Notices. Notice of Availability of Final Environmental Impact Statement

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BILLING CODE 7555-01-M NUCLEAR REGULATORY COMMISSION [Docket No. 72-25] Foster Wheeler Environmental Corporation's Proposed Idaho Spent Fuel Facility; Notice of Availability of Final Environmental Impact Statement AGENCY: Nuclear Regulatory Commission. ACTION: Notice of Availability of Final Environmental Impact Statement. SUMMARY: Notice is hereby given that the U.S. Nuclear Regulatory Commission

(NRC)is issuing a Final Environmental Impact Statement (FEIS), “Environmental Impact Statement for the Proposed Idaho Spent Fuel Facility at the Idaho National Engineering and Environmental Laboratory in Butte County, Idaho,” NUREG-1773, January 2004. This FEIS was prepared to evaluate the environmental impacts of the Foster Wheeler Environmental Corporation (FWENC) proposal to construct and operate an independent spent fuel storage installation as described in it's license application dated November 19, 2001, and docketed on June 27, 2002 (67 FR 43358). If approved, the proposed facility would be licensed in accordance with NRC regulations found at 10 CFR Part 72. The FEIS discusses the purpose and need for the proposed facility and reasonable alternatives to the proposed action, including the no-action alternative. The FEIS also discusses the environment potentially affected by the proposed facility, presents and compares the potential environmental impacts resulting from the proposed action and its alternatives, and identifies mitigation measures that could eliminate or lessen the potential environmental impacts. The FEIS is being issued as part of the NRC's decision-making process on whether to issue a license to FWENC. Based on the evaluation in the FEIS, the NRC environmental review staff have concluded that the proposed action will have small effects on the public and existing environment. This FEIS reflects the final analysis of environmental impacts of FWENC's proposal and it's alternatives, including the consideration of public comments received by the NRC. In addition, the FEIS provides summaries of the substantive public comments on the draft EIS, and responses, as appropriate. ADDRESSES: The NRC maintains an Agencywide Documents Access and Management System (ADAMS), which provides text and image files of NRC's public documents. The FEIS and its appendices may be accessed through the NRC's Public Electronic Reading Room on the Internet at *http://www.nrc.gov/reading-rm/adams.html* . If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room

(PDR)Reference staff at 1-800-397-4209, 301-415-4737, or by email to *pdr@nrc.gov* . The FEIS is also available for inspection at the Commission's Public Document Room, U.S. NRC's Headquarters Building, 11555 Rockville Pike (first floor), Rockville, Maryland. Upon written request and to the extent supplies are available, a single copy of the FEIS can be obtained for a fee by writing to the Office of the Chief Information Officer, Reproduction and Distribution Services Section, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; by electronic mail at *DISTRIBUTION@nrc.gov* ; or by fax at

(301)415-2289. Information and documents associated with the Idaho Spent Fuel Facility project may also be obtained from the Internet on NRC's Idaho Spent Fuel Facility Web page: *http://www.nrc.gov/waste/spent-fuel-storage.html* (case sensitive). FOR FURTHER INFORMATION CONTACT: For environmental review questions, please contact Matthew Blevins at

(301)415-7684. For questions related to the safety review or overall licensing of the Idaho Spent Fuel Facility, please contact Randall Hall at

(301)415-1336. SUPPLEMENTARY INFORMATION: A Settlement Agreement dated October 17, 1995, among the Department of Energy (DOE), the U.S. Navy, and the State of Idaho requires, among other things, the transfer and dry storage of spent nuclear fuel

(SNF)until it can be removed from Idaho. As part of it's efforts to meet the Settlement Agreement, the DOE has contracted with FWENC to design, license, construct, and operate the proposed Idaho Spent Fuel Facility for portions of the SNF currently in storage at the Idaho National Environmental and Engineering Laboratory (INEEL). Subsequently, FWENC submitted a license application to the NRC for the receipt, transfer, and storage of SNF at the proposed facility. The proposed facility would provide the ability to remove the SNF from existing canisters, place it in specially designed storage containers, then seal and place the loaded containers into interim storage. The new containers are designed to be compatible with transportation systems and with future permanent disposal systems. The proposed facility would store SNF and associated radioactive material from the Peach Bottom Unit 1 High-Temperature Gas-Cooled Reactor, the Shippingport Atomic Power Station, and various Training, Research, and Isotope reactors built by General Atomics (TRIGA) reactors. The majority of this SNF is currently in storage at the Idaho Nuclear Technology Center located on the INEEL immediately adjacent to the proposed facility. DOE plans to transfer the SNF to the proposed facility using existing INEEL and DOE procedures. The transfers from DOE to the proposed facility would take place completely within the boundaries of the INEEL. Upon arrival at the proposed facility, the SNF would be

(1)remotely removed from the containers in which it is currently stored,

(2)visually inspected,

(3)inventoried,

(4)placed into new storage canisters, and

(5)placed into interim dry storage. The FEIS for the proposed Idaho Spent Fuel Facility was prepared by the staff of the NRC and its contractor, Center for Nuclear Waste Regulatory Analyses, in compliance with the National Environmental Policy Act (NEPA), and the NRC's regulations for implementing NEPA (10 CFR part 51). The proposed action involves a decision by NRC of whether to issue a license under the provisions of 10 CFR part 72 that would authorize FWENC to receive, transfer, and store SNF and associated radioactive materials at the proposed facility. The NRC published a Notice of Intent to prepare an environmental impact statement

