Rules and Regulations. NUCLEAR REGULATORY COMMISSION

33,172 words·~151 min read·/register/2006/01/17/06-320·

A research copy — for the controlling text, always check the official state or federal source. Not legal advice.

BILLING CODE 7555-01-M NUCLEAR REGULATORY COMMISSION [IA-05-052] David Geisen; Order Prohibiting Involvement in NRC-Licensed Activities (Effective Immediately) I Mr. David Geisen was previously employed, at times relevant to this Order, as the Manager of Design Engineering at the Davis-Besse Nuclear Power Station (Davis-Besse) operated by FirstEnergy Nuclear Operating Company (FENOC or licensee). The licensee holds License No. NPF-3 which was issued by the Nuclear Regulatory Commission (NRC or Commission) pursuant to 10 CFR part 50 on April 22, 1977.

The license authorizes the operation of Davis-Besse in accordance with the conditions specified therein. The facility is located on the licensee's site near Oak Harbor, Ohio. II On August 3, 2001, the NRC issued Bulletin 2001-001, “Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles,” (Bulletin). In the Bulletin, the NRC requested that all holders of operating licenses for pressurized water nuclear power reactors (PWR), including FENOC for the Davis-Besse facility, provide information to the NRC relating to the structural integrity of the reactor pressure vessel

(RPV)head penetration nozzles at their respective facilities. The information requested from the licensees included the extent of RPV head penetration nozzle leakage and cracking that had been found to date, a description of the inspections and repairs undertaken to satisfy applicable regulatory requirements, and the basis for concluding that a licensee's plans for future inspections would ensure compliance with applicable regulatory requirements. The NRC also required that all Bulletin addressees, including FENOC, submit a written response to the NRC in accordance with the provisions of 10 CFR 50.54(f). That regulation provides, in part, that upon request of the NRC, an NRC-licensee must submit written statements, signed under oath or affirmation, to enable the NRC to determine whether the license should be modified, suspended, or revoked. On September 4, October 17, and October 30, 2001, the licensee provided written responses to the Bulletin. Additionally, the licensee met with the NRC staff on numerous occasions during October and November of 2001 to provide clarifying information. Based, in part, on the information provided by FENOC in its written responses to the Bulletin and during meetings with the NRC staff, the NRC staff allowed the licensee to continue operation of the Davis-Besse facility until February 2002, rather than requiring FENOC to shut the unit down to perform inspections by December 31, 2001, as provided in the Bulletin. On February 16, 2002, FENOC shut down Davis-Besse for refueling and inspection of control rod drive mechanism

(CRDM)RPV head penetration nozzles. Using ultrasonic testing, the licensee found cracks in three CRDM RPV head penetration nozzles and on March 6, 2002, the licensee discovered a cavity in the RPV head in the vicinity of CRDM Penetration Nozzle No. 3. The cavity measured approximately 5 to 7 inches long, 4 to 5 inches wide, and penetrated through the 6.63 inch-thick low-alloy steel portion of the RPV head, leaving the stainless steel cladding material (measuring 0.202 to 0.314 inches-thick) as the sole reactor coolant system

(RCS)pressure boundary. A smaller cavity was also found near CRDM Penetration Nozzle No. 2. The licensee conducted a root cause evaluation and determined that, contrary to the earlier information provided to the NRC, the cavities were caused by boric acid from the RCS released through cracks in the CRDM RPV head penetration nozzles. The root cause evaluation found that the licensee conducted limited cleaning and inspections of the RPV head during the Twelfth Refueling Outage (12RFO) that ended on May 18, 2000. However, neither the limited RPV head cleaning nor the resultant inspections during 12RFO were sufficient to ensure that the significant boric acid deposits on the RPV head were only a result of CRDM flange leakage, as supposed, and were not a result of RCS pressure boundary leakage. On March 6 and March 10, 2002, the licensee provided information to the NRC concerning the identification of a large cavity in the RPV head adjacent to CRDM Penetration Nozzle No. 3. The NRC conducted an Augmented Inspection Team

(AIT)inspection at Davis-Besse from March 12 to April 5, 2002, to determine the facts and circumstances related to the significant degradation of the RPV head. The results of the AIT inspection were documented in NRC Inspection Report No. 50-346/2002-03, issued on May 3, 2002. A follow-up Special Inspection was conducted from May 15 to August 9, 2002, and on October 2, 2002, the NRC issued the AIT Follow-up Special Inspection Report No. 50-346/2002-08 documenting ten apparent violations associated with the RPV head degradation. On April 22, 2002, the NRC Office of Investigations

(CR)2000-1037 which included the following problem statement associated with the identification of five leaking control rod drives: “Identified at locations: F10, D10, C11, F8, and G9 * * * There are no boron deposits on the vertical faces of the flange of G9 drive. The bottom of the flange of G9 drive is inaccessible for inspection due to the boron buildup on the reactor head insulation, not allowing full camera insertion. Since the boron is evident only under the flange and not on the vertical surfaces, there is a high probability that G9 is a leaking CRD.” • On June 27, 2001, Mr. Giesen approved and signed an intra-company memorandum that indicated that “large boron leakage from a control rod drive mechanism

