittlement trend curve, are compared with heat-specific surveillance data that are collected as part of 10 CFR Part 50, Appendix H surveillance programs. The purpose of this comparison is to assess how well the surveillance data are represented by the generic embrittlement trend curve. If the surveillance data are close (closeness is assessed statistically) to the VerDate Aug<31>2005 17:37 Aug 08, 2008 Jkt 214001 generic embrittlement trend curve, then the predictions of this embrittlement trend curve are used. This is expected to normally be the case. However, if the heat-specific surveillance data deviate significantly, and non-conservatively, from the predictions of the generic embrittlement trend curve, this indicates that alternative methods (i.e., other than, or in addition to, the generic embrittlement trend curve) may be needed to reliably predict the temperature shift trends, and to estimate RTMAX–X, for the conditions being assessed. However, alternative methods for temperature shift prediction are not prescribed by § 50.61a(f) of the proposed rule. Although standard and accepted procedures exist to assess the statistical significance of the differences between heat-specific surveillance data and the generic embrittlement trend curve, similarly standard and acceptable procedures are not available to assess the practical importance of such differences. The practical importance of statistically significant deviations is best assessed by licensees on a case-by-case basis, which would be submitted for the review of the Director of NRR, as prescribed by § 50.61a(f). The method described in the proposed rulemaking to compare the heat-specific surveillance data collected as part of 10 CFR part 50, Appendix H surveillance programs to the generic temperature shift embrittlement trend curve included a single statistical test. This statistical test was set forth by Equations 9 and 10, and Table 5. This test determined if, on average, the temperature shift from the surveillance data was significantly higher than the temperature shift of the generic embrittlement trend curve. The NRC has determined that, while necessary, this single test is not sufficient to ensure that the temperature shift predicted by the embrittlement trend curve well represents the heat-specific surveillance data. Specifically, this single statistical test cannot determine if the temperature shift from the surveillance data shows a more rapid increase after significant radiation exposure than the progression predicted by the generic embrittlement trend curve. To address this potential deficiency, which could be particularly important during a plant’s period of extended operation, the NRC added two more statistical tests in this supplemental proposed rulemaking, which are expressed by Equations 11 and 12 and by Tables 6 and 7. Together, these two additional tests determine if the surveillance data from a particular heat show a more rapid increase after significant radiation exposure than the PO 00000 Frm 00003 Fmt 4702 Sfmt 4702 46559 progression predicted by the generic embrittlement trend curve. The NRC documented the technical basis for the proposed alternative in the following reports: (1) ‘‘Statistical Procedures for Assessing Surveillance Data for 10 CFR Part 50.61a,’’ (ADAMS Accession No. ML081290654), and (2) ‘‘A Physically Based Correlation of Irradiation Induced Transition Temperature Shifts for RPV Steel,’’ (ADAMS Accession No. ML081000630). IV. Responses to Comments on the Proposed Rule The NRC received 5 comment letters on the proposed 10 CFR 50.61a rule published on October 3, 2007 (72 FR 56275). The following paragraphs discuss those comments which are directly associated with the supplemental proposed rule’s provisions on the applicability of the rule and surveillance data procedures. The remainder of the comments and the NRC responses will be provided in the Federal Register notice for the final rule. Comments on the Applicability of the Proposed Rule Comment: The commenters stated that the rule, as written, is only applicable to the existing fleet of PWRs. The characteristics of advanced PWR designs were not considered in the analysis. The commenters suggested adding a statement to state that this rule is applicable to the current PWR fleet and not the new plant designs. [PWROG–5, EPRI–5] Response: The NRC agrees with the comment that this rule is only applicable to the existing fleet of PWRs. The NRC cannot be assured that reactors that commence commercial power operation after the effective date of this rule will have operating characteristics, in particular identified PTS event sequences and thermal-hydraulic responses, which are consistent with the reactors that were evaluated as part of the technical basis for § 50.61a. Other factors, including materials of fabrication and welding methods, could also vary. Therefore, the NRC agrees with the commenters that it would be prudent to restrict the use of § 50.61a to current plants. As a result of this comment, the NRC proposes to modify § 50.61a(b) and the statement of considerations of the rule to reflect this position to limit the use of the rule to currently operating plants. Comments on Surveillance Data Comment: The commenters stated that there is little added value in the requirement to assess the surveillance E:\FR\FM\11AUP1.SGM 11AUP1 rwilkins on PROD1PC63 with PROPOSALS 46560 Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules data as a part of this rule because variability in data has already been accounted for in the derivation of the embrittlement correlation. The commenters also stated that there is no viable methodology for adjusting the projected DT30 for the vessel based on the surveillance data. Any effort to make this adjustment is likely to introduce additional error into the prediction. Note that the embrittlement correlation described in the basis for the revised PTS rule (i.e., NUREG–1874) was derived using all of the currently available industry-wide surveillance data. In the event that the surveillance data does not match the DT30 value predicted by the embrittlement correlation, the best estimate value for the pressure vessel material is derived using the embrittlement correlation. The likely source of the discrepancy is an error in the characterization of the surveillance material or of the irradiation environment. Therefore, unless the discrepancy can be resolved, obtaining the DT30 prediction based on the best estimate chemical composition for the heat of the material is more reliable than a prediction based on a single set of surveillance measurements. The commenters suggested removing the requirement to assess surveillance data, including Table 5, of this rule. [PWROG–4, EPRI–4, NEI–2] Response: The NRC does not agree with the proposed change. The NRC believes that there is added value in the requirement to assess surveillance data. Although variability has been accounted for in the derivation of the embrittlement correlation, it is the NRC’s view that the surveillance assessment required in § 50.61a(f)(6) is needed to determine if the embrittlement for a specific heat of material in a reactor vessel is consistent with the embrittlement predicted by the embrittlement correlation. The commenters also assert that there is no viable methodology for adjusting the projected DT30 for the vessel based on the surveillance data, and that any adjustment is likely to introduce additional error into the prediction. The NRC believes that although there is no single methodology for adjusting the projected DT30 for the vessel based on the surveillance data, it is possible, on a case-specific basis, to justify adjustments to the generic DT30 prediction. For this reason the rule does not specify a method for adjusting the DT30 value based on surveillance data, but rather requires the licensee to propose a case-specific DT30 adjustment procedure for review and approval from the Director. Although the commenters VerDate Aug<31>2005 17:37 Aug 08, 2008 Jkt 214001 assert that it is possible that error could be introduced, it is the NRC view that appropriate plant-specific adjustments based upon available surveillance data may be necessary to project reactor pressure vessel embrittlement for the purpose of this rule. As the result of these public comments, the NRC has continued to work on statistical procedures to identify deviations from generic embrittlement trends, such as those described in § 50.61a(f)(6) of the proposed rule. Based on this work, the NRC is considering further enhancing the procedure described in paragraph (f)(6) to, among other things, detect signs from the plant- and heat-specific surveillance data that may emerge at high fluences of embrittlement trends that are not reflected by Equations 5, 6, and 7. The empirical basis for the NRC’s concern regarding the potential for unmodeled high fluence effects is described in documents located at ADAMS Accession Nos. ML081120253, ML081120289, ML081120365, ML081120380, and ML081120600. The technical basis for the enhanced surveillance assessment procedure is described in the document located at ADAMS Accession No. ML081290654. V. Section-by-Section Analysis The following section-by-section analysis only discusses the modifications in the provisions related to the applicability of the rule and surveillance data procedures that the NRC is considering as an alternative in this supplemental proposed rule. The NRC is only seeking comments on these alternative provisions. This supplemental proposed rule does not reflect other modifications or editorial and conforming changes that the NRC is considering to incorporate as a result of the public comments on the proposed rule that were not discussed in this notice as they will be provided in the Federal Register notice for the final rule. would evaluate the surveillance for consistency with the embrittlement model by following the procedures specified by §§ 50.61a(f)(6)(ii), (f)(6)(iii), and (f)(6)(iv) of the supplemental proposed rule. Proposed § 50.61a(f)(6)(ii) The proposed language for § 50.61a(f)(6)(ii) would establish the requirements to perform an estimate of the mean deviation of the data set from the embrittlement model. The mean deviation for the data set would be compared to values given in Table 5 or Equation 10 of this section. The NRC proposes to modify this paragraph to state that the surveillance data analysis would follow the criteria in §§ 50.61a(f)(6)(v) and (f)(6)(vi) of the supplemental proposed rule. Proposed § 50.61a(f)(6)(iii) The NRC proposes to modify § 50.61a(f)(6)(iii) to establish the requirements to estimate the slope of the embrittlement model residuals (i.e., the difference between the measured and predicted value for a specific data point). The licensee would estimate the slope using Equation 11 and compare this value to the maximum permissible value in Table 6, both from the supplemental proposed rule. This surveillance data analysis would follow the criteria in §§ 50.61a(f)(6)(v) and (f)(6)(vi) of the supplemental proposed rule. Proposed § 50.61a(f)(6)(iv) The NRC proposes to modify § 50.61a(f)(6)(iv) to establish the requirements to estimate an outlier deviation from the embrittlement model for the specific data set using Equations 8 and 12. The licensee would compare the normalized residuals to the allowable values in Table 7 of the supplemental proposed rule. This surveillance data analysis would follow the criteria in §§ 50.61a(f)(6)(v) and (f)(6)(vi) of the supplemental proposed rule. Proposed § 50.61a(b) The proposed language for § 50.61a(b) would establish the applicability of the rule. The NRC proposes to modify this paragraph to limit the use of this rule to currently-operating plants only. Proposed § 50.61a(f)(6)(v) The NRC proposes to add paragraph (f)(6)(v) to establish the criteria to be satisfied in order to calculate the DT30 shift values. Proposed § 50.61a(f)(6)(i) The proposed language for § 50.61a(f)(6)(i) would establish the requirements to perform data checks to determine if the surveillance data show a significantly different trend than what the embrittlement model in this rule predicts. The NRC proposes to modify § 50.61a(f)(6)(i)(B) to state that licensees Proposed § 50.61a(f)(6)(vi) The NRC proposes to add paragraph (f)(6)(vi) to establish the actions to be taken by a licensee if the criteria in paragraph (f)(6)(v) of this section are not met. The licensee would need to submit an evaluation of the surveillance data and propose values for DT30, considering their plant-specific PO 00000 Frm 00004 Fmt 4702 Sfmt 4702 E:\FR\FM\11AUP1.SGM 11AUP1 46561 Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules surveillance data, for the review and approval by the Director. The licensee would need to submit an evaluation of each surveillance capsule removed from the vessel after the submittal of the initial application for review and approval by the Director no later than 2 years after the capsule is withdrawn from the vessel. Proposed § 50.61a(g) The proposed language for § 50.61a(g) would provide the necessary equations and variables required by the proposed changes in § 50.61a(f)(6). The NRC proposes to modify Equation 10 to account for 1 percent of significance level. Equations 11 and 12 would be added to provide the means for estimating the slope and the outlier deviation from the embrittlement model. Proposed Tables 5, 6, and 7 Tables 5, 6, and 7 would provide values to be used in the proposed changes in § 50.61a(f)(6). The NRC proposes to modify Table 5 to account for the use of a 1 percent of significance level. Tables 6 and 7 would be added to provide the threshold values for the slope and the outlier deviation tests. VI. Specific Request for Comments The NRC seeks comments on §§ 50.61a(b), (f)(6)(i) through (f)(6)(vi); Equations 10, 11, and 12 in § 50.61a(g), and Tables 5, 6, and 7 of the supplemental proposed rule. The NRC is not seeking comments on any other provisions of the proposed § 50.61a which remain unchanged from the October 2007 proposed rule. In addition, the NRC also requests comments on the following question: Adjustments of the Inservice Inspection Volumetric Examination and Flaw Assessments The flaw sizes in Tables 2 and 3 are selected so that reactor vessels with flaw sizes less than or equal to those in the tables will have a TWCF less than or equal to 1 × 10¥6 per reactor year at the maximum permissible embrittlement. The NRC recognizes that the flaw sizes in these tables represent actual flaw dimensions while the results from the ASME Code examinations are estimated dimensions. The available information indicates that, for most flaw sizes in Tables 2 and 3, qualified inspectors will oversize flaws. Comparing oversized flaws to the size and density distributions in Tables 2 and 3 is conservative and acceptable, but not necessary. Therefore, NRC is considering to permit flaw sizes to be adjusted to account for the effects of sizing error before comparing the estimated size and density distribution to the acceptable size and density distributions in Tables 2 and 3. This would be accomplished by requiring licensees to base the methodology to account for the effects of sizing error on statistical data collected from ASME Code inspector qualification tests. An acceptable method would include a demonstration, that accounting for the effects of sizing error, is unlikely to rwilkins on PROD1PC63 with PROPOSALS Document 17:37 Aug 08, 2008 Jkt 214001 PO 00000 Frm 00005 VII. Availability of Documents 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. Regulations.gov (Web). These documents may be viewed and downloaded electronically through the Federal eRulemaking Portal http:// www.regulations.gov, Docket number NRC–2007–0008. NRC’s Electronic Reading Room (ERR). The NRC’s public electronic reading room is located at http:// www.nrc.gov/reading-rm.html. PDR Federal Register Notice—Proposed Rule: Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (RIN 3150–AI01), 72 FR 56275, October 3, 2007 ................................................................................................................................................. Letter from Thomas P. Harrall, Jr., dated December 17, 2007, ‘‘Comments on Proposed Rule 10 CFR 50, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, RIN 3150–AI01’’ [Identified as Duke] ............................................................... Letter from Jack Spanner, dated December 17, 2007, ‘‘10 CFR 50.55a Proposed Rulemaking Comments RIN 3150–AI01’’ [Identified as EPRI] ............................................................................ Letter from James H. Riley, dated December 17, 2007, ‘‘Proposed Rulemaking—Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (RIN 3150–AI01), 72 FR 56275, October 3, 2007 [Identified as NEI] ..................................................... Letter from Melvin L. Arey, dated December 17, 2007, ‘‘Transmittal of PWROG Comments on the NRC Proposed Rule on Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events’’, RIN 3150–AI01, PA-MSC–0232 [Identified as PWROG] .... Letter from T. Moser, dated December 17, 2007, ‘‘Strategic Teaming and Resource Sharing (STARS) Comments on RIN 3150–AI01, Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events 72 FR 56275 (October 3,2007) [Identified as STARS] ............................................................................................................................................ ‘‘Statistical Procedures for Assessing Surveillance Data for 10 CFR Part 50.61a’’ ........................... ‘‘A Physically Based Correlation of Irradiation Induced Transition Temperature Shifts for RPV Steel’’ ................................................................................................................................................ Supplemental Regulatory Analysis ...................................................................................................... Supplemental OMB Supporting Statement ......................................................................................... Memo from J. Uhle, dated May 15, 2008, ‘‘Embrittlement Trend Curve Development for Reactor Pressure Vessel Materials’’ .............................................................................................................. Draft ‘‘Technical Basis for Revision of Regulatory Guide 1.99: NRC Guidance on Methods to Estimate the Effects of Radiation Embrittlement on the Charpy V-Notch Impact Toughness of Reactor Vessel Materials’’ ........................................................................................................................ VerDate Aug<31>2005 result in accepting actual flaw size distribution that cause the TWCF to exceed the acceptance criteria. Adjusting flaw sizes to account for sizing error can change an unacceptable examination result into an acceptable result; further, collecting, evaluating, and using data from ASME Code inspector qualification tests will require extensive engineering judgment. Therefore, the methodology would have to be reviewed and approved by the Director of the NRC’s Office of Nuclear Reactor Regulation (NRR) to ensure that the risk associated with PTS is acceptable. The NRC requests specific comments on whether there should be additional language added to 10 CFR 50.61a(e) to allow licensees to account for the effects of sizing errors. Fmt 4702 Sfmt 4702 Web ERR (ADAMS) X NRC–2007–0008 ML072750659 X NRC–2007–0008 ML073521542 X NRC–2007–0008 ML073521545 X NRC–2007–0008 ML073521543 X NRC–2007–0008 ML073521547 X X NRC–2007–0008 ML073610558 ML081290654 X X X NRC–2007–0008 NRC–2007–0008 ML081000630 ML081440673 ML081440736 X ML081120253 X ML081120289 E:\FR\FM\11AUP1.SGM 11AUP1 46562 Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules Document PDR ‘‘Comparison of the Predictions of RM–9 to the IVAR and RADAMO Databases’’ ........................... Memo from M. Erickson Kirk, dated December 12, 2007, ‘‘New Data from Boiling Water Reactor Vessel Integrity Program (BWRVIP) Integrated Surveillance Project (ISP)’’ .................................. ‘‘Further Evaluation of High Fluence Data’’ ......................................................................................... X ML081120365 X X ML081120380 ML081120600 VIII. Plain Language The Presidential memorandum ‘‘Plain Language in Government Writing’’ published in 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 heading of this notice. rwilkins on PROD1PC63 with PROPOSALS IX. Voluntary Consensus Standards The National Technology Transfer and Advancement Act of 1995, Public Law 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 determined that there is only one technical standard developed that could be utilized for characterizing the embrittlement correlations. That standard is the 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 supplemental proposed rule. However, the correlation developed by the NRC 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 supplemental proposed rule because it does not represent the broad range of conditions necessary to justify a revision to the regulations. The ASME Code requirements are utilized as part of the volumetric examination analysis requirements of the supplemental 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 Part 50, Appendix H and utilized to determine 30-foot-pound transition temperatures. These standards were selected for use in the supplemental proposed rule based on their use in other regulations within 10 CFR Part 50 and their applicability VerDate Aug<31>2005 17:37 Aug 08, 2008 Jkt 214001 to the subject of the desired requirements. The NRC will consider using a voluntary consensus standard in the final rule if an appropriate standard is identified in the public comment period for this supplemental proposed rule. X. Finding of No Significant Environmental Impact: Availability The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission’s regulations in 10 CFR Part 51, Subpart A, 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. This determination was made as part of the proposed rulemaking issued on October 3, 2007 (72 FR 56275), and remains applicable to this supplemental proposed rulemaking. XI. Paperwork Reduction Act Statement This supplemental 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 supplemental 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 50.61 and 50.61a)’’ supplemental 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 an alternative to the requirements of 10 CFR 50.61. Collections would also be required, after implementation of the new 10 CFR 50.61a, when any change is made to the design or operation of the PO 00000 Frm 00006 Fmt 4702 Sfmt 4702 Web ERR (ADAMS) facility that affects the calculated RTMAX-X value. Collections would also be required during the scheduled periodic ultrasonic examination of beltline welds. Who will be required or asked to report: Licensees of currently operating PWRs 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: 363 hours (253 hours annually for record keeping plus 110 hours annually for reporting). Abstract: The NRC is proposing to amend its regulations to provide updated fracture toughness requirements for protection against PTS events for PWR pressure vessels. The supplemental 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 is expected to 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 utilized by licensees of currently operating PWRs as an alternative to complying with the existing requirements. The NRC is seeking public comment on the potential impact of the information collections contained in this supplemental 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? E:\FR\FM\11AUP1.SGM 11AUP1 Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules rwilkins on PROD1PC63 with PROPOSALS 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–1F21, 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. The document will be available on the NRC home page site 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 September 10, 2008. 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. Comments submitted in writing or in electronic form will be made available for public inspection. Because your comments will not be edited to remove any identifying or contact information, the NRC cautions you against including any information in your submission that you do not want to be publicly disclosed. Comments submitted should reference Docket No. NRC–2007–0008. Comments can be submitted in electronic form via the Federal eRulemaking Portal at http:// www.regulations.gov by search for Docket No. NRC–2007–0008. Comments can be mailed to NRC Clearance Officer, Russell Nichols (T–5F52), U.S. Nuclear Regulatory Commission, Washington, DC 20555–0001. Questions about the information collection requirements may be directed to the NRC Clearance Officer, Russell Nichols (T–5 F52), U.S. Nuclear Regulatory Commission, Washington, DC 20555–0001, by telephone at (301) 415–6874, or by email to INFOCOLLECTS.Resource@ nrc.gov. Comments can be mailed to the Desk Officer, Office of Information and Regulatory Affairs, NEOB–10202, (3150–0011), Office of Management and Budget, Washington, DC 20503, or by email to Nathan_J._Frey@omb.eop.gov, or by telephone at (202) 395–7345. Public Protection Notification 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. XII. Regulatory Analysis The NRC has issued a supplemental regulatory analysis for this supplemental proposed rulemaking. The analysis examines the costs and benefits VerDate Aug<31>2005 17:37 Aug 08, 2008 Jkt 214001 of the alternatives considered by the NRC. The NRC requests public comments on this supplemental draft regulatory analysis. Availability of the supplemental regulatory analysis is provided in Section VII of this notice. Comments on the supplemental draft regulatory analysis may be submitted to the NRC as indicated under the ADDRESSES heading of this notice. XIII. Regulatory Flexibility Act Certification In accordance with the Regulatory Flexibility Act (5 U.S.C. 605(b)), the NRC certifies that this rule would not, if promulgated, have a significant economic impact on a substantial number of small entities. This supplemental proposed rule would affect only the licensing and operation of currently operating 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). XIV. Backfit Analysis The NRC has determined that the requirements in this supplemental proposed rule would 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 supplemental proposed rule. The requirements of the proposed PTS rule, including those in the supplemental proposed rule, would not be required, but could be utilized by PWR licensees with currently operating plants. Licensees choosing to implement the proposed PTS rule would be required to comply with its requirements as an 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, the NRC finds that this amendment would not constitute backfitting. List of Subjects for 10 CFR Part 50 Antitrust, Classified information, Criminal penalties, Fire protection, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements. For the reasons set out in the preamble and under the authority of the PO 00000 Frm 00007 Fmt 4702 Sfmt 4702 46563 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. PART 50—DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES 1. The authority citation for part 50 continues to read as follows: Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88 Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note); sec. 651(e), Pub. L. 109–58, 119 Stat. 806–810 (42 U.S.C. 2014, 2021, 2021b, 2111). Section 50.7 also issued under Pub. L. 95– 601, sec. 10, 92 Stat. 2951 as amended by Pub. L. 102–486, sec. 2902, 106 Stat. 3123 (42 U.S.C. 5841). Section 50.10 also issued under secs. 101, 185, 68 Stat. 955, as amended (42 U.S.C. 2131, 2235); sec. 102, Pub. L. 91–190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd), and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 50.55a and appendix Q also issued under sec. 102, Pub. L. 91–190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58, 50.91, and 50.92 also issued under Pub. L. 97–415, 96 Stat. 2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42 U.S.C. 2152). Sections 50.80–50.81 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued under sec. 187, 68 Stat. 955 (42 U.S.C. 2237). 2. Section 50.8(b) is revised to read as follows: § 50.8 Information collection requirements: OMB approval. * * * * * (b) The approved information collection requirements contained in this part appear in §§ 50.30, 50.33, 50.34, 50.34a, 50.35, 50.36, 50.36a, 50.36b, 50.44, 50.46, 50.47, 50.48, 50.49, 50.54, 50.55, 50.55a, 50.59, 50.60, 50.61, 50.61a, 50.62, 50.63, 50.64, 50.65, 50.66, 50.68, 50.69, 50.70, 50.71, 50.72, 50.74, 50.75, 50.80, 50.82, 50.90, 50.91, 50.120, and appendices A, B, E, G, H, I, J, K, M, N,O, Q, R, and S to this part. * * * * * 3. Section 50.61a is added to read as follows: § 50.61a Alternate fracture toughness requirements for protection against pressurized thermal shock events. (a) Definitions. Terms in this section have the same meaning as those set E:\FR\FM\11AUP1.SGM 11AUP1 rwilkins on PROD1PC63 with PROPOSALS 46564 Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules forth in 10 CFR 50.61(a), with the exception of the term ‘‘ASME Code’’. (1) ASME Code means the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division I, ‘‘Rules for the Construction of Nuclear Power Plant Components,’’ and Section XI, Division I, ‘‘Rules for Inservice Inspection of Nuclear Power Plant Components,’’ edition and addenda and any limitations and modifications thereof as specified in § 50.55a. (2) RTMAX–AW means the material property which characterizes the reactor vessel’s resistance to fracture initiating from flaws found along axial weld fusion lines. RTMAX–AW is determined under the provisions of paragraph (f) of this section and has units of °F. (3) RTMAX–PL means the material property which characterizes the reactor vessel’s resistance to fracture initiating from flaws found in plates in regions that are not associated with welds found in plates. RTMAX–PL is determined under the provisions of paragraph (f) of this section and has units of °F. (4) RTMAX–FO means the material property which characterizes the reactor vessel’s resistance to fracture initiating from flaws in forgings that are not associated with welds found in forgings. RTMAX–FO is determined under the provisions of paragraph (f) of this section and has units of °F. (5) RTMAX–CW means the material property which characterizes the reactor vessel’s resistance to fracture initiating from flaws found along the circumferential weld fusion lines. RTMAX–CW is determined under the provisions of paragraph (f) of this section and has units of °F. (6) RTMAX–X means any or all of the material properties RTMAX–AW, RTMAX–PL, RTMAX–FO, or RTMAX–CW for a particular reactor vessel. (7) jt means fast neutron fluence for neutrons with energies greater than 1.0 MeV. jt is determined under the provisions of paragraph (g) of this section and has units of n/cm2. (8) j means average neutron flux. j is determined under the provisions of paragraph (g) of this section and has units of n/cm2/sec. (9) ∆T30 means the shift in the Charpy V-notch transition temperature produced by irradiation defined at the 30 ft-lb energy level. The DT30 value is determined under the provisions of paragraph (g) of this section and has units of °F. (10) Surveillance data means any data that demonstrates the embrittlement trends for the beltline materials, including, but not limited to, data from test reactors or surveillance programs at VerDate Aug<31>2005 17:37 Aug 08, 2008 Jkt 214001 other plants with or without a surveillance program integrated under 10 CFR Part 50, Appendix H. (11) Tc means cold leg temperature under normal full power operating conditions, as a time-weighted average from the start of full power operation through the end of licensed operation. Tc has units of °F. (b) Applicability. Each licensee of a pressurized water nuclear power reactor, whose original operating license was issued prior to [EFFECTIVE DATE OF FINAL RULE], and the holder of any operating license issued under this part or part 54 for the Watts Bar Unit 2 facility, may utilize the requirements of this section as an alternative to the requirements of 10 CFR 50.61. (c) Request for Approval. Prior to implementation of this section, each licensee shall submit a request for approval in the form of a license amendment together with the documentation required by paragraphs (c)(1), (c)(2), and (c)(3) of this section for review and approval to the Director, Office of Nuclear Reactor Regulation (Director). The information required by paragraphs (c)(1), (c)(2), and (c)(3) of this section must be submitted for review and approval by the Director at least three years before the limiting RTPTS value calculated under 10 CFR 50.61 is projected to exceed the PTS screening criteria in 10 CFR 50.61 for plants licensed under this part. (1) Each licensee shall have projected values of RTMAX–X for each reactor vessel beltline material for the EOL fluence of the material. The assessment of RTMAX–X values must use the calculation procedures given in paragraphs (f) and (g) of this section, except as provided in paragraphs (f)(6) and (f)(7) of this section. The assessment must specify the bases for the projected value of RTMAX–X for each reactor vessel beltline material, including the assumptions regarding future plant operation (e.g., core loading patterns, projected capacity factors, etc.); the copper (Cu), phosphorus (P), manganese (Mn), and nickel (Ni) contents; the reactor cold leg temperature (TC); and the neutron flux and fluence values used in the calculation for each beltline material. (2) Each licensee shall perform an examination and an assessment of flaws in the reactor vessel beltline as required by paragraph (e) of this section. The licensee shall verify that the requirements of paragraphs (e)(1) through (e)(3) have been met and submit all documented indications and the neutron fluence map required by paragraph (e)(1)(iii) to the Director in its application to utilize 10 CFR 50.61a. If PO 00000 Frm 00008 Fmt 4702 Sfmt 4702 analyses performed under paragraph (e)(4) of this section are used to justify continued operation of the facility, approval by the Director is required prior to implementation. (3) Each licensee shall compare the projected RTMAX–X values for plates, forgings, axial welds, and circumferential welds to the PTS screening criteria for the purpose of evaluating a reactor vessel’s susceptibility to fracture due to a PTS event. If any of the projected RTMAX–X values are greater than the PTS screening criteria in Table 1 of this section, then the licensee may propose the compensatory actions or plantspecific analyses as required in paragraphs (d)(3) through (d)(7) of this section, as applicable, to justify operation beyond the PTS screening criteria in Table 1 of this section. (d) Subsequent Requirements. Licensees who have been approved to utilize 10 CFR 50.61a under the requirements of paragraph (c) of this section shall comply with the requirements of this paragraph. (1) Whenever there is a significant change in projected values of RTMAX–X, such that the previous value, the current value, or both values, exceed the screening criteria prior to the expiration of the plant operating license; or upon the licensee’s request for a change in the expiration date for operation of the facility; a reassessment of RTMAX–X values documented consistent with the requirements of paragraph (c)(1) and (c)(3) of this section must be submitted for review and approval to the Director. If the Director does not approve the assessment of RTMAX–X values, then the licensee shall perform the actions required in paragraphs (d)(3) through (d)(7) of this section, as necessary, prior to operation beyond the PTS screening criteria in Table 1 of this section. (2) Licensees shall determine the impact of the subsequent flaw assessments required by paragraphs (e)(1)(i), (e)(1)(ii), (e)(2), and (e)(3) of this section and shall submit the assessment for review and approval to the Director within 120 days after completing a volumetric examination of reactor vessel beltline materials as required by Section XI of the ASME Code. If a licensee is required to implement paragraphs (e)(4) and (e)(5) of this section, a reanalysis in accordance with paragraphs (e)(4) and (e)(5) of this section is required within one year of the subsequent ASME Code inspection. (3) If the value of RTMAX–X is projected to exceed the PTS screening criteria, then the licensee shall implement those flux reduction E:\FR\FM\11AUP1.SGM 11AUP1 rwilkins on PROD1PC63 with PROPOSALS Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules programs that are reasonably practicable to avoid exceeding the PTS screening criteria. The schedule for implementation of flux reduction measures may take into account the schedule for review and anticipated approval by the Director of detailed plant-specific analyses which demonstrate acceptable risk with RTMAX–X values above the PTS screening criteria due to plant modifications, new information, or new analysis techniques. (4) If the analysis required by paragraph (d)(3) of this section indicates that no reasonably practicable flux reduction program will prevent the RTMAX–X value for one or more reactor vessel beltline materials from exceeding the PTS screening criteria, then the licensee shall perform a safety analysis to determine what, if any, modifications to equipment, systems, and operation are necessary to prevent the potential for an unacceptably high probability of failure of the reactor vessel as a result of postulated PTS events if continued operation beyond the PTS screening criteria is to be allowed. In the analysis, the licensee may determine the properties of the reactor vessel materials based on available information, research results and plant surveillance data, and may use probabilistic fracture mechanics techniques. This analysis must be submitted to the Director at least three years before RTMAX–X is projected to exceed the PTS screening criteria. (5) After consideration of the licensee’s analyses, including effects of proposed corrective actions, if any, submitted under paragraphs (d)(3) and (d)(4) of this section, the Director may, on a case-by-case basis, approve operation of the facility with RTMAX–X values in excess of the PTS screening criteria. The Director will consider factors significantly affecting the potential for failure of the reactor vessel in reaching a decision. (6) If the Director concludes, under paragraph (d)(5) of this section, that operation of the facility with RTMAX–X values in excess of the PTS screening criteria cannot be approved on the basis of the licensee’s analyses submitted under paragraphs (d)(3) and (d)(4) of this section, then the licensee shall request a license amendment, and receive approval by the Director, prior to any operation beyond the PTS screening criteria. The request must be based on modifications to equipment, systems, and operation of the facility in addition to those previously proposed in the submitted analyses that would reduce the potential for failure of the reactor vessel due to PTS events, or on VerDate Aug<31>2005 17:37 Aug 08, 2008 Jkt 214001 46565 further analyses based on new information or improved methodology. (7) If the limiting RTMAX–X value of the facility is projected to exceed the PTS screening criteria and the requirements of paragraphs (d)(3) through (d)(6) of this section cannot be satisfied, the reactor vessel beltline may be given a thermal annealing treatment under the requirements of § 50.66 to recover the fracture toughness of the material. The reactor vessel may be used only for that service period within which the predicted fracture toughness of the reactor vessel beltline materials satisfy the requirements of paragraphs (d)(1) through (d)(6) of this section, with RTMAX–X values accounting for the effects of annealing and subsequent irradiation. (e) Examination and Flaw Assessment Requirements. The volumetric examinations results evaluated under paragraphs (e)(1), (e)(2), and (e)(3) of this section must be acquired using procedures, equipment and personnel that have been qualified under the ASME Code, Section XI, Appendix VIII, Supplement 4 and Supplement 6. (1) The licensee shall verify that the indication density and size distributions within the ASME Code, Section XI, Appendix VIII, Supplement 4 inspection volume 1 are within the flaw density and size distributions in Tables 2 and 3 of this section based on the test results from the volumetric examination. The allowable number of flaws specified in Tables 2 and 3 of this section represent a cumulative flaw size distribution for each ASME flaw size increment. The allowable number of flaws for a particular ASME flaw size increment represents the maximum total number of flaws in that and all larger ASME flaw size increments. The licensee shall also demonstrate that no flaw exceeds the size limitations specified in Tables 2 and 3 of this section. (i) The licensee shall determine the allowable number of weld flaws for the reactor vessel beltline by multiplying the values in Table 2 of this section by the total length of the reactor vessel beltline welds that were volumetrically inspected and dividing by 1000 inches of weld length. (ii) The licensee shall determine the allowable number of plate or forging flaws for their reactor vessel beltline by multiplying the values in Table 3 of this section by the total plate or forging surface area that was volumetrically inspected in the beltline plates or forgings and dividing by 1000 square inches. (iii) For each indication detected in the ASME Code, Section XI, Appendix VIII, Supplement 4 inspection volume, the licensee shall document the dimensions of the indication, including depth and length, the orientation of the indication relative to the axial direction, and the location within the reactor vessel, including its azimuthal and axial positions and its depth embedded from the clad-to-base metal interface. The licensee shall also document a neutron fluence map, projected to the date of license expiration, for the reactor vessel beltline clad-to-base metal interface and indexed in a manner that allows the determination of the neutron fluence at the location of the detected indications. (2) The licensee shall identify, as part of the examination required by paragraph (c)(2) of this section and any subsequent ASME Code, Section XI ultrasonic examination of the beltline welds, any indications within the ASME Code, Section XI, Appendix VIII, Supplement 4 inspection volume that are located at the clad-to-base metal interface. The licensee shall verify that such indications do not open to the vessel inside surface using a qualified surface or visual examination. (3) The licensee shall verify, as part of the examination required by paragraph (c)(2) of this section and any subsequent ASME Code, Section XI ultrasonic examination of the beltline welds, all indications between the clad-to-base metal interface and three-eighths of the reactor vessel thickness from the interior surface are within the allowable values in ASME Code, Section XI, Table IWB–3510–1. (4) The licensee shall perform analyses to demonstrate that the reactor vessel will have a through-wall crack frequency (TWCF) of less than 1 × 10¥6 per reactor year if the ASME Code, Section XI volumetric examination required by paragraph (c)(2) or (d)(2) of this section indicates any of the following: (i) The indication density and size in the ASME Code, Section XI, Appendix VIII, Supplement 4 inspection volume is not within the flaw density and size limitations specified in Tables 2 and 3 of this section; (ii) Any indication in the ASME Code, Section XI, Appendix VIII, Supplement 4 inspection volume that is larger 2 than 1 The ASME Code, Section XI, Appendix VIII, Supplement 4 weld volume is the weld volume from the clad-to-base metal interface to the inner 1.0 inch or 10 percent of the vessel thickness, whichever is greater. 2 Table 2 for the weld flaws is limited to flaw sizes that are expected to occur and were modeled from the technical basis supporting this rule. Similarly, Table 3 for the plate and forging flaws PO 00000 Frm 00009 Fmt 4702 Sfmt 4702 E:\FR\FM\11AUP1.SGM Continued 11AUP1 46566 Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules rwilkins on PROD1PC63 with PROPOSALS the sizes in Tables 2 and 3 of this section; (iii) There are linear indications that penetrate through the clad into the low alloy steel reactor vessel shell; or (iv) Any indications between the cladto-base metal interface and threeeighths 3 of the vessel thickness exceed the size allowable in ASME Code, Section XI, Table IWB–3510–1. (5) The analyses required by paragraph (e)(4) of this section must address the effects on TWCF of the known sizes and locations of all indications detected by the ASME Code, Section XI, Appendix VIII, Supplement 4 and Supplement 6 ultrasonic examination out to three-eighths of the vessel thickness from the inner surface, and may also take into account other reactor vessel-specific information, including fracture toughness information. (f) Calculation of RTMAX–X values. Each licensee shall calculate RTMAX–X values for each reactor vessel beltline material using jt. jt must be calculated using an NRC-approved methodology. (1) The values of RTMAX–AW, RTMAX–PL, RTMAX–FO, and RTMAX–CW must be determined using Equations 1 through 4 of this section. (2) The values of DT30 must be determined using Equations 5 through 7 of this section, unless the conditions specified in paragraph (f)(6)(vi) of this section are met, for each axial weld fusion line, plate, and circumferential weld fusion line. The DT30 value for each axial weld fusion line calculated as specified by Equation 1 of this section must be calculated for the maximum fluence (jtFL) occurring along a particular axial weld fusion line. The DT30 value for each plate calculated as specified by Equation 1 of this section must be calculated for jtFL occurring along a particular axial weld fusion line. The DT30 value for each plate or forging calculated as specified by Equations 2 and 3 of this section are calculated for the maximum fluence (jtMAX) occurring at the clad-to-base metal interface of each plate or forging. In Equation 4, the jtFL value used for calculating the plate, forging, and circumferential weld RTMAX–CW value is the maximum jt occurring for each material along the circumferential weld fusion line. (3) The values of Cu, Mn, P, and Ni in Equations 6 and 7 of this section must represent the best estimate values for the material weight percentages. For a plate or forging, the best estimate value is normally the mean of the measured values for that plate or forging. For a weld, the best estimate value is normally the mean of the measured values for a weld deposit made using the same weld wire heat number as the critical vessel weld. If these values are not available, either the upper limiting values given in the material specification to which the vessel material was fabricated, or conservative estimates (mean plus one standard deviation) based on generic data 4 as shown in Table 4 of this section for P and Mn, must be used. (4) The values of RTNDT(U) must be evaluated according to the procedures in the ASME Code, Section III, paragraph NB–2331. If any other method is used for this evaluation, the licensee shall submit the proposed method for review and approval by the Director along with the calculation of RTMAX–X values required in paragraph (c)(1) of this section. (i) If a measured value of RTNDT(U) is not available, a generic mean value of RTNDT(U) for the class 5 of material must be used if there are sufficient test results to establish a mean. (ii) The following generic mean values of RTNDT(U) must be used unless justification for different values is provided: 0 °F for welds made with Linde 80 weld flux; and ¥56 °F for welds made with Linde 0091, 1092, and 124 and ARCOS B–5 weld fluxes. (5) The value of Tc in Equation 6 of this section must represent the weighted time average of the reactor cold leg temperature under normal operating full power conditions from the beginning of full power operation through the end of licensed operation. (6) The licensee shall verify that an appropriate RTMAX–X value has been calculated for each reactor vessel beltline material. The licensee shall consider plant-specific information that could affect the use of Equations 5 though 7 of this section for the determination of a material’s DT30 value. (i) The licensee shall evaluate the results from a plant-specific or integrated surveillance program if the surveillance data satisfy the criteria described in paragraphs (f)(6)(i)(A) and (f)(6)(i)(B) of this section: stops at the maximum flaw size modeled for these materials in the technical basis supporting this rule. 3 Because flaws greater than three-eighths of the vessel wall thickness from the inside surface do not contribute to TWCF, flaws greater than threeeighths of the vessel wall thickness from the inside surface need not be analyzed for their contribution to PTS. 4 Data from reactor vessels fabricated to the same material specification in the same shop as the vessel in question and in the same time period is an example of ‘‘generic data.’’ 5 The class of material for estimating RT NDT(U) must be determined by the type of welding flux (Linde 80, or other) for welds or by the material specification for base metal. VerDate Aug<31>2005 17:37 Aug 08, 2008 Jkt 214001 PO 00000 Frm 00010 Fmt 4702 Sfmt 4702 (A) The surveillance material must be a heat-specific match for one or more of the materials for which RTMAX–X is being calculated. The 30-foot-pound transition temperature must be determined as specified by the requirements of 10 CFR Part 50, Appendix H. (B) If three or more surveillance data points measured at three or more different neutron fluences exist for a specific material, the licensee shall determine if the surveillance data show a significantly different trend than the embrittlement model predicts. This must be achieved by evaluating the surveillance data for consistency with the embrittlement model by following the procedures specified by paragraphs (f)(6)(ii), (f)(6)(iii), and (f)(6)(iv) of this section. If fewer than three surveillance data points exist for a specific material, then the embrittlement model must be used without performing the consistency check. (ii) The licensee shall estimate the mean deviation from the embrittlement model for the specific data set (i.e. , a group of surveillance data points representative of a given material). The mean deviation from the embrittlement model for a given data set must be calculated using Equations 8 and 9 of this section. The mean deviation for the data set must be compared to the maximum heat-average residual given in Table 5 or derived using Equation 10 of this section. The maximum heat-average residual is based on the material group into which the surveillance material falls and the number of surveillance data points. The surveillance data analysis must use the criteria in paragraphs (f)(6)(v) and (f)(6)(vi) of this section. For surveillance data sets with greater than 8 shift points, the maximum credible heat-average residual must be calculated using Equation 10 of this section. The value of s used in Equation 10 of this section must be obtained from Table 5 of this section. (iii) The licensee shall estimate the slope of the embrittlement model residuals (estimated using Equation 8) plotted as a function of the base 10 logarithm of neutron fluence for the specific data set. The licensee shall estimate the T-statistic for this slope (TSURV) using Equation 11 and compare this value to the maximum permissible T-statistic (TMAX) in Table 6. The surveillance data analysis must follow the criteria in paragraphs (f)(6)(v) and (f)(6)(vi) of this section. For surveillance data sets with greater than 15 shift points, the TMAX value must be calculated using Student’s T distribution with a significance level (a) of 1 percent for a one-tailed test. E:\FR\FM\11AUP1.SGM 11AUP1 46567 Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules permissible T-statistic (TMAX) in Table 6; and (C) The largest normalized residual value is equal to or less than the appropriate allowable value from the third column in Table 7 and the second largest normalized residual value is equal to or less than the appropriate allowable value from the second column in Table 7. (vi) If any of the criteria described in paragraph (f)(6)(v) of this section are not satisfied, the licensee shall review the data base for that heat in detail, including all parameters used in Equations 4, 5, and 6 of this section and the data used to determine the baseline Charpy V-notch curve for the material in an unirradiated condition. The licensee shall submit an evaluation of the surveillance data and shall, on the basis of this review, propose DT30 and RTMAX–X values, considering their plant-specific surveillance data, to be used for evaluation relative to the (iv) The licensee shall estimate the two largest positive deviations (i.e. , outliers) from the embrittlement model for the specific data set using Equations 8 and 12. The licensee shall compare the largest normalized residual (r*) to the appropriate allowable value from the third column in Table 7 and the second largest normalized residual to the appropriate allowable value from the second column in Table 7. The surveillance data analysis must follow the criteria in paragraphs (f)(6)(v) and (f)(6)(vi) of this section. (v) The DT30 value must be determined using Equations 5, 6, and 7 of this section if all three of the following criteria are satisfied: (A) The mean deviation from the embrittlement model for the data set is equal to or less than the value in Table 5 or the value derived using Equation 10 of this section; (B) The T-statistic for the slope (TSURV) estimated using Equation 11 is equal to or less than the maximum acceptance criteria of this rule. These evaluations shall be submitted for the review and approval by the Director at the time of the initial application. For each surveillance capsule removed from the reactor vessel after the submittal of the initial application, the licensee shall perform the analyses required by paragraph (f)(6) of this section. The analyses must be submitted for the review and approval by the Director in the form of a license amendment, and must be submitted no later than two years after the capsule is withdrawn from the vessel. (7) The licensee shall report any information that significantly improves the accuracy of the RTMAX–X value to the Director. Any value of RTMAX–X that has been modified as specified in paragraph (f)(6)(iv) of this section is subject to the approval of the Director when used as provided in this section. (g) Equations and variables used in this section. { } Equation 1: RTMAX-AW = MAX  RTNDT(u) - plate + ∆T30 - plate ( ϕt FL )  ,  RTNDT ( u ) - axial weld + ∆T30 - axial weld ( ϕt FL )      Equation 2: RTMAX-PL = RTNDT ( u ) - plate + ∆T30 - plate ( ϕt MAX ) rwilkins on PROD1PC63 with PROPOSALS Where: P [wt-%] = phosphorus content Mn [wt-%] = manganese content Ni [wt-%] = nickel content Cu [wt-%] = copper content A = 1.140 × 10¥7 for forgings = 1.561 × 10¥7 for plates = 1.417 × 10¥7 for welds B = 102.3 for forgings = 102.5 for plates in non-Combustion Engineering manufactured vessels VerDate Aug<31>2005 17:37 Aug 08, 2008 Jkt 214001 = 135.2 for plates in Combustion Engineering vessels = 155.0 for welds jte = j for j ≥ 4.39 × 1010 n/cm2/sec = jt × (4.39 × 1010 / j)0.2595 for j < 4.39 × 1010 n/cm2/sec Where: j[n/cm2/sec] = average neutron flux t[sec] = time that the reactor has been in full power operation jt[n/cm2] = j × t f(Cue,P) = 0 for Cu ≤ 0.072 PO 00000 Frm 00011 Fmt 4702 Sfmt 4702 = [Cue¥0.072]0.668 for Cu > 0.072 and P ≤ 0.008 = [Cue¥0.072 + 1.359 × (P¥0.008)]0.668 for Cu > 0.072 and P > 0.008 and Cue = 0 for Cu ≤ 0.072 = MIN (Cu, maximum Cue) for Cu > 0.072 and maximum Cue = 0.243 for Linde 80 welds = 0.301 for all other materials g(Cue,Ni,jte) = 0.5 + (0.5 × tanh {[log10(jte) + (1.1390 × Cue)¥(0.448 × Ni)¥18.120] / 0.629}) E:\FR\FM\11AUP1.SGM 11AUP1 EP11AU08.021</MATH> EP11AU08.020</MATH> Equation 7: CRP = B × (1 + 3.77 × Ni1.191 ) × f ( Cu e , P ) × g ( Cu e , Ni, ϕt e ) EP11AU08.019</MATH> Equation 6: MD = A × (1 − 0.001718 × TC ) × (1 + 6.13 × P × Mn 2.471 ) × ϕt 0.5 e EP11AU08.018</MATH> Equation 5: ∆T30 = MD + CRP EP11AU08.017</MATH> } EP11AU08.016</MATH> { Equation 4: RTMAX-CW = MAX  RTNDT ( u ) - plate + ∆T30 - plate ( ϕt MAX )  ,  RTNDT ( u ) - circweld + ∆T30 - circweld ( ϕt MAX )  ,  RTNDT ( u ) - forging + ∆T30 - forging ( ϕt MAX )        EP11AU08.022</MATH> Equation 3: RTMAX-FO = RTNDT ( u ) - forging + ∆T30 - forging ( ϕt MAX ) 46568 Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules Equation 8: Residual (r) = measured ∆T30 − predicted ∆T30 (by Equations 5, 6 and 7) n Equation 9: Mean deviation for a data set of n data points = (1/n ) × ∑ ri i =1 Equation 10: Maximum credible heat-average residual = 2.33σ/n 0.5 σ Where: n = number of surveillance shift data points (sample size) in the specific data set s = standard deviation of the residuals about the model for a relevant material group given in Table 5. Equation 11: TSURV = Where: m = the slope of a plot of all of the r values (estimated using Equation 8) versus the base 10 logarithm of the neutron fluence for each r value. The slope shall be estimated using the method of least squares. se(m) = the least squares estimate of the standard-error associated with the estimated slope value m. m se(m) Equation 12: r* = r σ Where: r is defined using Equation 8 and s is given in Table 5. TABLE 1—PTS SCREENING CRITERIA RTMAX-X limits [°F] for different vessel wall thicknesses 6 (TWALL) Product form and RTMAX-X values TWALL ≤ 9.5in. Axial Weld, RTMAX-AW ............................................................................................................................. Plate, RTMAX-PL ....................................................................................................................................... Forging without underclad cracks, RTMAX-FO .......................................................................................... Axial Weld and Plate, RTMAX-AW + RTMAX-PL ........................................................................................ Circumferential Weld, RTMAX-CW7 ........................................................................................................... Forging with underclad cracks, RTMAX-FO ............................................................................................... 6 Wall 9.5in. < TWALL ≤ 10.5in. 269 356 356 538 312 246 230 305 305 476 277 241 10.5in. < TWALL ≤ 11.5in. 222 293 293 445 269 239 thickness is the beltline wall thickness including the clad thickness. limits contributes 1 × 10¥8 per reactor year to the ractor vessel TWCF. 7 RT PTS TABLE 2—ALLOWABLE NUMBER OF FLAWS IN WELDS TWE TWE TWE TWE TWE TWE TWE TWE TWE < < < < < < < < < 0.075 0.125 0.175 0.225 0.275 0.325 0.375 0.425 0.475 .................................................................................... .................................................................................... .................................................................................... .................................................................................... .................................................................................... .................................................................................... .................................................................................... .................................................................................... .................................................................................... Unlimited. 166.70. 90.80. 22.82. 8.66. 4.01. 3.01. 1.49. 1.00. rwilkins on PROD1PC63 with PROPOSALS TABLE 3—ALLOWABLE NUMBER OF FLAWS IN PLATES OR FORGING Allowable number of cumulative flaws per 1000 square inches of inside diameter surface area in forgings or plates in the ASME section XI Appendix VIII supplement 4 inspection volume 8 ASME section XI flaw size per IWA– 3200 Range of Through-Wall Extent (TWE) of flaw [in.] 0.05 .................................................... 0.10 .................................................... 0.15 .................................................... 0.025 ≤ TWE < 0.075 .................................................................................... 0.075 ≤ TWE < 0.125 .................................................................................... 0.125 ≤ TWE < 0.175 .................................................................................... VerDate Aug<31>2005 17:37 Aug 08, 2008 Jkt 214001 PO 00000 Frm 00012 Fmt 4702 Sfmt 4702 E:\FR\FM\11AUP1.SGM 11AUP1 Unlimited 8.049 3.146 EP11AU08.024</MATH> ≤ ≤ ≤ ≤ ≤ ≤ ≤ ≤ ≤ EP11AU08.023</MATH> 0.025 0.075 0.125 0.175 0.225 0.275 0.325 0.375 0.425 EP11AU08.031</MATH> .................................................... .................................................... .................................................... .................................................... .................................................... .................................................... .................................................... .................................................... .................................................... EP11AU08.030</MATH> 0.05 0.10 0.15 0.20 0.25 0.30 0.35 0.40 0.45 Range of Through-Wall Extent (TWE) of flaw [in.] EP11AU08.029</MATH> ASME section XI flaw size per IWA– 3200 Allowable number of cumulative flaws per 1000 inches of weld length in the ASME section XI Appendix VIII supplement 4 inspection volume 46569 Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules TABLE 3—ALLOWABLE NUMBER OF FLAWS IN PLATES OR FORGING—Continued ASME section XI flaw size per IWA– 3200 0.20 0.25 0.30 0.35 Range of Through-Wall Extent (TWE) of flaw [in.] .................................................... .................................................... .................................................... .................................................... 8 Excluding Allowable number of cumulative flaws per 1000 square inches of inside diameter surface area in forgings or plates in the ASME section XI Appendix VIII supplement 4 inspection volume 8 0.175 0.225 0.275 0.325 ≤ ≤ ≤ ≤ TWE TWE TWE TWE < < < < 0.225 0.275 0.325 0.375 .................................................................................... .................................................................................... .................................................................................... .................................................................................... 0.853 0.293 0.0756 0.0144 underclad cracks in forgings. TABLE 4—CONSERVATIVE ESTIMATES FOR CHEMICAL ELEMENT WEIGHT PERCENTAGES Materials P Plates ................ TABLE 4—CONSERVATIVE ESTIMATES FOR CHEMICAL ELEMENT WEIGHT PERCENTAGES—Continued Mn 0.014 Materials 1.45 P Forgings ............ Materials Mn 0.016 TABLE 4—CONSERVATIVE ESTIMATES FOR CHEMICAL ELEMENT WEIGHT PERCENTAGES—Continued 1.11 P Welds ................ Mn 0.019 1.63 TABLE 5—MAXIMUM HEAT-AVERAGE RESIDUAL [°F] FOR RELEVANT MATERIAL GROUPS BY NUMBER OF AVAILABLE DATA POINTS [Significance level = 1%] Number of available data points s [°F] Material group 3 Welds, for Cu > 0.072 ...................................................................................... Plates, for Cu > 0.072 ...................................................................................... Forgings, for Cu > 0.072 .................................................................................. Weld, Plate or Forging, for Cu ≤ 0.072 ........................................................... TABLE 6—TMAX VALUES FOR THE SLOPE DEVIATION TEST 3 ................................................ 4 ................................................ 5 ................................................ 6 ................................................ 7 ................................................ 8 ................................................ 9 ................................................ 10 .............................................. 11 .............................................. 12 .............................................. 14 .............................................. 15 .............................................. TMAX 31.82 6.96 4.54 3.75 3.36 3.14 3.00 2.90 2.82 2.76 2.68 2.65 rwilkins on PROD1PC63 with PROPOSALS TABLE 7—THRESHOLD VALUES FOR THE OUTLIER DEVIATION TEST (SIGNIFICANCE LEVEL = 1%) Number of available data points (n) Second largest allowable normalized residual value (r*) Largest allowable normalized residual value (r*) 3 ........................ 4 ........................ 5 ........................ 1.55 1.73 1.84 2.71 2.81 2.88 VerDate Aug<31>2005 17:37 Aug 08, 2008 Jkt 214001 5 6 7 8 35.5 28.5 26.4 25.0 30.8 24.7 22.8 21.7 27.5 22.1 20.4 19.4 25.1 20.2 18.6 17.7 23.2 18.7 17.3 16.4 21.7 17.5 16.1 15.3 TABLE 7—THRESHOLD VALUES FOR DEPARTMENT OF TRANSPORTATION THE OUTLIER DEVIATION TEST (SIGNIFICANCE LEVEL = 1%)—Contin- Federal Aviation Administration ued 14 CFR Part 39 [Significance level = 1%] Number of available data points (n) 26.4 21.2 19.6 18.6 4 Number of available data points (n) Second largest allowable normalized residual value (r*) Largest allowable normalized residual value (r*) 1.93 2.00 2.05 2.11 2.16 2.19 2.23 2.26 2.29 2.32 2.93 2.98 3.02 3.06 3.09 3.12 3.14 3.17 3.19 3.21 6 ........................ 7 ........................ 8 ........................ 9 ........................ 10 ...................... 11 ...................... 12 ...................... 13 ...................... 14 ...................... 15 ...................... Dated at Rockville, Maryland, this 24th day of July 2008. For the Nuclear Regulatory Commission. R.W. Borchardt, Executive Director for Operations. [FR Doc. E8–18429 Filed 8–8–08; 8:45 am] BILLING CODE 7590–01–P PO 00000 Frm 00013 Fmt 4702 Sfmt 4702 [Docket No. FAA–2008–0857; Directorate Identifier 2007–NM–317–AD] RIN 2120–AA64 Airworthiness Directives; Dornier Model 328–300 Airplanes Federal Aviation Administration (FAA), Department of Transportation (DOT). ACTION: Notice of proposed rulemaking (NPRM). AGENCY: SUMMARY: The FAA proposes to supersede an existing airworthiness directive (AD) that applies to all AvCraft Dornier Model 328–300 airplanes. The existing AD currently requires modifying the electrical wiring of the fuel pumps; installing insulation at the flow control and shut-off valves, and other components of the environmental control system; installing markings at fuel wiring harnesses; replacing the wiring harness of the auxiliary fuel system with a new wiring harness; and installing insulated couplings in the fuel E:\FR\FM\11AUP1.SGM 11AUP1