(EIS)for the proposed Idaho Spent Fuel Facility and to hold a public scoping period in the **Federal Register** on July 26, 2002 (67 FR 48953). The NRC accepted scoping comments through September 16, 2002, and subsequently issued a Scoping Summary Report on December 2, 2002. The NRC published a draft EIS on June 26, 2003, and provided an opportunity to comment until August 18, 2003 (68 FR 38105, 68 FR 39940). The FEIS describes the proposed action and alternatives to the proposed action, including the no-action alternative. The FEIS assesses the impacts of the proposed action and its alternatives on human health, air quality, water resources, waste management, geology, noise, ecology, land use, cultural resources, socioeconomics, accident impacts, and environmental justice. Additionally, the FEIS analyzes and compares the costs and benefits of the proposed action. After weighing the impacts, costs, and benefits of the proposed action and comparing alternatives ( *see* Sections 2.6, 4.15, and 7 of the FEIS), the NRC staff, in accordance with 10 CFR 51.91 (d), sets forth its final NEPA recommendation regarding the proposed action. The NRC staff recommend that, unless safety issues mandate otherwise, the action called for is the issuance of the proposed license to FWENC. In this regard, the NRC staff concludes

(i)the applicable environmental monitoring program described in Section 6 of the FEIS, and

(ii)the proposed mitigation measures discussed in Section 5 of the FEIS would eliminate or substantially lessen any potential adverse environmental impacts associated with the proposed action. The NRC staff has concluded that the overall benefits of the proposed Idaho Spent Fuel Facility outweigh the disadvantages and costs, based on consideration of the following: —The proposed Idaho Spent Fuel Facility will have small impacts on the physical environment and human communities in the vicinity. Long-term impacts of the no-action alternative are likely to be similar to the impacts of the proposed action. —The proposed action is designed to support the INEEL mission and comply with agreements and commitments negotiated by DOE, including the 1995 Settlement Agreement among DOE, the State of Idaho, and the U.S. Navy to remove SNF from Idaho by 2035. —Currently, most of the SNF to be received by the proposed Idaho Spent Fuel Facility is stored at the Idaho Nuclear Technology Center. Transfer distances from current storage locations to the proposed facility are relatively short. —The current storage configuration does not prepare the SNF for shipment from INEEL to a national HLW repository. NRC staff in the Spent Fuel Project Office are currently completing the licensing and safety review of FWENC's proposed action. The final licensing decision is currently scheduled for the Spring of 2004. Dated at Rockville, Maryland, this 3rd day of February 2004. For the Nuclear Regulatory Commission. Lawrence E. Kokajko, Chief, Environmental and Performance Assessment Branch, Division of Waste Management, Office of Nuclear Material Safety and Safeguards. [FR Doc. E4-413 Filed 2-26-04; 8:45 am] BILLING CODE 7590-01-P NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on Reliability and Probabilistic Risk Assessment; Notice of Meeting The ACRS Subcommittee on Reliability and Probabilistic Risk Assessment will hold a meeting on March 25, 2004, Room T-2B1, 11545 Rockville Pike, Rockville, Maryland. The entire meeting will be open to public attendance. The agenda for the subject meeting shall be as follows: Thursday, March 25, 2004—1 p.m. Until the Conclusion of Business The purpose of this meeting is to discuss the NRC staff's draft action plan for the implementation of the phased approach to PRA Quality. The Subcommittee will hear presentations by and hold discussions with representatives of the NRC staff regarding this matter. The Subcommittee will gather information, analyze relevant issues and facts, and formulate proposed positions and actions, as appropriate, for deliberation by the full Committee. Members of the public desiring to provide oral statements and/or written comments should notify the Designated Federal Official, Mr. Michael R. Snodderly (telephone: 301-415-6927) five days prior to the meeting, if possible, so that appropriate arrangements can be made. Electronic recordings will be permitted during the meeting. Further information regarding this meeting can be obtained by contacting the Designated Federal Official between 7:30 a.m. and 4:15 p.m. (ET). Persons planning to attend this meeting are urged to contact the above named individual at least two working days prior to the meeting to be advised of any potential changes to the agenda. Dated: February 23, 2004. Sam Duraiswamy, Acting Associate Director for Technical Support, ACRS/ACNW. [FR Doc. E4-414 Filed 2-26-04; 8:45 am] BILLING CODE 7590-01-P NUCLEAR REGULATORY COMMISSION [Docket Nos. (as shown in the Attachment), License Nos. (as shown in the Attachment), EA-03-009] In the Matter of All Pressurized Water Reactor Licensees; First Revised Order Modifying Licenses I The Licensees identified in the Attachment to this Order hold licenses issued by the Nuclear Regulatory Commission (NRC or Commission) authorizing operation of pressurized water reactor

(PWR)nuclear power plants in accordance with the Atomic Energy Act of 1954 and title 10 of the Code of Federal Regulations (10 CFR) part 50. II The reactor pressure vessel