(CRDM)flange was observed during 12RFO inspection” and “This leakage did not permit the detailed inspection of CRDM nozzles.” • On August 11, 2001, Mr. Geisen received an E-mail that stated, in part: “it was pointed out that we cannot clean our head thru the mouse holes and a system engineer is requesting that three large holes be cut in the Service Structure for viewing [inspection] and cleaning.” • Mr. Geisen reviewed a Piedmont Management and Technical Services, Inc., report, dated September 14, 2001, that indicated, in part, that at the completion of 12RFO the RPV head had boric acid deposits of considerable depth left at the center top area of the head. • A Senior Staff Nuclear Advisor (former inservice inspector), FENOC, at the request of a system engineer from Davis-Besse plant engineering, reviewed a CD ROM video that the system engineer had made from videos of the reactor vessel head. The purpose of the review was to assist in locating or determining the location of some nozzles. Shortly after completing the review, Mr. Geisen asked the Senior Staff Nuclear Advisor what he thought, from a visual standpoint, of the data he had seen on the video. The Senior Staff Nuclear Advisor replied, in part, that, based on an Electric Power Research Institute

(EPRI)head examination document being developed, boron on the Davis-Besse head would preclude an examination of that nature [EPRI] from being performed. • In March 2002, a consultant from Martin Sigmund Consulting Services, Inc., conducted an assessment of reactor head management issues at Davis-Besse. The consultant provided his assessment to the Davis-Besse Site Vice President via a memorandum dated March 28, 2002. The assessment, in part, consisted of interviews with many of the personnel involved with the reactor head corrosion issues. Mr. Geisen was interviewed for this assessment on March 27, 2002, and stated, in part, that some boric acid was left on the head in 2000 and that the condition report was not very thoroughly evaluated. Mr. Geisen also stated that he became aware that the reactor vessel head had not been cleaned completely when reviewing the videos of the inspections in preparation for interacting with the NRC in August, 2001. • On June 18, 2002, the licensee interviewed Mr. Geisen regarding the Davis-Besse responses to Bulletin 2001-001. When asked whether the reactor vessel head was inspected in accordance with plant procedure, Mr. Geisen stated, in part, that we did the inspection but clearly not with [in accordance with] the procedure. Mr. Geisen further stated that Davis-Besse was taking credit for a general inspection which clearly did not meet the requirements in Bulletin 2001-001. The above information demonstrates that Mr. Geisen had sufficient knowledge of the results of previous inspections of the RPV head and that he knew that the licensee's written and oral responses to NRC Bulletin 2001-001 were incomplete and inaccurate. Several FENOC employees, including Mr. David Geisen, were responsible for the information provided to the NRC by FENOC in response to the Bulletin. III David Geisen was employed by FENOC as the Manager of Design Engineering at Davis-Besse at the time the licensee developed and transmitted to the NRC its written responses to the Bulletin and at the time the licensee met with the NRC to provide clarifying information regarding its written responses. On August 28, October 17, and October 30, 2001, respectively, Mr. Geisen concurred in the issuance of the licensee's September 4, October 17, and October 30, 2001, responses to the Bulletin. On the concurrence sheets, Mr. Geisen was listed as the FENOC manager responsible for ensuring the completeness and accuracy of the responses. Mr. Geisen participated in the development and presentation of information to the NRC during information briefings held on October 3, October 11, and November 9, 2001. Item 1.d of the Bulletin requested each pressurized water reactor

(RFO)in April 1998 and during the 12th RFO in April 2000. The scope of the visual inspection was to inspect the bare metal RPV head area that was accessible through the weep holes to identify any boric acid leaks/deposits. The DBNPS also inspected 100% of Control Rod Drive Mechanism

(CRDM)flanges for leaks in response to Generic Letter 88-05, ‘Boric Acid Corrosion of Carbon Steel Reactor pressure Boundary Components in PWR Plants.’ The results of these two recent inspections are described below. Inspections of the RPV head are performed with the RPV head insulation installed in accordance with DBNPS procedure NG-EN-0324, ‘Boric Acid Corrosion Control Program,’ which was developed in response to Generic Letter 88-05. As stated previously, a gap exists between the RPV head and the insulation, the minimum gap being at the dome center of the RPV head where it is approximately 2 inches, and does not impede visual inspection. The service structure envelopes the DBNPS RPV head and has 18 openings (weep holes) at the bottom through which inspections are performed. There are 69 CRDM nozzles that penetrate the RPV head. The metal reflective insulation is located above the head and does not interfere with the visual inspection. The visual inspection is performed by the use of a small camera. This camera is inserted through the weep holes.” Item 1.d of the licensee's September 4, 2001, response, under the section entitled, “April 2000 Inspection Results (12RFO),” stated: “The boric acid deposits were located beneath the leaking flanges with clear evidence of downward flow. No visible evidence of nozzle leakage was detected.” Item 1.d of the licensee's September 4, 2001, response, under the section entitled, “Subsequent Review of 1998 and 2000 Inspection Videotapes Results,” stated: “Since May 2001, a review of the 1998 and 2000 inspection videotapes of the RPV head has been performed. This review was conducted to re-confirm the indications of boron leakage experienced at the DBNPS were not similar to the indications seen at ONS and ANO-1; i.e., was not indicative of RPV nozzle leakage. This review determined that indications such as those that would result from RPV head penetration leakage were not evident.” The licensee's September 4, 2001, response was materially incomplete and inaccurate in that the response:

(1)Mischaracterized the accumulation of boric acid on the RVP head as a result of the 12RFO RPV head inspection;

(2)failed to include information that during the Eleventh Refueling Outage (11RFO) and 12RFO, the licensee's access to the RPV head bare metal was impeded by the presence of significant accumulations of boric acid deposits;

(3)failed to indicate that the presence of boric acid deposits was not limited to the area beneath control rod drive mechanism flanges; and