Agencies

[Federal Register Volume 73, Number 155 (Monday, August 11, 2008)]
[Proposed Rules]
[Pages 46557-46569]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E8-18429]


========================================================================
Proposed Rules
                                                Federal Register
________________________________________________________________________

This section of the FEDERAL REGISTER contains notices to the public of 
the proposed issuance of rules and regulations. The purpose of these 
notices is to give interested persons an opportunity to participate in 
the rule making prior to the adoption of the final rules.

========================================================================


Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / 
Proposed Rules

[[Page 46557]]



NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

RIN 3150-AI01
[NRC-2007-0008]


Alternate Fracture Toughness Requirements for Protection Against 
Pressurized Thermal Shock Events

AGENCY: Nuclear Regulatory Commission.

ACTION: Supplemental Proposed Rule.

-----------------------------------------------------------------------

SUMMARY: The Nuclear Regulatory Commission (NRC) is considering the 
adoption of provisions regarding applicability of the rule and new 
provisions regarding procedures to perform surveillance data checks 
related to the updated fracture toughness requirements for protection 
against pressurized thermal shock (PTS) events for pressurized water 
reactor (PWR) pressure vessels. The NRC is considering these provisions 
as an alternative to the provisions previously noticed for public 
comment on October 3, 2007 (72 FR 56275).

DATES: Submit comments on this proposed rule by September 10, 2008. 
Submit comments on the information collection aspects on this proposed 
rule by September 10, 2008.

ADDRESSES: You may submit comments by any one of the following methods. 
Please include the following number RIN 3150-AI01 in the subject line 
of your comments. Comments submitted in writing or in electronic form 
will be made available for public inspection. Because your comments 
will not be edited to remove any identifying or contact information, 
the NRC cautions you against including any information in your 
submission that you do not want to be publicly disclosed.
    Federal e Rulemaking Portal: Go to http://www.regulations.gov and 
search for documents filed under Docket ID NRC-2007-0008. Address 
questions about NRC dockets to Carol Gallagher (301) 415-5905; e-mail 
Carol.Gallager@nrc.gov.
    Mail comments to: Secretary, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.
    E-mail comments to: Rulemaking.Comments@nrc.gov. If you do not 
receive a reply e-mail confirming that we have received your comments, 
contact us directly at (301) 415-1966.
    Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland 
20852, between 7:30 am and 4:15 pm during Federal workdays. (Telephone 
(301) 415-1966).
    Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at 
(301) 415-1101.
    You can access publicly available documents related to this 
document using the following methods:
    NRC's Public Document Room (PDR): The public may examine publicly 
available documents at the NRC's PDR, Public File Area O-F21, One White 
Flint North, 11555 Rockville Pike, Rockville, Maryland. The PDR 
reproduction contractor will copy documents for a fee.
    NRC's Agencywide Document Access and Management System (ADAMS): 
Publicly available documents created or received at the NRC are 
available electronically at the NRC's Electronic Reading Room at http:/
/www.nrc.gov/reading-rm/adams.html. From this page, the public can gain 
entry into ADAMS, which provides text and image files of NRC's public 
documents. If you do not have access to ADAMS or if there are problems 
in accessing the documents located in ADAMS, contact the NRC's PDR 
reference staff at 1-800-397-4209, or (301) 415-4737, or by e-mail to 
PDR.Resource@nrc.gov.

FOR FURTHER INFORMATION CONTACT: Ms. Veronica M. Rodriguez, Office of 
Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001; telephone (301) 415-3703; e-mail: 
Veronica.Rodriguez@nrc.gov, Mr. Barry Elliot, Office of Nuclear Reactor 
Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001; telephone (301) 415-2709; e-mail: Barry.Elliot@nrc.gov, or Mr. 
Mark Kirk, Office of Nuclear Regulatory Research, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001; telephone (301) 415-
6015; e-mail: Mark.Kirk@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Introduction
II. Background
III. Discussion
IV. Responses to Comments on the Proposed Rule
V. Section-by-Section Analysis
VI. Specific Request for Comments
VII. Availability of Documents
VIII. Plain Language
IX. Voluntary Consensus Standards
X. Finding of No Significant Environmental Impact: Availability
XI. Paperwork Reduction Act Statement
XII. Regulatory Analysis
XIII. Regulatory Flexibility Act Certification
XIV. Backfit Analysis

I. Introduction

    The NRC published a proposed rule on alternate fracture toughness 
requirements for protection against Pressurized Thermal Shock (PTS) for 
public comments in the Federal Register on October 3, 2007 (72 FR 
56275). This rule provides new PTS requirements based on updated 
analysis methods. This action is desirable 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, while 
maintaining adequate safety. These new requirements would be utilized 
by any Pressurized Water Reactor (PWR) licensee as an alternative to 
complying with the existing requirements.
    During the development of the PTS final rule, the NRC determined 
that several changes to the proposed rule language may be needed to 
adequately address issues raised in stakeholder's comments. The NRC 
also determined, in response to a stakeholder comment, that the 
characteristics of advanced PWR designs were not considered in the 
technical analysis made for the proposed rule. The NRC does not have 
assurance that reactors that commence commercial power operation after 
the effective date of this rule will have operating characteristics and 
materials of fabrication similar to those evaluated as part of the 
technical basis for the proposed rule. Therefore, the NRC has concluded 
that it would be prudent to

[[Page 46558]]

limit the applicability and the use of Sec.  50.61a to currently-
operating plants only, and proposes to modify the applicability 
provisions of the proposed rule accordingly.
    Also, several stakeholders questioned the accuracy and validity of 
the generic embrittlement curves in the proposed rule. The NRC wants to 
ensure that the predicted values from the proposed embrittlement trend 
curves provide an adequate basis for implementation of the rule. 
Therefore, the NRC has continued to work on statistical procedures to 
identify deviations from generic embrittlement trends, such as those 
described in Sec.  50.61a(f)(6) of the proposed rule. Based on this 
work, the NRC is considering enhancing the procedure described in 
paragraph Sec.  50.61a(f)(6) to, among other things, detect signs from 
the plant- and heat-specific surveillance data of embrittlement trends 
that are not reflected by Equations 5, 6 and 7 of the rule that may 
emerge at high fluences.
    Because these proposed modifications may not represent a logical 
outgrowth from the October 2007 proposed rule's provisions, the NRC 
concludes that obtaining stakeholder feedback on the proposed 
alternative provisions through the use of a supplemental proposed rule 
is appropriate. As discussed in Section VI of this notice, the NRC will 
consider comments on Sec. Sec.  50.61a(b); (f)(6)(i) through 
(f)(6)(vi); Equations 10, 11, and 12 in Sec.  50.61a(g); and Tables 5, 
6, and 7 of this supplemental proposed rule. The NRC is also requesting 
comments on whether there should be additional language added to Sec.  
50.61a(e) to allow licensees to account for the effects of sizing 
errors. This supplemental proposed rule does not reflect other 
modifications or editorial and conforming changes that the NRC is 
considering to incorporate in the final rule as a result of the public 
comments on the October 2007 proposed rule.

II. Background

    PTS events are system transients in a PWR in which severe 
overcooling occurs coincident with high pressure. The thermal stresses 
are caused by rapid cooling of the reactor vessel inside surface, which 
combine with the stresses caused by high pressure. The aggregate effect 
of these stresses is an increase in the potential for fracture if a 
pre-existing 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, can be susceptible to brittle 
fracture.
    The PTS rule, described in Sec.  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 Sec.  50.61 may 
continue to operate provided they can demonstrate a mean through-wall 
crack frequency (TWCF) from PTS-related events of no greater than 5 x 
10-6 per reactor year.
    Any reactor vessel with materials predicted to exceed the screening 
criteria in Sec.  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, plant modifications to reduce PTS event probability or 
severity, and reactor vessel annealing, which are addressed in 
Sec. Sec.  50.61(b)(3), (b)(4), and (b)(7); and Sec.  50.66, 
``Requirements for Thermal Annealing of the Reactor Pressure Vessel.''
    Currently, no operating PWR reactor vessel is projected to exceed 
the Sec.  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) developed a 
technical basis that supports updating the PTS regulations. This 
technical basis concluded that the risk of through-wall cracking due to 
a PTS event is much lower than previously estimated. This finding 
indicated that the screening criteria in Sec.  50.61 are unnecessarily 
conservative and may impose an unnecessary burden on some licensees. 
Therefore, the NRC created a new rule, Sec.  50.61a, which provides 
alternate screening criteria and corresponding embrittlement 
correlations based on the updated technical basis. 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, in a Staff Requirements Memorandum 
(SRM) dated June 30, 2006, to prepare a rulemaking which would allow 
current PWR licensees to implement the new requirements of Sec.  50.61a 
or continue to comply with the current requirements of Sec.  50.61. 
Alternatively, the NRC could have revised Sec.  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 NRC published the proposed rulemaking on the alternate fracture 
toughness requirements for protection against PTS for public comment in 
the Federal Register on October 3, 2007 (72 FR 56275). The proposed 
rule provided an alternative to the current rule, which a licensee may 
choose to adopt. This prompted the NRC to keep the current requirements 
separate from the new alternative requirements. As a result, the 
proposed rule retained the current requirements in Sec.  50.61 for PWR 
licensees choosing not to implement the less restrictive screening 
limits, and presented new requirements in Sec.  50.61a as an 
alternative relaxation for PWR licensees.