(RPV)heads of PWRs have penetrations for control rod drive mechanisms and instrumentation systems. Nickel-based alloys ( *e.g.* , Alloy 600) are used in the penetration nozzles and related welds. Primary coolant water and the operating conditions of PWR plants can cause cracking of these nickel-based alloys through a process called primary water stress corrosion cracking (PWSCC). The susceptibility of RPV head penetrations to PWSCC appears to be strongly linked to the operating time and temperature of the RPV head. Problems related to PWSCC have, therefore, increased as plants have operated for longer periods of time. Inspections of the RPV head nozzles at the Oconee Nuclear Station, Units 2 and 3 (Oconee), in early 2001 identified circumferential cracking of the nozzles above the J-groove weld, which joins the nozzle to the RPV head. Circumferential cracking above the J-groove weld is a safety concern because of the possibility of a nozzle ejection if the circumferential cracking is not detected and repaired. Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), which is incorporated into NRC regulations by 10 CFR 50.55a, “Codes and standards,” currently specifies that inspections of the RPV head need only include a visual check for leakage on the insulated surface or surrounding area. These inspections may not detect small amounts of leakage from an RPV head penetration with cracks extending through the nozzle or the J-groove weld. Such leakage can create an environment that leads to circumferential cracks in RPV head penetration nozzles or corrosion of the RPV head. In response to the inspection findings at Oconee and because existing requirements in the ASME Code and NRC regulations do not adequately address inspections of RPV head penetrations for degradation due to PWSCC, the NRC issued Bulletin 2001-01, “Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles,” dated August 3, 2001. In response to the bulletin, PWR licensees provided their plans for inspecting RPV head penetrations and the outside surface of the heads to determine whether any nozzles were leaking. In early March 2002, while conducting inspections of RPV head penetrations prompted by Bulletin 2001-01, the licensee for the Davis-Besse Nuclear Power Station (Davis-Besse) identified a cavity in the RPV head near the top of the dome. The cavity was next to a leaking nozzle with a through-wall axial crack and was in an area of the RPV head that the licensee had left covered with boric acid deposits for several years. On March 18, 2002, the NRC issued Bulletin 2002-01, “Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity,” which requested PWR licensees to provide information on their RPV head inspection and maintenance programs, the material condition of their reactor vessel heads, and their boric acid inspection programs. In their responses, the licensees provided information about their boric acid inspection programs and their inspections and assessments to ensure that their respective plant did not have reactor vessel head degradation like that identified at Davis-Besse. The experience at Davis-Besse and the discovery of leaks and nozzle cracking at other plants reinforced the need for more effective inspections of RPV head penetration nozzles. The absence of an effective inspection regime could, over time, result in unacceptable circumferential cracks in RPV head penetration nozzles or in the degradation of the RPV head by corrosion. These degradation mechanisms increase the probability of a more significant loss of reactor coolant pressure boundary through ejection of a nozzle or other rupture of the RPV head. The NRC issued Bulletin 2002-02, “Reactor Pressure Vessel Head and Vessel Head Penetration Nozzle Inspection Programs,” dated August 9, 2002, requesting that licensees provide information about their inspection programs and any plans to supplement existing visual inspections with additional measures ( *e.g.* , volumetric and surface examinations). Licensees have responded to Bulletin 2002-02 with descriptions of their inspection plans for at least the first refueling outage following the issuance of Bulletin 2002-02 or with a schedule to submit such descriptions before the next refueling outage. Many of the licensees' responses to Bulletin 2002-02 did not describe long-term inspection plans. Instead the licensees stated that they would follow guidance being developed by the industry-sponsored Materials Reliability Program. Inspections performed at several PWR plants in late 2002 found leakage and cracks in nozzles or J-groove welds that have required repairs or prompted the replacement of the RPV head. In addition, as discussed in NRC Information Notice 2003-02, “Recent Experience with Reactor Coolant System Leakage and Boric Acid Corrosion,” issued January 16, 2003, leakage has recently occurred at some plants from connections above the RPV head and has required additional assessments and inspections to ensure that the leakage has not caused significant degradation of RPV heads. The NRC issued an Order Modifying Licenses (Effective Immediately) (EA-03-009), dated February 11, 2003 (Order), to establish required inspections of RPV heads and associated penetration nozzles at PWRs. These requirements were necessary to provide reasonable assurance that plant operations did not pose an undue risk to the public health and safety. The requirements of that Order were expected to remain in effect pending long-term resolution of RPV head penetration inspection requirements, which is expected to involve changes to the NRC regulations, specifically 10 CFR 50.55a. Research being conducted by the NRC and industry is increasing our understanding of material performance, improving inspection capabilities, and supporting assessments of the risks to public health and safety associated with potential degradation of the RPV head and associated penetration nozzles. These research activities are important to the long-term development of revisions to the NRC regulations. III Revising the NRC regulations will take several years. The licensees' actions to date in response to the NRC bulletins and the February 11, 2003, Order have provided reasonable assurance of adequate protection of public health and safety. That Order required inspections of RPV heads and associated penetration nozzles at PWRs which were necessary to provide reasonable assurance that plant operations do not pose an undue risk to the public health and safety. Since the issuance of that Order, the NRC staff has reviewed and granted many requests for relaxation thereof. The arguments in the relaxation requests provide reasonable assurance of the continued structural integrity of the RPV head, and the associated nozzle penetrations and J-groove welds. As a result, it is appropriate to revise that Order with respect to bare metal visual inspections, penetration nozzle inspection coverage, flexibility in combining nondestructive examination