(4)failed to indicate that the build-up of boric acid deposits was so significant that the licensee could not inspect all of the RPV head penetration nozzles. Mr. Geisen was aware that the licensee's September 4, 2001, response to the Bulletin was materially incomplete and inaccurate, but nevertheless concurred on the response, thereby allowing it to be submitted to the NRC. The NRC staff determined that the September 4, 2001 response did not include sufficient information to justify the NRC permitting FENOC to operate Davis-Besse beyond December 31, 2001. As a result, FENOC met with the NRC staff, Commissioners' Technical Assistants, the Advisory Committee on Reactor Safeguards, and Congressional staff members, and developed supplemental responses in an effort to better communicate its justification for continued operations beyond December 31, 2001. On October 3, 2001, Mr. Geisen participated in a conference call with the NRC staff. Mr. Geisen was also involved in preparatory meetings for the October 3rd conference call. The agenda for the conference call stated “Video Inspection Review from RFO10, RFO11, and RFO12: Further Confirmation of no indication of leakage attributable to CRDM nozzle leakage; clearly CRDM flange leakage.” During the conference call, Mr. Geisen informed the NRC that 100% of the reactor pressure vessel head had been inspected during the last outage (RFO12) but some areas were precluded from inspection and that videotapes of the 10RFO, 11RFO, and 12RFO reactor pressure vessel head inspections had been reviewed. The information communicated by the Mr. Geisen during the conference call was materially incomplete and inaccurate in that the licensee did not conduct a 100% inspection of the RPV head during 12RFO due to the presence of significant amount of boric acid on the reactor pressure vessel head which obscured a significant number of RPV head nozzles. On October 10, 2001, Mr. Geisen attended a meeting with other FENOC management officials for the purposes of finalizing presentation slides for an October 11, 2001, meeting with the NRC Commissioner's Technical Assistants. Draft Presentation Slide 20 stated: “Reviewed video inspections of Reactor Vessel head taken during 11RFO (April 1998) and 12RFO (April 2000) and confirmed that Davis-Besse has not experienced boron leakage as seen at Oconee or Arkansas Nuclear.” Presentation Draft Slide 21 stated: “Reviewed past 3 outages of Reactor Vessel Head inspection video tapes which were taken to satisfy Generic Letter 97-01: No telltale “popcorn” type boron deposits; During 12RFO (Spring 2000), Davis-Besse identified sources of boron that precluded the visual inspection of some CRDM penetrations, as five leaking flanges above the mirror insulation; Viewed past 3 outages of inspection video tapes of area masked by boron in 12 RFO did not have previous leakage.” On October 11, 2001, Mr. Geisen and other licensee staff briefed the NRC Commissioners' Technical Assistants as to FENOC's basis for determining that Davis-Besse was safe to operate until the next refueling outage (March 2002). During the briefing, FENOC and Mr. Geisen, as a presenter, discussed the presentation slides that were finalized the previous day. Presentation Slide 6, as presented by FENOC stated, in part: “Conducted and recorded video inspections of the head during 11RFO (April 1998) and 12RFO (April 2000)—No head penetration leakage was identified.” Presentation Slide 7, as presented by Mr. Geisen stated, in part: “All CRDM [control rod drive mechanism] penetrations were verified to be free from “popcorn” type boron deposits using video recordings from 11RFO or 12RFO.” The licensee's October 11, 2001, presentation to the NRC Commissioners' Technical Assistants was materially incomplete and inaccurate in that the presentation slides did not state that the build-up of boric acid on the RPV head was so significant that the licensee could not inspect all of the RPV head penetration nozzles. Due to the significant amount of boric acid present on the RPV head, of which he was aware, Mr. Geisen did not have a basis for stating that no visible evidence of RPV penetration nozzle leakage was detected. On October 17, 2001, the licensee provided a supplemental response to the Bulletin. The second paragraph under the section entitled, “Previous Inspection Results,” on Page 2 of Attachment 1 of the licensee's October 17, 2001, supplemental response stated, in part: “The inspections performed during the 10th, 11th, and 12th Refueling Outage (10RFO, conducted April 8 to June 2, 1996; 11RFO, conducted April 10 to May 23, 1998; and, 12RFO, conducted April 1 to May 18, 2000) consisted of a whole head visual inspection of the RPV head in accordance with the DBNPS Boric Acid Control Program pursuant to Generic Letter 88-05 ‘Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants.’ The visual inspections were conducted by remote camera and included below insulation inspections of the RPV bare head such that the Control Rod Drive Mechanism

(1998)and 12RFO

(2000)inspections were focused on inspecting the RPV for indications of the impact of boric acid leakage from leaking flanges. Mr. Geisen stated that the 1998 and 2000 inspections (video tapes) did not give a good view of the control rod drives because the camera angle was looking upwards at the structural material of the service structure on top of the head. Mr. Geisen stated that the video tape of the 10RFO

(1996)inspection was a better video because the camera was following around a vacuum and probe that were specifically looking for head wastage as a result of boron deposits on the head. The information provided by the licensee and Mr. Geisen to the ACRS was materially incomplete and inaccurate in that each of the video tapes was helpful in understanding the significant boron accumulations present at the start of each outage, the clear impediments to 100% inspection of the RPV head nozzles, and difficulty the licensee encountered in its attempts to fully clean the RPV head of boron or to complete a comprehensive inspection of the RPV head nozzles. Following the 1996 RPV head inspection, the licensee generated Potential Condition Adverse to Quality Report 96-0551, which stated, in part, on Continuation Sheet Page 9, Part C, Item 1: “The extent of the inspection was limited to approximately 50 to 60% of the head area because of the restrictions imposed by the location and size of mouseholes. The inspection showed varying sizes of boric acid mounds scattered in various areas of the head. It is extremely difficult to develop an estimate of the amount of boric acid deposit because of the deposit scatter and limited inspection.” Based on the above information, the NRC concludes that Mr. Geisen had knowledge of the RPV head conditions and the limitations experienced during RPV head inspections, and that, notwithstanding that knowledge, he deliberately provided materially incomplete and inaccurate information when he:

(1)Concurred, on August 28, October 17, and October 30, 2001, respectively, in the licensee's September 4, October 17, and October 30, 2001, responses to the Bulletin; and

(2)assisted in the preparation and presentation of incomplete or inaccurate information during internal meetings on October 2 and 10, 2001, and during meetings or teleconferences held with the NRC on October 3, 11, and November 9, 2001. The information provided by the licensee under oath in the Bulletin responses based, in part on the concurrence of Mr. Geisen, was material to the NRC because the NRC used the information, in part, to allow FENOC to operate Davis-Besse until February 2002 rather than requiring the plant to shut down by December 31, 2001, to conduct inspections of the head as discussed in Item 3.v.1. of the Bulletin. The information provided to the NRC during teleconferences and meetings was material to the NRC because the information gave the impression to the NRC staff that the Davis-Besse RPV head had been completely inspected and that the licensee had not identified any indications of RPV head penetration nozzle cracks when this was not the case at the time the response was submitted. Based on the above information, Mr. David Geisen, while employed by the licensee, engaged in deliberate misconduct by deliberately providing FENOC and the NRC information that he knew was not complete or accurate in all material respects to the NRC, a violation of 10 CFR 50.5(a)(2). Mr. Geisen's actions also placed FENOC in violation of 10 CFR 50.9. The NRC determined that these violations were of very high safety and regulatory significance because they demonstrated a pattern of deliberate inaccurate or incomplete documentation of information that was required to be submitted to the NRC. Had the NRC been aware of this incomplete and inaccurate information, the NRC would likely have taken immediate regulatory action to shut down the plant and require the licensee to implement appropriate corrective actions. The NRC must be able to rely on the licensee and its employees to comply with NRC requirements, including the requirement to provide information that is complete and accurate in all material respects. Mr. Geisen's action violated 10 CFR 50.5(a)(2) and caused the licensee to violate 10 CFR 50.9, and raise serious doubt as to whether he can be relied upon to comply with NRC requirements and to provide complete and accurate information to the NRC. Consequently, I lack the requisite reasonable assurance that licensed activities can be conducted in compliance with the Commission's requirements and that the health and safety of the public will be protected if Mr. Geisen is permitted to be involved in NRC-licensed activities. Therefore, the public health, safety and interest require that Mr. Geisen be prohibited from any involvement in NRC-licensed activities for a period of five years from the effective date of this Order. Additionally, Mr. Geisen is required to notify the NRC of his first employment in NRC-licensed activities for a period of five years following the prohibition period. V Accordingly, pursuant to sections 103, 104, 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, 10 CFR 50.5, and 10 CFR 150.20, *It is hereby ordered* that effective immediately: 1. Mr. David Geisen is prohibited for five years from the date of this Order from engaging in NRC-licensed activities. The NRC considers NRC-licensed activities to be those activities that are conducted pursuant to a specific or general license issued by the NRC, including those activities of Agreement State licensees conducted pursuant to the authority granted by 10 CFR 150.20. 2. If Mr. Geisen is currently involved with another licensee in NRC-licensed activities, he must immediately cease those activities, and inform the NRC of the name, address and telephone number of the employer, and provide a copy of this Order to the employer. 3. For a period of five years after the five-year period of prohibition has expired, Mr. Geisen shall, within 20 days of acceptance of his first employment offer involving NRC-licensed activities or his becoming involved in NRC-licensed activities, as defined in Paragraph IV.1 above, provide notice to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555, of the name, address, and telephone number of the employer or the entity where he is, or will be, involved in NRC-licensed activities. In the notification, Mr. Geisen shall include a statement of his commitment to compliance with regulatory requirements and the basis why the Commission should have confidence that he will now comply with applicable NRC requirements. The Director, Office of Enforcement, may, in writing, relax or rescind any of the above conditions upon demonstration by Mr. Geisen of good cause. VI In accordance with 10 CFR 2.202, David Geisen 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. However, since this enforcement action is being proposed prior to the U.S. Department of Justice completing its review of the OI investigation results, consideration may be given to extending the response time for submitting an answer as well as the time for requesting a hearing, for good cause shown. A request for extension of time must be made in writing to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 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 admit or deny each allegation or charge made in this Order and shall set forth the matters of fact and law on which Mr. Geisen or other person adversely affected relies and the reasons as to why the Order should not have been issued. Pursuant to 10 CFR 2.202(c)(2)(i), Mr. Giesen, may, in addition to demanding a hearing, at the time the answer is filed or sooner, move the presiding officer to set aside the immediate effectiveness of the Order on the ground that the Order, including the need for immediate effectiveness, is not based on adequate evidence but on mere suspicion, unfounded allegations, or error. Any answer or request for a hearing shall be submitted to the Secretary, U.S. Nuclear Regulatory Commission, Attn: Rulemakings and Adjudications Staff, Washington, DC 20555. Copies also shall be sent to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555, to the Assistant General Counsel for Materials Litigation and Enforcement at the same address, to the Regional Administrator, NRC Region III, 2443 Warrenville Road, Lisle, IL 60532-4352, and to Mr. Geisen if the answer or hearing request is by a person other than Mr. Geisen. Because of continuing 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 Office of the General Counsel either by means of facsimile transmission to 301-415-3725 or by e-mail to *OGCMailCenter@nrc.gov* . If a person other than Mr. Geisen 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 2.309. If a hearing is requested by Mr. Geisen 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. Pursuant to 10 CFR 2.202(c)(2)(i), Mr. Goyal, may, in addition to demanding a hearing, at the time the answer is filed or sooner, move the presiding officer to set aside the immediate effectiveness of the Order on the ground that the Order, including the need for immediate effectiveness, is not based on adequate evidence but on mere suspicion, unfounded allegations, or error. In the absence of any request for hearing, or written approval of an extension of time in which to request a hearing, the provisions specified in Section V above shall be effective immediately and shall be 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 V shall be final when the extension expires if a hearing request has not been received. Dated this 4th day of January 2006. For the Nuclear Regulatory Commission. Martin J. Virgilio, Deputy Executive Director for Materials, Research, State and Compliance Programs, Office of the Executive Director for Operations. [FR Doc. E6-437 Filed 1-13-06; 8:45 am] BILLING CODE 7590-01-P NUCLEAR REGULATORY COMMISSION [ IA-05-055] Prasoon Goyal; Order Prohibiting Involvement in NRC-Licensed Activities (Effective Immediately) I Mr. Prasoon Goyal was previously employed, at times relevant to this Order, as a Senior Engineer at the Davis-Besse Nuclear Power Station (Davis-Besse) operated by FirstEnergy Nuclear Operating Company (FENOC or licensee). The licensee holds License No. NPF-3 which was issued by the Nuclear Regulatory Commission (NRC or Commission) pursuant to 10 CFR part 50 on April 22, 1977. The license authorizes the operation of Davis-Besse in accordance with the conditions specified therein. The facility is located on the licensee's site near Oak Harbor, Ohio. II On August 3, 2001, the NRC issued Bulletin 2001-001, “Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles,” (Bulletin). In the Bulletin, the NRC requested that all holders of operating licenses for pressurized water nuclear power reactors (PWR), including FENOC for the Davis-Besse facility, provide information to the NRC relating to the structural integrity of the reactor pressure vessel