III. Discussion

    The NRC published a proposed new rule, Sec.  50.61a (October 3, 
2007, 72 FR 56275), that would provide new PTS requirements based on 
updated analysis methods because the existing requirements are based on 
unnecessarily conservative probabilistic fracture mechanics analyses. 
Stakeholders' comments raised concerns related to the applicability of 
the rule and the accuracy and validity of the generic embrittlement 
curves. The NRC reconsidered the technical and regulatory issues in 
these areas and is considering adopting the modified provisions 
regarding the applicability of the rule and new provisions regarding 
procedures to perform surveillance data checks described in this 
supplemental proposed rule. The NRC will consider comments on 
Sec. Sec.  50.61a(b), (f)(6)(i) through (f)(6)(vi); Equations 10, 11 
and 12 in Sec.  50.61a(g); and Tables 5, 6, and 7 of this supplemental 
proposed rule. As described in Section VI of this notice, the NRC is 
also requesting comments on whether there should be additional language 
added to Sec.  50.61a(e) to allow licensees to account for the effects 
of sizing errors. The NRC will consider the October 2007 proposed rule, 
the supplemental proposed rule, and the comments received in response

[[Page 46559]]

to both, when deciding whether to adopt a final PTS rule.

Applicability of the Proposed Rule, Sec.  50.61a(b)

    The supplemental proposed rule differs from the proposed rule and 
from Sec.  50.61 in that it proposes to limit the use of Sec.  50.61a 
to currently operating plants only. It cannot be demonstrated, a 
priori, that reactors which commence commercial power operation after 
the effective date of this rule will have operating characteristics, in 
particular identified PTS event sequences and thermal-hydraulic 
responses, which are consistent with the reactors which were evaluated 
as part of the technical basis for this rule. Other factors, including 
materials of fabrication and welding methods, could also vary. Hence, 
the use of Sec.  50.61a would be limited to currently operating PWR 
facilities which are known to have characteristics consistent with 
those assumed in the technical basis. The NRC also proposes to allow 
the holder of the operating license for Watts Bar Unit 2 to adopt the 
requirements in Sec.  50.61a as this facility has operating 
characteristics consistent with those assumed in the technical basis. 
The NRC recognizes that licensees for reactors who commence commercial 
power operation after the effective date of this rule may, under the 
provisions of Sec.  50.12, seek an exemption from Sec.  50.61a(b) to 
apply this rule if a plant-specific basis analyzing their operating 
characteristics, materials of fabrications, and welding methods is 
provided.

Surveillance Data, Sec.  50.61a(f)

    Section 50.61a(f) of the proposed rule defines the process for 
calculating the values for the material properties (i.e. , 
RTMAX-X) for a particular reactor vessel. These values would 
be based on the vessel material's copper, manganese, phosphorus, and 
nickel weight percentages, reactor cold leg temperature, and fast 
neutron flux and fluence values, as well as the unirradiated nil-
ductility transition reference temperature (i.e., RTNDT).
    Section 50.61a(f) of the proposed rule included a procedure by 
which the RTMAX-X values, which are predicted for plant-
specific materials using a generic temperature shift (i.e., 
[Delta]T30) embrittlement trend curve, are compared with 
heat-specific surveillance data that are collected as part of 10 CFR 
Part 50, Appendix H surveillance programs. The purpose of this 
comparison is to assess how well the surveillance data are represented 
by the generic embrittlement trend curve. If the surveillance data are 
close (closeness is assessed statistically) to the generic 
embrittlement trend curve, then the predictions of this embrittlement 
trend curve are used. This is expected to normally be the case. 
However, if the heat-specific surveillance data deviate significantly, 
and non-conservatively, from the predictions of the generic 
embrittlement trend curve, this indicates that alternative methods 
(i.e., other than, or in addition to, the generic embrittlement trend 
curve) may be needed to reliably predict the temperature shift trends, 
and to estimate RTMAX-X, for the conditions being assessed. 
However, alternative methods for temperature shift prediction are not 
prescribed by Sec.  50.61a(f) of the proposed rule.
    Although standard and accepted procedures exist to assess the 
statistical significance of the differences between heat-specific 
surveillance data and the generic embrittlement trend curve, similarly 
standard and acceptable procedures are not available to assess the 
practical importance of such differences. The practical importance of 
statistically significant deviations is best assessed by licensees on a 
case-by-case basis, which would be submitted for the review of the 
Director of NRR, as prescribed by Sec.  50.61a(f).
    The method described in the proposed rulemaking to compare the 
heat-specific surveillance data collected as part of 10 CFR part 50, 
Appendix H surveillance programs to the generic temperature shift 
embrittlement trend curve included a single statistical test. This 
statistical test was set forth by Equations 9 and 10, and Table 5. This 
test determined if, on average, the temperature shift from the 
surveillance data was significantly higher than the temperature shift 
of the generic embrittlement trend curve. The NRC has determined that, 
while necessary, this single test is not sufficient to ensure that the 
temperature shift predicted by the embrittlement trend curve well 
represents the heat-specific surveillance data. Specifically, this 
single statistical test cannot determine if the temperature shift from 
the surveillance data shows a more rapid increase after significant 
radiation exposure than the progression predicted by the generic 
embrittlement trend curve. To address this potential deficiency, which 
could be particularly important during a plant's period of extended 
operation, the NRC added two more statistical tests in this 
supplemental proposed rulemaking, which are expressed by Equations 11 
and 12 and by Tables 6 and 7. Together, these two additional tests 
determine if the surveillance data from a particular heat show a more 
rapid increase after significant radiation exposure than the 
progression predicted by the generic embrittlement trend curve.
    The NRC documented the technical basis for the proposed alternative 
in the following reports: (1) ``Statistical Procedures for Assessing 
Surveillance Data for 10 CFR Part 50.61a,'' (ADAMS Accession No. 
ML081290654), and (2) ``A Physically Based Correlation of Irradiation 
Induced Transition Temperature Shifts for RPV Steel,'' (ADAMS Accession 
No. ML081000630).

IV. Responses to Comments on the Proposed Rule

    The NRC received 5 comment letters on the proposed 10 CFR 50.61a 
rule published on October 3, 2007 (72 FR 56275). The following 
paragraphs discuss those comments which are directly associated with 
the supplemental proposed rule's provisions on the applicability of the 
rule and surveillance data procedures. The remainder of the comments 
and the NRC responses will be provided in the Federal Register notice 
for the final rule.

Comments on the Applicability of the Proposed Rule

    Comment: The commenters stated that the rule, as written, is only 
applicable to the existing fleet of PWRs. The characteristics of 
advanced PWR designs were not considered in the analysis. The 
commenters suggested adding a statement to state that this rule is 
applicable to the current PWR fleet and not the new plant designs. 
[PWROG-5, EPRI-5]
    Response: The NRC agrees with the comment that this rule is only 
applicable to the existing fleet of PWRs. The NRC cannot be assured 
that reactors that commence commercial power operation after the 
effective date of this rule will have operating characteristics, in 
particular identified PTS event sequences and thermal-hydraulic 
responses, which are consistent with the reactors that were evaluated 
as part of the technical basis for Sec.  50.61a. Other factors, 
including materials of fabrication and welding methods, could also 
vary. Therefore, the NRC agrees with the commenters that it would be 
prudent to restrict the use of Sec.  50.61a to current plants. As a 
result of this comment, the NRC proposes to modify Sec.  50.61a(b) and 
the statement of considerations of the rule to reflect this position to 
limit the use of the rule to currently operating plants.

Comments on Surveillance Data

    Comment: The commenters stated that there is little added value in 
the requirement to assess the surveillance

[[Page 46560]]

data as a part of this rule because variability in data has already 
been accounted for in the derivation of the embrittlement correlation.
    The commenters also stated that there is no viable methodology for 
adjusting the projected [Delta]T30 for the vessel based on 
the surveillance data. Any effort to make this adjustment is likely to 
introduce additional error into the prediction. Note that the 
embrittlement correlation described in the basis for the revised PTS 
rule (i.e., NUREG-1874) was derived using all of the currently 
available industry-wide surveillance data.
    In the event that the surveillance data does not match the 
[Delta]T30 value predicted by the embrittlement correlation, 
the best estimate value for the pressure vessel material is derived 
using the embrittlement correlation. The likely source of the 
discrepancy is an error in the characterization of the surveillance 
material or of the irradiation environment. Therefore, unless the 
discrepancy can be resolved, obtaining the [Delta]T30 
prediction based on the best estimate chemical composition for the heat 
of the material is more reliable than a prediction based on a single 
set of surveillance measurements.
    The commenters suggested removing the requirement to assess 
surveillance data, including Table 5, of this rule. [PWROG-4, EPRI-4, 
NEI-2]
    Response: The NRC does not agree with the proposed change. The NRC 
believes that there is added value in the requirement to assess 
surveillance data. Although variability has been accounted for in the 
derivation of the embrittlement correlation, it is the NRC's view that 
the surveillance assessment required in Sec.  50.61a(f)(6) is needed to 
determine if the embrittlement for a specific heat of material in a 
reactor vessel is consistent with the embrittlement predicted by the 
embrittlement correlation.
    The commenters also assert that there is no viable methodology for 
adjusting the projected [Delta]T30 for the vessel based on 
the surveillance data, and that any adjustment is likely to introduce 
additional error into the prediction. The NRC believes that although 
there is no single methodology for adjusting the projected 
[Delta]T30 for the vessel based on the surveillance data, it 
is possible, on a case-specific basis, to justify adjustments to the 
generic [Delta]T30 prediction. For this reason the rule does 
not specify a method for adjusting the [Delta]T30 value 
based on surveillance data, but rather requires the licensee to propose 
a case-specific [Delta]T30 adjustment procedure for review 
and approval from the Director. Although the commenters assert that it 
is possible that error could be introduced, it is the NRC view that 
appropriate plant-specific adjustments based upon available 
surveillance data may be necessary to project reactor pressure vessel 
embrittlement for the purpose of this rule.
    As the result of these public comments, the NRC has continued to 
work on statistical procedures to identify deviations from generic 
embrittlement trends, such as those described in Sec.  50.61a(f)(6) of 
the proposed rule. Based on this work, the NRC is considering further 
enhancing the procedure described in paragraph (f)(6) to, among other 
things, detect signs from the plant- and heat-specific surveillance 
data that may emerge at high fluences of embrittlement trends that are 
not reflected by Equations 5, 6, and 7. The empirical basis for the 
NRC's concern regarding the potential for un-modeled high fluence 
effects is described in documents located at ADAMS Accession Nos. 
ML081120253, ML081120289, ML081120365, ML081120380, and ML081120600. 
The technical basis for the enhanced surveillance assessment procedure 
is described in the document located at ADAMS Accession No. 
ML081290654.

V. Section-by-Section Analysis

    The following section-by-section analysis only discusses the 
modifications in the provisions related to the applicability of the 
rule and surveillance data procedures that the NRC is considering as an 
alternative in this supplemental proposed rule. The NRC is only seeking 
comments on these alternative provisions. This supplemental proposed 
rule does not reflect other modifications or editorial and conforming 
changes that the NRC is considering to incorporate as a result of the 
public comments on the proposed rule that were not discussed in this 
notice as they will be provided in the Federal Register notice for the 
final rule.

Proposed Sec.  50.61a(b)

    The proposed language for Sec.  50.61a(b) would establish the 
applicability of the rule. The NRC proposes to modify this paragraph to 
limit the use of this rule to currently-operating plants only.

Proposed Sec.  50.61a(f)(6)(i)

    The proposed language for Sec.  50.61a(f)(6)(i) would establish the 
requirements to perform data checks to determine if the surveillance 
data show a significantly different trend than what the embrittlement 
model in this rule predicts. The NRC proposes to modify Sec.  
50.61a(f)(6)(i)(B) to state that licensees would evaluate the 
surveillance for consistency with the embrittlement model by following 
the procedures specified by Sec. Sec.  50.61a(f)(6)(ii), (f)(6)(iii), 
and (f)(6)(iv) of the supplemental proposed rule.

Proposed Sec.  50.61a(f)(6)(ii)

    The proposed language for Sec.  50.61a(f)(6)(ii) would establish 
the requirements to perform an estimate of the mean deviation of the 
data set from the embrittlement model. The mean deviation for the data 
set would be compared to values given in Table 5 or Equation 10 of this 
section. The NRC proposes to modify this paragraph to state that the 
surveillance data analysis would follow the criteria in Sec. Sec.  
50.61a(f)(6)(v) and (f)(6)(vi) of the supplemental proposed rule.

Proposed Sec.  50.61a(f)(6)(iii)

    The NRC proposes to modify Sec.  50.61a(f)(6)(iii) to establish the 
requirements to estimate the slope of the embrittlement model residuals 
(i.e., the difference between the measured and predicted value for a 
specific data point). The licensee would estimate the slope using 
Equation 11 and compare this value to the maximum permissible value in 
Table 6, both from the supplemental proposed rule. This surveillance 
data analysis would follow the criteria in Sec. Sec.  50.61a(f)(6)(v) 
and (f)(6)(vi) of the supplemental proposed rule.

Proposed Sec.  50.61a(f)(6)(iv)

    The NRC proposes to modify Sec.  50.61a(f)(6)(iv) to establish the 
requirements to estimate an outlier deviation from the embrittlement 
model for the specific data set using Equations 8 and 12. The licensee 
would compare the normalized residuals to the allowable values in Table 
7 of the supplemental proposed rule. This surveillance data analysis 
would follow the criteria in Sec. Sec.  50.61a(f)(6)(v) and (f)(6)(vi) 
of the supplemental proposed rule.