(NDE)methods, flaw evaluation, and requirements for plants which have replaced their reactor pressure vessel head. It is appropriate and necessary to the protection of public health and safety to establish a clear regulatory framework, pending the incorporation of revised inspection requirements into 10 CFR 50.55a. To provide reasonable assurance of adequate protection of public health and safety for the interim period, all PWR Licenses identified in the Attachment to this Order shall be modified to include the inspection requirements for RPV heads and associated penetration nozzles identified in section IV of this Order. The NRC requirements imposed by this Order are based on the body of evidence available through December 2003. Continuing research and operating experience may support future changes to the requirements imposed through this Order. IV Accordingly, pursuant to sections 103, 104b, 161b, 161i, 161o, 182, and 186 of the Atomic Energy Act of 1954, as amended, and the Commission's regulations in 10 CFR 2.202 and 10 CFR part 50, it is hereby ordered that all licenses identified in the Attachment to this Order are modified as follows: A. To determine the required inspection(s) for each refueling outage at their facility, all Licensees shall calculate the susceptibility category of each RPV head to PWSCC-related degradation, as represented by a value of effective degradation years

(EDY)for the end of each operating cycle, using the following equation: EN27FE04.000 Where: *EDY* = total effective degradation years, normalized to a reference temperature of 600 °F *ΔEFPY* j = operating time in years at *T* head,j *Q* i = activation energy for crack initiation (50 kcal/mole) *R* = universal gas constant (1.103x10 − 3 kcal/mole °R) *T* head,j = 100 percent power head temperature during time period *j* (°R = °F + 459.67) *T* ref = reference temperature (600 °F = 1059.67 °R) *n* = number of different head temperatures during plant history This calculation shall be performed with best estimate values for each parameter at the end of each operating cycle for the RPV head that will be in service during the subsequent operating cycle. The calculated value of EDY shall determine the susceptibility category and the appropriate inspection for the RPV head during each refueling outage. B. All Licensees shall use the following criteria to assign the RPV head at their facility to the appropriate PWSCC susceptibility category: High:

(1)Plants with a calculated value of EDY greater than 12, or

(2)Plants with an RPV head that has experienced cracking in a penetration nozzle or J-groove weld due to PWSCC. Moderate: Plants with a calculated value of EDY less than or equal to 12 and greater than or equal to 8 and no previous inspection findings requiring classification as High. Low: Plants with a calculated value of EDY less than 8 and no previous inspection findings requiring classification as High. Replaced: Plants with a replaced RPV head and with a calculated value of EDY less than 8 AND no previous inspection findings requiring classification as High. C. All Licensees shall perform inspections of the RPV head 1 using the following frequencies 2 and techniques: 1 This Order imposes additional inspection requirements. Licensees are required to address any findings from these inspections ( *i.e.* , perform analyses and repairs) in accordance with existing requirements in the ASME Code and 10 CFR 50.55a. The NRC has issued guidance to address flaw evaluations for RPV head penetration nozzles ( *see* letter dated April 11, 2003, from R. Barrett, NRC, to A. Marion, Nuclear Energy Institute, ADAMS Accession No. ML030980322) and will, as necessary, issue revised guidance pending the updating of the NRC regulations. 2 The requirements of this Order are generally consistent with inspection plans that the NRC staff accepted in letters to some Licensees regarding their responses to Bulletin 2002-02. If the NRC staff has already accepted a specific variation from the requirements of this Order ( *e.g.* , inspections to less than 2 inches above the J-groove weld), the Licensee may continue with the previously accepted inspection plan for the first refueling outage after February 11, 2003, provided that in its response to this Order the Licensee identifies all discrepancies between the requirements of this Order and the previously accepted inspection plan. Licensees proposing to deviate from the requirements of this Order for subsequent refueling outages shall seek relaxation of this Order pursuant to the procedure specified at the end of this section.

(1)For those plants in the High category, RPV head and head penetration nozzle inspections shall be performed using the techniques of paragraph IV.C.(5)(a) and paragraph IV.C.(5)(b) every refueling outage. 3 3 For repaired RPV head penetration nozzles that establish a new pressure boundary, the ultrasonic testing inspection shall include the weld and at least 1-inch above the weld in the nozzle base material. For RPV head penetration nozzles or J-groove welds repaired using a weld overlay, the overlay shall be examined by either ultrasonic, eddy current, or dye penetrant testing in addition to the examinations required by paragraph IV.C.(5)(a) and paragraph IV.(C).(5)(b).

(2)For those plants in the Moderate category, RPV head and head penetration inspections shall be performed such that at least the requirements of paragraph IV.C.(5)(a) *or* paragraph IV.C.(5)(b) are performed each refueling outage. In addition the requirements of paragraph IV.C.(5)(a) *and* paragraph IV.C.(5)(b) shall each be performed at least once over the course of every 2 refueling outages.