(RCS)pressure boundary. A smaller cavity was also found near CRDM Penetration Nozzle No. 2. The licensee conducted a root cause evaluation and determined, contrary to the earlier information provided to the NRC, that the cavities were caused by boric acid from the RCS released through cracks in the CRDM RPV head penetration nozzles. The root cause evaluation found that the licensee conducted limited cleaning and inspections of the RPV head during the Twelfth Refueling Outage (12RFO) that ended on May 18, 2000. However, neither the limited RPV head cleaning nor the resultant inspections during 12RFO were sufficient to ensure that the significant boric acid deposits on the RPV head were only a result of CRDM flange leakage, as supposed, and were not a result of RCS pressure boundary leakage. On March 6 and March 10, 2002, the licensee provided information to the NRC concerning the identification of a large cavity in the RPV head adjacent to CRDM Penetration Nozzle No. 3. The NRC conducted an Augmented Inspection Team

(PWR)licensee, including FENOC for Davis-Besse, to provide a description of the RPV head penetration nozzles and RPV head inspection (including type, scope, qualification requirements, and acceptance criteria) that were performed at PWRs in the 4 years preceding the date of the Bulletin, and the findings resulting from the inspections. The licensees were requested to include a description of any limitations (insulation or other impediments) to accessibility of the bare metal of the RPV head for visual examinations. On September 4, 2001, FENOC submitted its written response to the Bulletin for Davis-Besse. On October 17, 2001, FENOC submitted a supplemental response to the Bulletin for Davis-Besse and included information not provided in the September 4, 2001, response with regard to RPV inspections and cleaning conducted during 10RFO. Attachment 1 to the licensee's October 17, 2001, supplemental response to the Bulletin stated under the section entitled, “Summary,” in part: “In May 1996, during a refueling outage, the RPV head was inspected. No leakage was identified, and these results have been recently verified by a re-review of the video tapes obtained from that inspection.” The October 17, 2001, supplemental response to the Bulletin also stated under the section entitled, “Previous Inspection Results,” in part: “The inspections performed during the 10th, 11th, and 12th Refueling Outage (10RFO, conducted April 8 to June 2, 1996; 11RFO, conducted April 10, to May 23, 1998; and, 12RFO, conducted April 1 to May 28, 2000) consisted of a whole head visual inspection of the RPV head in accordance with the DBNPS Boric Acid Control Program pursuant to Generic Letter 88-05, ‘Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants.' The visual inspections were conducted by remote camera and included below insulation inspections of the RPV bare head such that the Control Rod Drive Mechanism