Proposed Sec.  50.61a(f)(6)(v)

    The NRC proposes to add paragraph (f)(6)(v) to establish the 
criteria to be satisfied in order to calculate the 
[Delta]T30 shift values.

Proposed Sec.  50.61a(f)(6)(vi)

    The NRC proposes to add paragraph (f)(6)(vi) to establish the 
actions to be taken by a licensee if the criteria in paragraph 
(f)(6)(v) of this section are not met. The licensee would need to 
submit an evaluation of the surveillance data and propose values for 
[Delta]T30, considering their plant-specific

[[Page 46561]]

surveillance data, for the review and approval by the Director. The 
licensee would need to submit an evaluation of each surveillance 
capsule removed from the vessel after the submittal of the initial 
application for review and approval by the Director no later than 2 
years after the capsule is withdrawn from the vessel.

Proposed Sec.  50.61a(g)

    The proposed language for Sec.  50.61a(g) would provide the 
necessary equations and variables required by the proposed changes in 
Sec.  50.61a(f)(6). The NRC proposes to modify Equation 10 to account 
for 1 percent of significance level. Equations 11 and 12 would be added 
to provide the means for estimating the slope and the outlier deviation 
from the embrittlement model.

Proposed Tables 5, 6, and 7

    Tables 5, 6, and 7 would provide values to be used in the proposed 
changes in Sec.  50.61a(f)(6). The NRC proposes to modify Table 5 to 
account for the use of a 1 percent of significance level. Tables 6 and 
7 would be added to provide the threshold values for the slope and the 
outlier deviation tests.

VI. Specific Request for Comments

    The NRC seeks comments on Sec. Sec.  50.61a(b), (f)(6)(i) through 
(f)(6)(vi); Equations 10, 11, and 12 in Sec.  50.61a(g), and Tables 5, 
6, and 7 of the supplemental proposed rule. The NRC is not seeking 
comments on any other provisions of the proposed Sec.  50.61a which 
remain unchanged from the October 2007 proposed rule. In addition, the 
NRC also requests comments on the following question:

Adjustments of the Inservice Inspection Volumetric Examination and Flaw 
Assessments

    The flaw sizes in Tables 2 and 3 are selected so that reactor 
vessels with flaw sizes less than or equal to those in the tables will 
have a TWCF less than or equal to 1 x 10-6 per reactor year 
at the maximum permissible embrittlement. The NRC recognizes that the 
flaw sizes in these tables represent actual flaw dimensions while the 
results from the ASME Code examinations are estimated dimensions. The 
available information indicates that, for most flaw sizes in Tables 2 
and 3, qualified inspectors will oversize flaws. Comparing oversized 
flaws to the size and density distributions in Tables 2 and 3 is 
conservative and acceptable, but not necessary. Therefore, NRC is 
considering to permit flaw sizes to be adjusted to account for the 
effects of sizing error before comparing the estimated size and density 
distribution to the acceptable size and density distributions in Tables 
2 and 3. This would be accomplished by requiring licensees to base the 
methodology to account for the effects of sizing error on statistical 
data collected from ASME Code inspector qualification tests. An 
acceptable method would include a demonstration, that accounting for 
the effects of sizing error, is unlikely to result in accepting actual 
flaw size distribution that cause the TWCF to exceed the acceptance 
criteria. Adjusting flaw sizes to account for sizing error can change 
an unacceptable examination result into an acceptable result; further, 
collecting, evaluating, and using data from ASME Code inspector 
qualification tests will require extensive engineering judgment. 
Therefore, the methodology would have to be reviewed and approved by 
the Director of the NRC's Office of Nuclear Reactor Regulation (NRR) to 
ensure that the risk associated with PTS is acceptable. The NRC 
requests specific comments on whether there should be additional 
language added to 10 CFR 50.61a(e) to allow licensees to account for 
the effects of sizing errors.

VII. Availability of Documents

    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.
    Regulations.gov (Web). These documents may be viewed and downloaded 
electronically through the Federal eRulemaking Portal http://
www.regulations.gov, Docket number NRC-2007-0008.
    NRC's Electronic Reading Room (ERR). The NRC's public electronic 
reading room is located at http://www.nrc.gov/reading-rm.html.

------------------------------------------------------------------------
          Document               PDR           Web          ERR (ADAMS)
------------------------------------------------------------------------
Federal Register Notice--           X      NRC-2007-0008     ML072750659
 Proposed Rule: Alternate
 Fracture Toughness
 Requirements for Protection
 Against Pressurized Thermal
 Shock Events (RIN 3150-
 AI01), 72 FR 56275, October
 3, 2007....................
Letter from Thomas P.               X      NRC-2007-0008     ML073521542
 Harrall, Jr., dated
 December 17, 2007,
 ``Comments on Proposed Rule
 10 CFR 50, Alternate
 Fracture Toughness
 Requirements for Protection
 Against Pressurized Thermal
 Shock Events, RIN 3150-
 AI01'' [Identified as Duke]
Letter from Jack Spanner,           X      NRC-2007-0008     ML073521545
 dated December 17, 2007,
 ``10 CFR 50.55a Proposed
 Rulemaking Comments RIN
 3150-AI01'' [Identified as
 EPRI]......................
Letter from James H. Riley,         X      NRC-2007-0008     ML073521543
 dated December 17, 2007,
 ``Proposed Rulemaking--
 Alternate Fracture
 Toughness Requirements for
 Protection Against
 Pressurized Thermal Shock
 Events (RIN 3150-AI01), 72
 FR 56275, October 3, 2007
 [Identified as NEI]........
Letter from Melvin L. Arey,         X      NRC-2007-0008     ML073521547
 dated December 17, 2007,
 ``Transmittal of PWROG
 Comments on the NRC
 Proposed Rule on Alternate
 Fracture Toughness
 Requirements for Protection
 Against Pressurized Thermal
 Shock Events'', RIN 3150-
 AI01, PA-MSC-0232
 [Identified as PWROG]......
Letter from T. Moser, dated         X      NRC-2007-0008     ML073610558
 December 17, 2007,
 ``Strategic Teaming and
 Resource Sharing (STARS)
 Comments on RIN 3150-AI01,
 Alternate Fracture
 Toughness Requirements for
 Protection against
 Pressurized Thermal Shock
 Events 72 FR 56275 (October
 3,2007) [Identified as
 STARS].....................
``Statistical Procedures for        X   ................     ML081290654
 Assessing Surveillance Data
 for 10 CFR Part 50.61a''...
``A Physically Based                X   ................     ML081000630
 Correlation of Irradiation
 Induced Transition
 Temperature Shifts for RPV
 Steel''....................
Supplemental Regulatory             X      NRC-2007-0008     ML081440673
 Analysis...................
Supplemental OMB Supporting         X      NRC-2007-0008     ML081440736
 Statement..................
Memo from J. Uhle, dated May        X   ................     ML081120253
 15, 2008, ``Embrittlement
 Trend Curve Development for
 Reactor Pressure Vessel
 Materials''................
Draft ``Technical Basis for         X   ................     ML081120289
 Revision of Regulatory
 Guide 1.99: NRC Guidance on
 Methods to Estimate the
 Effects of Radiation
 Embrittlement on the Charpy
 V-Notch Impact Toughness of
 Reactor Vessel Materials''.

[[Page 46562]]

 
``Comparison of the                 X   ................     ML081120365
 Predictions of RM-9 to the
 IVAR and RADAMO Databases''
Memo from M. Erickson Kirk,         X   ................     ML081120380
 dated December 12, 2007,
 ``New Data from Boiling
 Water Reactor Vessel
 Integrity Program (BWRVIP)
 Integrated Surveillance
 Project (ISP)''............
``Further Evaluation of High        X   ................     ML081120600
 Fluence Data''.............
------------------------------------------------------------------------

VIII. Plain Language

    The Presidential memorandum ``Plain Language in Government 
Writing'' published in 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 heading of this notice.

IX. Voluntary Consensus Standards

    The National Technology Transfer and Advancement Act of 1995, 
Public Law 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 determined that there is only one technical standard 
developed that could be utilized for characterizing the embrittlement 
correlations. That standard is the 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 supplemental proposed rule. However, the 
correlation developed by the NRC 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 supplemental 
proposed rule because it does not represent the broad range of 
conditions necessary to justify a revision to the regulations.
    The ASME Code requirements are utilized as part of the volumetric 
examination analysis requirements of the supplemental 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 Part 50, Appendix H 
and utilized to determine 30-foot-pound transition temperatures. These 
standards were selected for use in the supplemental proposed rule based 
on their use in other regulations within 10 CFR Part 50 and their 
applicability to the subject of the desired requirements.
    The NRC will consider using a voluntary consensus standard in the 
final rule if an appropriate standard is identified in the public 
comment period for this supplemental proposed rule.

X. Finding of No Significant Environmental Impact: Availability

    The Commission has determined under the National Environmental 
Policy Act of 1969, as amended, and the Commission's regulations in 10 
CFR Part 51, Subpart A, 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. 
This determination was made as part of the proposed rulemaking issued 
on October 3, 2007 (72 FR 56275), and remains applicable to this 
supplemental proposed rulemaking.

XI. Paperwork Reduction Act Statement

    This supplemental 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 supplemental 
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 50.61 and 50.61a)'' 
supplemental 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 an alternative to the requirements of 10 CFR 50.61. 
Collections would also be required, after implementation of the new 10 
CFR 50.61a, when any change is made to the design or operation of the 
facility that affects the calculated RTMAX-X value. 
Collections would also be required during the scheduled periodic 
ultrasonic examination of beltline welds.
    Who will be required or asked to report: Licensees of currently 
operating PWRs 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: 363 hours (253 hours annually for 
record keeping plus 110 hours annually for reporting).
    Abstract: The NRC is proposing to amend its regulations to provide 
updated fracture toughness requirements for protection against PTS 
events for PWR pressure vessels. The supplemental 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 is expected to 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 utilized by licensees of currently operating PWRs 
as an alternative to complying with the existing requirements.
    The NRC is seeking public comment on the potential impact of the 
information collections contained in this supplemental 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?

[[Page 46563]]

    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-1F21, 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. The document will be 
available on the NRC home page site 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 September 10, 2008. 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. 
Comments submitted in writing or in electronic form will be made 
available for public inspection. Because your comments will not be 
edited to remove any identifying or contact information, the NRC 
cautions you against including any information in your submission that 
you do not want to be publicly disclosed. Comments submitted should 
reference Docket No. NRC-2007-0008. Comments can be submitted in 
electronic form via the Federal e-Rulemaking Portal at http://
www.regulations.gov by search for Docket No. NRC-2007-0008. Comments 
can be mailed to NRC Clearance Officer, Russell Nichols (T-5F52), U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001. Questions 
about the information collection requirements may be directed to the 
NRC Clearance Officer, Russell Nichols (T-5 F52), U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, by telephone at (301) 
415-6874, or by e-mail to INFOCOLLECTS.Resource@nrc.gov. Comments can 
be mailed to the Desk Officer, Office of Information and Regulatory 
Affairs, NEOB-10202, (3150-0011), Office of Management and Budget, 
Washington, DC 20503, or by e-mail to Nathan_J._Frey@omb.eop.gov, or 
by telephone at (202) 395-7345.

Public Protection Notification

    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.

XII. Regulatory Analysis

    The NRC has issued a supplemental regulatory analysis for this 
supplemental proposed rulemaking. The analysis examines the costs and 
benefits of the alternatives considered by the NRC. The NRC requests 
public comments on this supplemental draft regulatory analysis. 
Availability of the supplemental regulatory analysis is provided in 
Section VII of this notice. Comments on the supplemental draft 
regulatory analysis may be submitted to the NRC as indicated under the 
ADDRESSES heading of this notice.

XIII. Regulatory Flexibility Act Certification

    In accordance with the Regulatory Flexibility Act (5 U.S.C. 
605(b)), the NRC certifies that this rule would not, if promulgated, 
have a significant economic impact on a substantial number of small 
entities. This supplemental proposed rule would affect only the 
licensing and operation of currently operating 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).

XIV. Backfit Analysis

    The NRC has determined that the requirements in this supplemental 
proposed rule would 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 supplemental proposed rule. The requirements of the proposed 
PTS rule, including those in the supplemental proposed rule, would not 
be required, but could be utilized by PWR licensees with currently 
operating plants. Licensees choosing to implement the proposed PTS rule 
would be required to comply with its requirements as an 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, the NRC finds that this 
amendment would not constitute backfitting.

List of Subjects for 10 CFR Part 50

    Antitrust, Classified information, Criminal penalties, Fire 
protection, Intergovernmental relations, Nuclear power plants and 
reactors, Radiation protection, Reactor siting criteria, Reporting and 
recordkeeping requirements.

    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.