(3)For those plants in the Low category, RPV head and head penetration nozzle inspections shall be performed as follows. An inspection meeting the requirements of paragraph IV.C.(5)(a) must be completed at least every third refueling outage or every 5 years, whichever occurs first. If an inspection meeting the requirements of paragraph IV.C.(5)(a) was not performed during the last refueling outage prior to February 11, 2003, the Licensee must complete an inspection meeting the requirements of paragraph IV.C.(5)(a) within the first 2 refueling outages after February 11, 2003. The requirements of paragraph IV.C.(5)(b) must be completed at least once prior to February 11, 2008, and thereafter, at least every 4 refueling outages or every 7 years, whichever occurs first.

(4)For those plants in the Replaced category, no RPV head and head penetration nozzle inspections shall be required during the outage for which the RPV head was replaced. Thereafter, until the replacement RPV head in accordance with paragraph IV.A reaches 8 EDY, RPV head and head penetration nozzle inspections shall be performed as follows. An inspection meeting the requirements of paragraph IV.C.(5)(a) must be completed at least every third refueling outage or every 5 years, whichever occurs first. The requirements of paragraph IV.C.(5)(b) must be completed at least every 4 refueling outages or every 7 years, whichever occurs first.

(5)Inspections of the RPV head shall be performed as directed in paragraphs IV.C.(1), IV.C.(2), IV.C.(3) and IV.C.(4) using the following techniques:

(a)Bare metal visual examination of 100 percent of the RPV head surface (including 360° around each RPV head penetration nozzle). For RPV heads with the surface obscured by support structure interferences which are located at RPV head elevations downslope from the outermost RPV head penetration, a bare metal visual inspection of no less than 95 percent of the RPV head surface may be performed provided that the examination shall include those areas of the RPV head upslope and downslope from the support structure interference to identify any evidence of boron or corrosive product. Should any evidence of boron or corrosive product be identified, the licensee shall examine the RPV head surface under the support structure to ensure that the RPV head is not degraded.

(b)For each penetration, perform a nonvisual NDE in accordance with either (i),

(ii)or (iii):

(i)Ultrasonic testing of the RPV head penetration nozzle volume ( *i.e.* , nozzle base material) from 2 inches above the highest point of the root of the J-groove weld (on a horizontal plane perpendicular to the nozzle axis) to 2 inches below the lowest point at the toe of the J-groove weld on a horizontal plane perpendicular to the nozzle axis (or the bottom of the nozzle if less than 2 inches ( *see* Figure IV-1)); or from 2 inches above the highest point of the root of the J-groove weld (on a horizontal plane perpendicular to the nozzle axis) to 1.0-inch below the lowest point at the toe of the J-groove weld (on a horizontal plane perpendicular to the nozzle axis) and including all RPV head penetration nozzle surfaces below the J-groove weld that have an operating stress level (including all residual and normal operation stresses) of 20 ksi tension and greater ( *see* Figure IV-2). In addition, an assessment shall be made to determine if leakage has occurred into the annulus between the RPV head penetration nozzle and the RPV head low-alloy steel.

(ii)Eddy current testing or dye penetrant testing of the entire wetted surface of the J-groove weld and the wetted surface of the RPV head penetration nozzle base material from at least 2 inches above the highest point of the root of the J-groove weld (on a horizontal plane perpendicular to the nozzle axis) to 2 inches below the lowest point at the toe of the J-groove weld on a horizontal plane perpendicular to the nozzle axis (or the bottom of the nozzle if less than 2 inches ( *see* Figure IV-3)); or from 2 inches above the highest point of the root of the J-groove weld (on a horizontal plane perpendicular to the nozzle axis) to 1.0-inch below the lowest point at the toe of the J-groove weld (on a horizontal plane perpendicular to the nozzle axis) and including all RPV head penetration nozzle surfaces below the J-groove weld that have an operating stress level (including all residual and normal operation stresses) of 20 ksi tension and greater ( *see* Figure IV-4).

(iii)A combination of

(i)and

(ii)to cover equivalent volumes, surfaces and leak paths of the RPV head penetration nozzle base material and J-groove weld as described in

(i)and (ii). Substitution of a portion of a volumetric exam on a nozzle with a surface examination may be performed with the following requirements: 1. On nozzle material below the J-groove weld, both the outside diameter and inside diameter surfaces of the nozzle must be examined. 2. On nozzle material above the J-groove weld, surface examination of the inside diameter surface of the nozzle is permitted provided a surface examination of the J-groove weld is also performed. BILLING CODE 7590-01-P EN27FE04.001 EN27FE04.002 EN27FE04.003 EN27FE04.004 BILLING CODE 7590-01-C D. During each refueling outage, visual inspections shall be performed to identify potential boric acid leaks from pressure-retaining components above the RPV head. For any plant with boron deposits on the surface of the RPV head or related insulation, discovered either during the inspections required by this Order or otherwise and regardless of the source of the deposit, before returning the plant to operation the Licensee shall perform inspections of the affected RPV head surface and penetrations appropriate to the conditions found to verify the integrity of the affected area and penetrations. E. For each inspection required in Paragraph C, the Licensee shall submit a report detailing the inspection results within sixty

(60)days after returning the plant to operation. For each inspection required in Paragraph D, the Licensee shall submit a report detailing the inspection results within sixty