(CRDM)nozzle penetrations were viewed. During 10RFO, 65 of 69 nozzles were viewed, during 11RFO, 50 of 69 nozzles were viewed, and during 12 RFO, 45 of 69 nozzles were viewed.” Information included under Column 6 of Attachment 2 of the licensee's October 17, 2001, supplemental response stated, in part, that 24 nozzles have a “flange leak evident.” Note 1 on the same table stated, in part: “In 1996 during 10 RFO, the entire RPV head was inspected. Since the video was void of head orientation narration, each specific nozzle view could not be correlated.” The licensee's October 17, 2001, supplemental response was materially incomplete and inaccurate in that the licensee did not view the stated number of RPV head penetration nozzles during the referenced outages, and the licensee believed that only five RPV head control rod drive mechanism flanges were leaking instead of the 24 RPV head control rod drive mechanism flanges noted in the response. Mr. Goyal was aware that the licensee's October 17, 2001, supplemental response was materially incomplete and inaccurate and concurred on the response, thereby allowing it to be submitted to the NRC. Based on the above information, the NRC concludes that Mr. Goyal had sufficient knowledge of the condition of the RPV head and the limitations experienced during the RPV head inspections conducted during 10RFO, and notwithstanding that knowledge, he deliberately provided materially incomplete and inaccurate information, when on October 17, 2001, he concurred on the licensee's October 17, 2001, supplemental response to the NRC. The information provided by the licensee under oath in the Bulletin supplemental response was material to the NRC because the NRC used the information, in part, to allow FENOC to operate Davis-Besse until February 2002 rather than requiring the plant to shut down by December 31, 2001, to conduct inspections of the head as discussed in Item 3.v.1. of the Bulletin. Based on the above information, Mr. Prasoon Goyal, while employed by the licensee, engaged in deliberate misconduct by deliberately providing incomplete or inaccurate information that he knew was not complete and accurate in all material respects to the NRC, a violation of 10 CFR 50.5(a)(2). Mr. Goyal's actions also placed FENOC in violation of 10 CFR 50.9. The NRC determined that these violations were of very high safety and regulatory significance because they involved a pattern of deliberate documentation of inaccurate or incomplete information that was required to be submitted to the NRC. Had the NRC been aware of this incomplete and inaccurate information, the NRC would likely have taken immediate regulatory action to shut down the plant and require the licensee to implement appropriate corrective actions. IV The NRC must be able to rely on the licensee and its employees to comply with NRC requirements, including the requirement to provide information and maintain records that are complete and accurate in all material respects. Mr. Goyal's deliberate actions raise serious doubt as to whether he can be relied upon to comply with NRC requirements and to provide complete and accurate information to the NRC. Consequently, I lack the requisite reasonable assurance that licensed activities can be conducted in compliance with the Commission's requirements and that the health and safety of the public will be protected if Mr. Goyal is permitted to be involved in NRC-licensed activities. Therefore, the public health, safety and interest require that Mr. Goyal be prohibited from any involvement in NRC-licensed activities for a period of one year effective immediately. Additionally, Mr. Goyal is required to notify the NRC of his first employment in NRC-licensed activities for a period of one year following the prohibition period. V Accordingly, pursuant to sections 103, 104, 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, 10 CFR 50.5, and 10 CFR 150.20, *It is hereby ordered* that effective immediately: 1. Mr. Prasoon Goyal is prohibited for one year from the date of this Order from engaging in NRC-licensed activities. The NRC considers NRC-licensed activities to be those activities that are conducted pursuant to a specific or general license issued by the NRC, including those activities of Agreement State licensees conducted pursuant to the authority granted by 10 CFR 150.20. 2. If Mr. Goyal is currently involved with another licensee in NRC-licensed activities, he must immediately cease those activities, and inform the NRC of the name, address and telephone number of the employer, and provide a copy of this Order to the employer. 3. For a period of one year after the one-year period of prohibition has expired, Mr. Goyal shall, within 20 days of acceptance of his first employment offer involving NRC-licensed activities or his becoming involved in NRC-licensed activities, as defined in Paragraph IV.1 above, provide notice to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555, of the name, address, and telephone number of the employer or the entity where he is, or will be, involved in NRC-licensed activities. In the notification, Mr. Goyal shall include a statement of his commitment to compliance with regulatory requirements and the basis why the Commission should have confidence that he will now comply with applicable NRC requirements. The Director, Office of Enforcement, may, in writing, relax or rescind any of the above conditions upon demonstration by Mr. Goyal of good cause. VI In accordance with 10 CFR 2.202, Prasoon Goyal 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, consideration may be given to extending the response time for submitting an answer as well as the time for requesting a hearing, for good cause shown. A request for extension of time must be made in writing to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 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 admit or deny each allegation or charge made in this Order and shall set forth the matters of fact and law on which Mr. Goyal 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, U.S. Nuclear Regulatory Commission, Attn: Rulemakings and Adjudications Staff, Washington, DC 20555. Copies also shall be sent to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555, to the Assistant General Counsel for Materials Litigation and Enforcement at the same address, to the Regional Administrator, NRC Region III, 2443 Warrenville Road, Lisle, IL 60532-4352, and to Mr. Goyal if the answer or hearing request is by a person other than Mr. Goyal. Because of continuing 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 Office of the General Counsel either by means of facsimile transmission to 301-415-3725 or by e-mail to *OGCMailCenter@nrc.gov.* If a person other than the Mr. Goyal 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 2.309. If a hearing is requested by Mr. Goyal 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. Pursuant to 10 CFR 2.202(c)(2)(i), Mr. Goyal, may, in addition to demanding a hearing, at the time the answer is filed or sooner, move the presiding officer to set aside the immediate effectiveness of the Order on the ground that the Order, including the need for immediate effectiveness, is not based on adequate evidence but on mere suspicion, unfounded allegations, or error. In the absence of any request for hearing, or written approval of an extension of time in which to request a hearing, the provisions specified in Section V above shall be effective immediately 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 V shall be final when the extension expires if a hearing request has not been received. Dated this 4th day of January 2006. For the Nuclear Regulatory Commission. Martin J. Virgilio, Deputy Executive Director for Materials, Research, State, and Compliance Programs, Office of the Executive Director for Operations. [FR Doc. E6-418 Filed 1-13-06; 8:45 am] BILLING CODE 7590-01-P NUCLEAR REGULATORY COMMISSION [IA-05-053] Dale Miller; Order Prohibiting Involvement in NRC-Licensed Activities (Effective Immediately) I Mr. Dale Miller was previously employed, at times relevant to this Order, as a Compliance Supervisor at the Davis-Besse Nuclear Power Station (Davis-Besse) operated by FirstEnergy Nuclear Operating Company (FENOC or licensee). The licensee holds License No. NPF-3 which was issued by the Nuclear Regulatory Commission (NRC or Commission) pursuant to 10 CFR Part 50 on April 22, 1977. The license authorizes the operation of Davis-Besse in accordance with the conditions specified therein. The facility is located on the licensee's site near Oak Harbor, Ohio. II On August 3, 2001, the NRC issued Bulletin 2001-001, “Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles,” (Bulletin). In the Bulletin, the NRC requested that all holders of operating licenses for pressurized water nuclear power reactors (PWR), including FENOC for the Davis-Besse facility, provide information to the NRC relating to the structural integrity of the reactor pressure vessel