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
FACILITIES

    1. The authority citation for part 50 continues to read as follows:

    Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 
Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 
83 Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 
2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88 
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); 
sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note); sec. 651(e), Pub. 
L. 109-58, 119 Stat. 806-810 (42 U.S.C. 2014, 2021, 2021b, 2111).
    Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat. 
2951 as amended by Pub. L. 102-486, sec. 2902, 106 Stat. 3123 (42 
U.S.C. 5841). Section 50.10 also issued under secs. 101, 185, 68 
Stat. 955, as amended (42 U.S.C. 2131, 2235); sec. 102, Pub. L. 91-
190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd), and 
50.103 also issued under sec. 108, 68 Stat. 939, as amended (42 
U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued 
under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 
50.55a and appendix Q also issued under sec. 102, Pub. L. 91-190, 83 
Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued 
under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58, 
50.91, and 50.92 also issued under Pub. L. 97-415, 96 Stat. 2073 (42 
U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 
(42 U.S.C. 2152). Sections 50.80-50.81 also issued under sec. 184, 
68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued 
under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).

    2. Section 50.8(b) is revised to read as follows:


Sec.  50.8  Information collection requirements: OMB approval.

* * * * *
    (b) The approved information collection requirements contained in 
this part appear in Sec. Sec.  50.30, 50.33, 50.34, 50.34a, 50.35, 
50.36, 50.36a, 50.36b, 50.44, 50.46, 50.47, 50.48, 50.49, 50.54, 50.55, 
50.55a, 50.59, 50.60, 50.61, 50.61a, 50.62, 50.63, 50.64, 50.65, 50.66, 
50.68, 50.69, 50.70, 50.71, 50.72, 50.74, 50.75, 50.80, 50.82, 50.90, 
50.91, 50.120, and appendices A, B, E, G, H, I, J, K, M, N,O, Q, R, and 
S to this part.
* * * * *
    3. Section 50.61a is added to read as follows:


Sec.  50.61a  Alternate fracture toughness requirements for protection 
against pressurized thermal shock events.

    (a) Definitions. Terms in this section have the same meaning as 
those set

[[Page 46564]]

forth in 10 CFR 50.61(a), with the exception of the term ``ASME Code''.
    (1) ASME Code means the American Society of Mechanical Engineers 
Boiler and Pressure Vessel Code, Section III, Division I, ``Rules for 
the Construction of Nuclear Power Plant Components,'' and Section XI, 
Division I, ``Rules for Inservice Inspection of Nuclear Power Plant 
Components,'' edition and addenda and any limitations and modifications 
thereof as specified in Sec.  50.55a.
    (2) RTMAX	AW means the material property which characterizes the 
reactor vessel's resistance to fracture initiating from flaws found 
along axial weld fusion lines. RTMAX-AW is determined under 
the provisions of paragraph (f) of this section and has units of 
[deg]F.
    (3) RTMAX	PL means the material property which characterizes the 
reactor vessel's resistance to fracture initiating from flaws found in 
plates in regions that are not associated with welds found in plates. 
RTMAX-PL is determined under the provisions of paragraph (f) 
of this section and has units of [deg]F.
    (4) RTMAX	FO means the material property which characterizes the 
reactor vessel's resistance to fracture initiating from flaws in 
forgings that are not associated with welds found in forgings. 
RTMAX-FO is determined under the provisions of paragraph (f) 
of this section and has units of [deg]F.
    (5) RTMAX	CW means the material property which characterizes the 
reactor vessel's resistance to fracture initiating from flaws found 
along the circumferential weld fusion lines. RTMAX-CW is 
determined under the provisions of paragraph (f) of this section and 
has units of [deg]F.
    (6) RTMAX	X means any or all of the material properties 
RTMAX-AW, RTMAX-PL, RTMAX-FO, or 
RTMAX-CW for a particular reactor vessel.
    (7) [phis]t means fast neutron fluence for neutrons with energies 
greater than 1.0 MeV. [phis]t is determined under the provisions of 
paragraph (g) of this section and has units of n/cm\2\.
    (8) [phis] means average neutron flux. [phis] is determined under 
the provisions of paragraph (g) of this section and has units of n/
cm\2\/sec.
    (9) [Delta]T30 means the shift in the Charpy V-notch transition 
temperature produced by irradiation defined at the 30 ft-lb energy 
level. The [Delta]T30 value is determined under the 
provisions of paragraph (g) of this section and has units of [deg]F.
    (10) Surveillance data means any data that demonstrates the 
embrittlement trends for the beltline materials, including, but not 
limited to, data from test reactors or surveillance programs at other 
plants with or without a surveillance program integrated under 10 CFR 
Part 50, Appendix H.
    (11) Tc means cold leg temperature under normal full power 
operating conditions, as a time-weighted average from the start of full 
power operation through the end of licensed operation. Tc 
has units of [deg]F.
    (b) Applicability. Each licensee of a pressurized water nuclear 
power reactor, whose original operating license was issued prior to 
[EFFECTIVE DATE OF FINAL RULE], and the holder of any operating license 
issued under this part or part 54 for the Watts Bar Unit 2 facility, 
may utilize the requirements of this section as an alternative to the 
requirements of 10 CFR 50.61.
    (c) Request for Approval. Prior to implementation of this section, 
each licensee shall submit a request for approval in the form of a 
license amendment together with the documentation required by 
paragraphs (c)(1), (c)(2), and (c)(3) of this section for review and 
approval to the Director, Office of Nuclear Reactor Regulation 
(Director). The information required by paragraphs (c)(1), (c)(2), and 
(c)(3) of this section must be submitted for review and approval by the 
Director at least three years before the limiting RTPTS 
value calculated under 10 CFR 50.61 is projected to exceed the PTS 
screening criteria in 10 CFR 50.61 for plants licensed under this part.
    (1) Each licensee shall have projected values of RTMAX-X 
for each reactor vessel beltline material for the EOL fluence of the 
material. The assessment of RTMAX-X values must use the 
calculation procedures given in paragraphs (f) and (g) of this section, 
except as provided in paragraphs (f)(6) and (f)(7) of this section. The 
assessment must specify the bases for the projected value of 
RTMAX-X for each reactor vessel beltline material, including 
the assumptions regarding future plant operation (e.g., core loading 
patterns, projected capacity factors, etc.); the copper (Cu), 
phosphorus (P), manganese (Mn), and nickel (Ni) contents; the reactor 
cold leg temperature (TC); and the neutron flux and fluence 
values used in the calculation for each beltline material.
    (2) Each licensee shall perform an examination and an assessment of 
flaws in the reactor vessel beltline as required by paragraph (e) of 
this section. The licensee shall verify that the requirements of 
paragraphs (e)(1) through (e)(3) have been met and submit all 
documented indications and the neutron fluence map required by 
paragraph (e)(1)(iii) to the Director in its application to utilize 10 
CFR 50.61a. If analyses performed under paragraph (e)(4) of this 
section are used to justify continued operation of the facility, 
approval by the Director is required prior to implementation.
    (3) Each licensee shall compare the projected RTMAX-X 
values for plates, forgings, axial welds, and circumferential welds to 
the PTS screening criteria for the purpose of evaluating a reactor 
vessel's susceptibility to fracture due to a PTS event. If any of the 
projected RTMAX-X values are greater than the PTS screening 
criteria in Table 1 of this section, then the licensee may propose the 
compensatory actions or plant-specific analyses as required in 
paragraphs (d)(3) through (d)(7) of this section, as applicable, to 
justify operation beyond the PTS screening criteria in Table 1 of this 
section.
    (d) Subsequent Requirements. Licensees who have been approved to 
utilize 10 CFR 50.61a under the requirements of paragraph (c) of this 
section shall comply with the requirements of this paragraph.
    (1) Whenever there is a significant change in projected values of 
RTMAX-X, such that the previous value, the current value, or 
both values, exceed the screening criteria prior to the expiration of 
the plant operating license; or upon the licensee's request for a 
change in the expiration date for operation of the facility; a 
reassessment of RTMAX-X values documented consistent with 
the requirements of paragraph (c)(1) and (c)(3) of this section must be 
submitted for review and approval to the Director. If the Director does 
not approve the assessment of RTMAX-X values, then the 
licensee shall perform the actions required in paragraphs (d)(3) 
through (d)(7) of this section, as necessary, prior to operation beyond 
the PTS screening criteria in Table 1 of this section.
    (2) Licensees shall determine the impact of the subsequent flaw 
assessments required by paragraphs (e)(1)(i), (e)(1)(ii), (e)(2), and 
(e)(3) of this section and shall submit the assessment for review and 
approval to the Director within 120 days after completing a volumetric 
examination of reactor vessel beltline materials as required by Section 
XI of the ASME Code. If a licensee is required to implement paragraphs 
(e)(4) and (e)(5) of this section, a reanalysis in accordance with 
paragraphs (e)(4) and (e)(5) of this section is required within one 
year of the subsequent ASME Code inspection.
    (3) If the value of RTMAX-X is projected to exceed the 
PTS screening criteria, then the licensee shall implement those flux 
reduction

[[Page 46565]]

programs that are reasonably practicable to avoid exceeding the PTS 
screening criteria. The schedule for implementation of flux reduction 
measures may take into account the schedule for review and anticipated 
approval by the Director of detailed plant-specific analyses which 
demonstrate acceptable risk with RTMAX-X values above the 
PTS screening criteria due to plant modifications, new information, or 
new analysis techniques.
    (4) If the analysis required by paragraph (d)(3) of this section 
indicates that no reasonably practicable flux reduction program will 
prevent the RTMAX-X value for one or more reactor vessel 
beltline materials from exceeding the PTS screening criteria, then the 
licensee shall perform a safety analysis to determine what, if any, 
modifications to equipment, systems, and operation are necessary to 
prevent the potential for an unacceptably high probability of failure 
of the reactor vessel as a result of postulated PTS events if continued 
operation beyond the PTS screening criteria is to be allowed. In the 
analysis, the licensee may determine the properties of the reactor 
vessel materials based on available information, research results and 
plant surveillance data, and may use probabilistic fracture mechanics 
techniques. This analysis must be submitted to the Director at least 
three years before RTMAX-X is projected to exceed the PTS 
screening criteria.
    (5) After consideration of the licensee's analyses, including 
effects of proposed corrective actions, if any, submitted under 
paragraphs (d)(3) and (d)(4) of this section, the Director may, on a 
case-by-case basis, approve operation of the facility with 
RTMAX-X values in excess of the PTS screening criteria. The 
Director will consider factors significantly affecting the potential 
for failure of the reactor vessel in reaching a decision.
    (6) If the Director concludes, under paragraph (d)(5) of this 
section, that operation of the facility with RTMAX-X values 
in excess of the PTS screening criteria cannot be approved on the basis 
of the licensee's analyses submitted under paragraphs (d)(3) and (d)(4) 
of this section, then the licensee shall request a license amendment, 
and receive approval by the Director, prior to any operation beyond the 
PTS screening criteria. The request must be based on modifications to 
equipment, systems, and operation of the facility in addition to those 
previously proposed in the submitted analyses that would reduce the 
potential for failure of the reactor vessel due to PTS events, or on 
further analyses based on new information or improved methodology.
    (7) If the limiting RTMAX-X value of the facility is 
projected to exceed the PTS screening criteria and the requirements of 
paragraphs (d)(3) through (d)(6) of this section cannot be satisfied, 
the reactor vessel beltline may be given a thermal annealing treatment 
under the requirements of Sec.  50.66 to recover the fracture toughness 
of the material. The reactor vessel may be used only for that service 
period within which the predicted fracture toughness of the reactor 
vessel beltline materials satisfy the requirements of paragraphs (d)(1) 
through (d)(6) of this section, with RTMAX-X values 
accounting for the effects of annealing and subsequent irradiation.
    (e) Examination and Flaw Assessment Requirements. The volumetric 
examinations results evaluated under paragraphs (e)(1), (e)(2), and 
(e)(3) of this section must be acquired using procedures, equipment and 
personnel that have been qualified under the ASME Code, Section XI, 
Appendix VIII, Supplement 4 and Supplement 6.
    (1) The licensee shall verify that the indication density and size 
distributions within the ASME Code, Section XI, Appendix VIII, 
Supplement 4 inspection volume \1\ are within the flaw density and size 
distributions in Tables 2 and 3 of this section based on the test 
results from the volumetric examination. The allowable number of flaws 
specified in Tables 2 and 3 of this section represent a cumulative flaw 
size distribution for each ASME flaw size increment. The allowable 
number of flaws for a particular ASME flaw size increment represents 
the maximum total number of flaws in that and all larger ASME flaw size 
increments. The licensee shall also demonstrate that no flaw exceeds 
the size limitations specified in Tables 2 and 3 of this section.
---------------------------------------------------------------------------

    \1\ The ASME Code, Section XI, Appendix VIII, Supplement 4 weld 
volume is the weld volume from the clad-to-base metal interface to 
the inner 1.0 inch or 10 percent of the vessel thickness, whichever 
is greater.
---------------------------------------------------------------------------

    (i) The licensee shall determine the allowable number of weld flaws 
for the reactor vessel beltline by multiplying the values in Table 2 of 
this section by the total length of the reactor vessel beltline welds 
that were volumetrically inspected and dividing by 1000 inches of weld 
length.
    (ii) The licensee shall determine the allowable number of plate or 
forging flaws for their reactor vessel beltline by multiplying the 
values in Table 3 of this section by the total plate or forging surface 
area that was volumetrically inspected in the beltline plates or 
forgings and dividing by 1000 square inches.
    (iii) For each indication detected in the ASME Code, Section XI, 
Appendix VIII, Supplement 4 inspection volume, the licensee shall 
document the dimensions of the indication, including depth and length, 
the orientation of the indication relative to the axial direction, and 
the location within the reactor vessel, including