(60)days after returning the plant to operation if a leak or boron deposit was found during the inspection. F. In the response required by section V of this Order, all Licensees shall notify the Commission if

(1)they are unable to comply with any of the requirements of section IV or

(2)compliance with any of the requirements of section IV is unnecessary. Licensees proposing to deviate from the requirements of this Order shall seek relaxation of this Order pursuant to the procedure specified below. Project Directors or higher management positions in the Division of Licensing Project Management of the Office of Nuclear Reactor Regulation may, in writing, relax or rescind any of the above conditions upon demonstration by the Licensee of good cause. A request for relaxation regarding inspection of specific nozzles shall also address the following criteria:

(1)The proposed alternative(s) for inspection of specific nozzles will provide an acceptable level of quality and safety, or

(2)Compliance with this Order for specific nozzles would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Requests for relaxation associated with specific penetration nozzles will be evaluated by the NRC staff using its procedure for evaluating proposed alternatives to the ASME Code in accordance with 10 CFR 50.55a(a)(3). V In accordance with 10 CFR 2.202, the Licensee must, and any other person adversely affected by this Order may, submit an answer to this Order, and may request a hearing on this Order, within 20 days of the date of this Order. Where good cause is shown, consideration will be given to extending the time to request a hearing. A request for extension of time in which to submit an answer or request a hearing must be made in writing to the Director, Office of Nuclear Reactor Regulation, Nuclear Regulatory Commission, Washington, DC 20555, and must include a statement of good cause for the extension. The answer may consent to this Order. Unless the answer consents to this Order, the answer shall, in writing and under oath or affirmation, specifically set forth the matters of fact and law on which the Licensee or other person adversely affected relies and the reasons as to why the Order should not have been issued. Any answer or request for a hearing shall be submitted to the Secretary, Office of the Secretary of the Commission, Nuclear Regulatory Commission, Attn: Rulemakings and Adjudications Staff, Washington, DC 20555. Copies shall also be sent to the Director, Office of Nuclear Reactor Regulation, U. S. Nuclear Regulatory Commission, Washington, DC 20555; to the Assistant General Counsel for Materials Litigation and Enforcement at the same address; to the Document Control Desk at the same address; to the Regional Administrator for NRC Region I, II, III, or IV, as appropriate for the specific plant; and to the Licensee if the answer or hearing request is by a person other than the Licensee. Because of possible disruptions in delivery of mail to United States government offices, it is requested that answers and requests for hearing be transmitted to the Secretary of the Commission either by means of facsimile transmission to 301-415-1101 or by e-mail to *hearingdocket@nrc.gov* and also to the Assistant General Counsel for Materials Litigation and Enforcement either by means of facsimile transmission to 301-415-3725 or by e-mail to *OGCMailCenter@nrc.gov.* If a person other than the Licensee requests a hearing, that person shall set forth with particularity the manner in which his interest is adversely affected by this Order and shall address the criteria set forth in 10 CFR 2.714(d). If a hearing is requested by the Licensee or a person whose interest is adversely affected, the Commission will issue an Order designating the time and place of any hearing. If a hearing is held, the issue to be considered at such hearing shall be whether this Order should be sustained. In the absence of any request for a hearing, or written approval of an extension of time in which to request a hearing, the provisions specified in section IV above shall be effective and final 20 days from the date of this Order without further order or proceedings. If an extension of time for requesting a hearing has been approved, the provisions specified in section IV shall be final when the extension expires if a hearing request has not been received. For the Nuclear Regulatory Commission. Dated this 20th day of February, 2004. R. William Borchardt, Acting Director, Office of Nuclear Reactor Regulation. Attachment Facilities Addressee Beaver Valley Power Station, Units 1 and 2, Docket Nos. 50-334 and 50-412, License Nos. DPR-66 and NPF-73 Mr. L. William Pearce, Vice President, FirstEnergy Nuclear Operating Company, Beaver Valley Power Station, Post Office Box 4, Shippingport, PA 15077. Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Docket Nos. 50-317 and 50-318, License Nos. DPR-53 and DPR-69 Mr. George Vanderheyden, Vice President, Calvert Cliffs Nuclear Power Plant, Inc., Calvert Cliffs Nuclear Power Plant, 1650 Calvert Cliffs Parkway, Lusby, MD 20657-4702. R.E. Ginna Nuclear Power Plant, Docket No. 50-244, License No. DPR-18 Dr. Robert C. Mecredy, Vice President, Nuclear Operations, Rochester Gas and Electric Corporation, 89 East Avenue, Rochester, NY 14649. Indian Point Nuclear Generating, Units 2 and 3, Docket Nos. 50-247 and 50-286, License Nos. DPR-26 and DPR-64 Mr. Michael R. Kansler, President, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601. Millstone Power Station, Units 2 and 3, Docket Nos. 50-336 and 50-423, License Nos. DPR-65 and NPF-49 Mr. David A. Christian, Sr. Vice President and Chief Nuclear Officer, Dominion Nuclear Connecticut, Inc., Innsbrook Technical Center, 5000 Dominion Boulevard, Glen Allen, VA 23060-6711. Salem Nuclear Generating Station, Units 1 and 2, Docket Nos. 50-272 and 50-311, License Nos. DPR-70 and DPR-75 Mr. Roy A. Anderson, President & Chief Nuclear Officer, PSEG Nuclear LLC-X04, Post Office Box 236, Hancocks Bridge, NJ 08038. Seabrook Station, Unit 1, Docket No. 50-443, License No. NPF-86 Mr. Mark E. Warner, Site Vice President, c/o James M. Peschel, Seabrook Station, PO Box 300, Seabrook, NH 03874. Three Mile Island Nuclear Station, Unit 1, Docket No. 50-289, License No. DPR-50 Mr. Christopher M. Crane, President and Chief Executive Officer, AmerGen Energy Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. Catawba Nuclear Station, Units 1 and 2, Docket Nos. 50-413 and 50-414, License Nos. NPF-35 and NPF-52 Mr. Dhiaa Jamil, Site Vice President, Catawba Nuclear Station, Duke Energy Corporation, 4800 Concord Road, York, South Carolina 29745-9635. Crystal River Nuclear Power Plant, Docket No. 50-302, License No. DPR-72 Mr. Dale E. Young, Vice President, Crystal River Nuclear Plant (NA1B), Attn: Supervisor, Licensing & Regulatory Programs, 15760 W. Power Line Street, Crystal River, Florida 34428-6708. Joseph M. Farley Nuclear Plant, Units 1 and 2, Docket Nos. 50-348 and 50-364, License Nos. NPF-2 and NPF-8 Mr. L.M. Stinson, Vice President—Farley Project, Southern Nuclear Operating Company, Inc., Post Office Box 1295, Birmingham, Alabama 35201-1295. Shearon Harris Nuclear Power Plant, Unit 1, Docket No. 50-400, License No. NPF-63 Mr. James Scarola, Vice President, Shearon Harris Nuclear Power Plant, Carolina Power & Light Company, Post Office Box 165, Mail Code: Zone 1, New Hill, North Carolina 27562-0165. William B. McGuire Nuclear Station, Units 1 and 2, Docket Nos. 50-369 and 50-370, License Nos. NPF-9 and NPF-17 Mr. G.R. Peterson, Vice President, McGuire Site, Duke Energy Corporation, 12700 Hagers Ferry Road, Huntersville, NC 28078-8985. North Anna Power Station, Units 1 and 2, Docket Nos. 50-338 and 50-339, License Nos. NPF-4 and NPF-7 Surry Power Station, Units 1 and 2, Docket Nos. 50-280 and 50-281, License Nos. DPR-32 and DPR-37 Mr. David A. Christian, Senior Vice President—Nuclear, Virginia Electric and Power Company, 5000 Dominion Blvd., Glen Allen, Virginia 23060. Oconee Nuclear Station, Units 1, 2, and 3, Docket Nos. 50-269, 50-270 and 50-287, License Nos. DPR-38, DPR-47 and DPR-55 Mr. Ronald A. Jones, Vice President, Oconee Site, Duke Energy Corporation, 7800 Rochester Highway, Seneca, SC 29672. H.B. Robinson Steam Electric Plant, Unit 2, Docket No. 50-261, License No. DPR-23 Mr. J.W. Moyer, Vice President, Carolina Power & Light Company, H.B. Robinson Steam Electric Plant, Unit No. 2, 3581 West Entrance Road, Hartsville, South Carolina 29550. St. Lucie Nuclear Plant, Units 1 and 2, Docket Nos. 50-335 and 50-389, License Nos. DPR-67 and NPF-16 Turkey Point Nuclear Generating Station, Units 3 and 4, Docket Nos. 50-250 and 50-251, License Nos. DPR-31 and DPR-41 Mr. J.A. Stall, Senior Vice President, Nuclear and Chief Nuclear Officer, Florida Power and Light Company, P.O. Box 14000, Juno Beach, Florida 33408-0420. Sequoyah Nuclear Plant, Units 1 and 2, Docket Nos. 50-327 and 50-328, License Nos. DPR-77 and DPR-79 Watts Bar Nuclear Plant, Unit 1, Docket No. 50-390, License No. NPF-90 Mr. J.A. Scalice, Chief Nuclear Officer and Executive Vice President, Tennessee Valley Authority, 6A Lookout Place, 1101 Market Street, Chattanooga, Tennessee 37402-2801. Virgil C. Summer Nuclear Station, Unit 1, Docket No. 50-395, License No. NPF-12 Mr. Stephen A. Byrne, Senior Vice President, Nuclear Operations, South Carolina Electric & Gas