(RPV)head penetration nozzles at their respective facilities. The information requested from the licensees included the extent of RPV head penetration nozzle leakage and cracking that had been found to date, a description of the inspections and repairs undertaken to satisfy applicable regulatory requirements, and the basis for concluding that a licensee's plans for future inspections would ensure compliance with applicable regulatory requirements. The NRC also required that all the Bulletin addressees, including FENOC, submit a written response to the NRC in accordance with the provisions of 10 CFR 50.54(f). That regulation provides, in part, that upon request of the NRC, an NRC-licensee must submit written statements, signed under oath or affirmation, to enable the NRC to determine whether the license should be modified, suspended, or revoked. On September 4, October 17, and October 30, 2001, the licensee provided written responses to the Bulletin. Additionally, the licensee met with the NRC staff on numerous occasions during October and November of 2001 to provide clarifying information. Based, in part, on the information provided by FENOC in the written responses to the Bulletin and during meetings with the NRC staff, the NRC staff allowed the licensee to continue operation of the Davis-Besse facility until February 2002, rather than requiring FENOC to shut the unit down to perform inspections by December 31, 2001, as provided in the Bulletin. On February 16, 2002, FENOC shut down Davis-Besse for refueling and inspection of control rod drive mechanism

(OI)initiated an investigation at Davis-Besse to determine, among other matters, whether FENOC and individual employees at the Davis-Besse facility failed to provide complete and accurate information to the NRC in its September 4, October 17, and October 30, 2001, responses to the Bulletin and during numerous conference calls and meetings in violation of 10 CFR 50.9 and 10 CFR 50.5(a)(2). The OI report (No. 3-2002-006) was issued on August 22, 2003. A copy of the OI report was provided to the U.S. Department of Justice (DOJ), Office of the United States Attorney, Northern District of Ohio for review. The matter remains under continued Federal investigation. Mr. Miller, through the performance of his duties as a supervisor in the licensee's regulatory affairs organization, and through oral and written communications with other FENOC employees was aware of the results of previous RPV head inspections. For example: • Mr. Miller received several E-mails during August 2001, while FENOC was preparing the September 4, 2001, response to the NRC. These E-mails, in part, made Mr. Miller aware that the boric acid deposits on the RPV head and the RPV head service structure weepholes were an impediment to viewing all RPV head nozzle penetrations. • Mr. Miller received a copy of an E-mail, dated August 28, 2001, that questioned whether a discussion in the licensee's draft response to the Bulletin relative to a subsequent review of 1998 and 2000 inspection videotaped results should be reworded. The August 28, 2001, E-mail received by Mr. Miller stated, in part: “the discussion gives an impression to the reader that we were able to look at all the CRDMs. It is very difficult to look at the CRDMs when there is boric acid around it.” • Mr. Miller also received a copy of an E-mail, dated August 30, 2001, in which the author stated, in part: “I have not seen any EWR [engineering work request] to cut openings in the service structure in the 13th RFO. If we need these it should be funded and P.O. [Purchase Order] issued to Framatone immediately. We do not say anywhere in our response to the Bulletin that inspection thru the mouse holes creates an impediment for 100% visual inspection examination. (Management need[s] to know this).” • During a sworn, transcribed interview with OI, Mr. Miller stated that if the author of the E-mail was concerned about addressing the impediments [discussed in the E-mails listed above] before the licensee issued its response to the Bulletin the individual should have brought it to the attention of his supervisor and his management chain in the Engineering Department. • Mr. Miller also told OI that he looked-up the word “impediment” in the dictionary upon being informed of the size of the RPV head service structure weepholes, the two inch gap between the RPV head and the insulation at the top of the RPV head, the RPV head curvature, and the inspection limitations resulting from the presence of boron deposits. Specifically, Mr. Miller stated: “I even went to the point of looking up the word “impede” in the dictionary, you know. It says obstruct or hinder. Obstruct. Does the mouse hole obstruct? No. Does the curvature of the head obstruct? No. Does the two inch gap obstruct? No. Does it hinder? It may hinder it, but again, I think the collective thought was that it could be done.” Mr. Miller concluded that impediment meant something that obstructed or hindered. Using the dictionary definition, Mr. Miller concluded that none of these issues obstructed an inspection, though these issues may hinder it. • Mr. Miller also stated in his interview with OI that at the time the September 4, 2001, response was being issued to the NRC: “From what I knew, at that time they were able to look at them to a degree, but because there was boron, you know, on the head in some areas, it couldn't be credited as a qualified visual inspection. It's very difficult to look at CRDMs when there is boric acid around it. And in a sense, we were looking—we were—and my understanding at that time was that we were looking, you know, can we inspect to see that there's, you know, popcorn boron, or whatever, and it's very difficult to look at the CRDMs when there's boric acid around it. In other words, to me, it doesn't really say, it doesn't talk about, you know, and I'm speaking now, you know, somewhat what I know now, too. And this is where it's very difficult. You look back at this stuff and you could say, oh, for sure, you know, oh, it was obvious to the casual observer. Well, not to me it wasn't, because, you know, I'm this licensing guy taking input from engineering. It is very difficult to look at CRDMs when there's boric acid around it.” The above information demonstrates that Mr. Miller had sufficient knowledge of the results of previous inspections of the RPV head and that he knew that the licensee's written response to NRC Bulletin 2001-001 was incomplete and inaccurate. Several FENOC employees, including Mr. Dale Miller, were responsible for the information provided to the NRC by FENOC in response to the Bulletin. III Dale Miller was employed by FENOC as a Compliance Supervisor in the Regulatory Affairs organization at Davis-Besse at the time the responses to the Bulletin were developed and transmitted to the NRC. Additionally, Mr. Miller was the supervisor of the individual assigned the responsibility to prepare the September 4, 2001, response to the Bulletin. On August 30, 2001, Mr. Miller concurred as the “Supervisor, DB Compliance” in the issuance of the licensee's September 4, 2001, response to the Bulletin. Item 1.d of the Bulletin requested each PWR licensee, including FENOC for Davis-Besse, provide a description of the RPV head penetration nozzles and RPV head inspection (including type, scope, qualification requirements, and acceptance criteria) that were performed at PWRs in the 4 years preceding the date of the Bulletin, and the findings resulting from the inspections. The licensee's were requested to include a description of any limitations (insulation or other impediments) to accessibility of the bare metal of the RPV head for visual examinations. On September 4, 2001, FENOC submitted its written response to the Bulletin for Davis-Besse. Item 1.d of the licensee's September 4, 2001, response to the Bulletin stated, in part, “a gap exits between the RPV head and the insulation, the minimum gap being at the dome center of the RPV head where it is approximately 2 inches, and does not impede visual inspection.” The licensee included a description of the Eleventh Refueling Outage (11RFO) (April 1998) inspection of RPV head penetration nozzles and RPV head at Davis-Besse in its September 4, 2001, letter to the NRC, and stated, in part, “The head was cleaned by use of a manual scrubber and vacuum through the weepholes.” The licensee's September 4, 2001, response also described the results of the inspections conducted during 12RFO (April 2000) and included a statement that: “Inspection of the RPV head/nozzles area indicated some accumulation of boric acid deposits. The boric acid deposits were located beneath the leaking flanges with clear evidence of downward flow. No visible evidence of nozzle leakage was detected.” The licensee's September 4, 2001, response was materially incomplete and inaccurate in that the response did not describe impediments to accessing the RPV head bare metal during the 11RFO