Company, Virgil C. Summer Nuclear Station, Post Office Box 88, Jenkinsville, South Carolina 29065. Vogtle Electric Generating Plant, Units 1 and 2, Docket Nos. 50-424 and 50-425, License Nos. NPF-68 and NPF-81 Mr. J.T. Gasser, Vice President—Vogtle Project, Southern Nuclear Operating Company, Inc., Post Office Box 1295, Birmingham, Alabama 35201-1295. Brainwood Station, Units 1 and 2, Docket Nos. STN 50-456 and STN 50-457, License Nos. NPF-72 and NPF-77 Byron Station, Units 1 and 2, Docket Nos. STN 50-454 and STN 50-455, License Nos. NPF-37 and NPF-66 Mr. Christopher M. Crane, President, Exelon Nuclear, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. Donald C. Cook Nuclear Plant, Units 1 and 2, Docket Nos. 50-315 and 50-316, License Nos. DPR-58 and DPR-74 Mr. A. Christopher Bakken III, Senior Vice President and Chief Nuclear Officer, Indiana Michigan Power Company, Nuclear Generation Group, 500 Circle Drive, Buchanan, MI 49107. Davis-Besse Nuclear Power Station, Unit 1, Docket No. 50-346, License No. NPF-3 Mr. Lew W. Myers, Chief Operating Officer, FirstEnergy Nuclear Operating Company, Davis-Besse Nuclear Power Station, 5501 North State Route 2, Oak Harbor, OH 43449-9760. Kewaunee Nuclear Power Plant, Docket No. 50-305, License No. DPR-43 Mr. Thomas Coutu, Site Vice President, Kewaunee Nuclear Power Plant, Nuclear Management Company, LLC, N490 State Highway 42, Kewaunee, WI 54216-9511. Palisades Plant, Docket No. 50-255, License No. DPR-20 Mr. Daniel J. Malone, Site Vice President, Palisades Nuclear Plant, 27780 Blue Star Memorial Highway, Covert, MI 49043. Point Beach Nuclear Plant, Units 1 and 2, Docket Nos. 50-266 and 50-301, License Nos. DPR-24 and DPR-27 Mr. Gary Van Middlesworth, Acting Site Vice President, Point Beach Nuclear Plant, Nuclear Management Company, LLC, 6610 Nuclear Road, Two Rivers, WI 54241-9516. Prairie Island Nuclear Generating Plant, Units 1 and 2, Docket Nos. 50-282 and 50-306, License Nos. DPR-42 and DPR-60 Mr. Joseph M. Solymossy, Site Vice President, Prairie Island Nuclear Generating Plant, Nuclear Management Company, LLC, 1717 Wakonade Drive East, Welch, MN 55089. Arkansas Nuclear One, Units 1 and 2, Docket Nos. 50-313 and 50-368, License Nos. DPR-51 and NPF-61 Mr. Jeffrey S. Forbes, Site Vice President, Arkansas Nuclear One, Entergy Operations, Inc., 1448 S.R. 333, Russellville, AR 72801. Callaway Plant, Unit 1, Docket No. 50-483, License No. NPF-30 Mr. Garry L. Randolph, Vice President and Chief Nuclear Officer, Union Electric Company, Post Office Box 620, Fulton, MO 65251. Comanche Peak Steam Electric Station, Units 1 and 2, Docket Nos. 50-445 and 50-446, License Nos. NPF-87 and NPF-89 Mr. Michael R. Blevins, Senior Vice President & Principal Nuclear Officer, TXU Energy, Attn: Regulatory Affairs, P.O. Box 1002, Glen Rose, TX 76043. Diablo Canyon Power Plant, Units 1 and 2, Docket Nos. 50-275 and 50-323, License Nos. DPR-80 and DPR-82 Mr. Gregory M. Rueger, Senior Vice President, Generation and Chief Nuclear Officer, Pacific Gas and Electric Company, Diablo Canyon Power Plant, P.O. Box 3, Avila Beach, CA 93424. Fort Calhoun Station, Unit 1, Docket No. 50-285, License No. DPR-40 Mr. R.T. Ridenoure, Division Manager—Nuclear Operations, Omaha Public Power District, Fort Calhoun Station FC-2-4 Adm., Post Office Box 550, Fort Calhoun, NE 68023-0550. Palo Verde Nuclear Generating Station, Units 1, 2 and 3, Docket Nos. STN 50-528, STN 50-529 and STN 50-530, License Nos. NPF-41, NPF-51 and NPF-74 Mr. Gregg R. Overbeck, Senior Vice President, Nuclear, Arizona Public Service Company, P.O. Box 52034, Phoenix, AZ 80572-2034. San Onofre Nuclear Station, Units 2 and 3, Docket Nos. 50-361 and 50-362, License Nos. NPF-10 and NPF-15 Mr. Harold B. Ray, Executive Vice President, Southern California Edison Company, San Onofre Nuclear Generating Station, P.O. Box 128, San Clemente, CA 92674-0128. South Texas Project Electric Generating, Station, Units 1 and 2, Docket Nos. 50-498 and 50-499, License Nos. NPF-76 and NPF-80 Mr. James J. Sheppard, President and Chief Executive Officer, STP Nuclear Operating Company, South Texas Project Electric Generating Station, P.O. Box 289, Wadsworth, TX 77483. Waterford Steam Electric Generating Station, Unit 3, Docket No. 50-382, License No. NPF-38 Mr. Joseph E. Venable, Vice President Operations, Entergy Operations, Inc., 17265 River Road, Killona, LA 70066-0751. Wolf Creek Generating Station, Unit 1, Docket No. 50-482, License No. NPF-42 Mr. Rick A. Muench, President and Chief Executive Officer, Wolf Creek Nuclear Operating Corporation, Post Office Box 411, Burlington, KS 66839. [FR Doc. 04-4341 Filed 2-26-04; 8:45 am]

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