(1998)and 12RFO (2000). Access to the RPV head bare metal was limited due to significant accumulations of boric acid deposits and the size of the service structure access holes. Based on the above information, the NRC concludes that Mr. Miller had sufficient knowledge of the condition of the RPV head and the limitations experienced during RPV head inspections, and he deliberately provided materially incomplete and inaccurate information when, on August 30, 2001, Mr. Miller concurred on the licensee's September 4, 2001, response to the NRC. The information provided by the licensee under oath in the Bulletin response, based, in part, on the concurrence of Mr. Miller, was material to the NRC because the NRC used the information, in part, to allow FENOC to operate Davis-Besse until February 2002 rather than requiring the plant to shut down by December 31, 2001, to conduct inspections of the head as discussed in Item 3.v.1. of the Bulletin. Based on the above information, Mr. Dale Miller, while employed by the licensee, engaged in deliberate misconduct by deliberately providing FENOC and the NRC information that he knew was not complete or accurate in all material respects to the NRC, a violation of 10 CFR 50.5(a)(2). Mr. Miller's actions also placed FENOC in violation of 10 CFR 50.9. The NRC determined that these violations were of very high safety and regulatory significance because they demonstrated a pattern of deliberate inaccurate or incomplete documentation of information that was required to be submitted to the NRC pursuant to 10 CFR 50.54(f). Had the NRC been aware of this incomplete and inaccurate information, the NRC would likely have taken immediate regulatory action to shut down the plant and require the licensee to implement appropriate corrective actions. IV The NRC must be able to rely on the licensee and its employees to comply with NRC requirements, including the requirement to provide information and maintain records that are complete and accurate in all material respects. Mr. Miller's deliberate actions raised serious doubt as to whether he can be relied upon to comply with NRC requirements and to provide complete and accurate information to the NRC. Consequently, I lack the requisite reasonable assurance that licensed activities can be conducted in compliance with the Commission's requirements and that the health and safety of the public will be protected if Mr. Miller is permitted to be involved in NRC-licensed activities. Therefore, the public health, safety and interest require that Mr. Miller be prohibited from any involvement in NRC-licensed activities for a period of five years effective immediately. Additionally, Mr. Miller is required to notify the NRC of his first employment in NRC-licensed activities for a period of five years following the prohibition period. V Accordingly, pursuant to sections 103, 104, 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, 10 CFR 50.5, and 10 CFR 150.20, *It is hereby ordered* that effective immediately: 1. Mr. Dale Miller is prohibited for five years from the date of this Order from engaging in NRC-licensed activities. The NRC considers NRC-licensed activities to be those activities that are conducted pursuant to a specific or general license issued by the NRC, including those activities of Agreement State licensees conducted pursuant to the authority granted by 10 CFR 150.20. 2. If Mr. Miller is currently involved with another licensee in NRC-licensed activities, he must immediately cease those activities, and inform the NRC of the name, address and telephone number of the employer, and provide a copy of this Order to the employer. 3. For a period of five years after the five-year period of prohibition has expired, Mr. Miller shall, within 20 days of acceptance of his first employment offer involving NRC-licensed activities or his becoming involved in NRC-licensed activities, as defined in Paragraph IV.1 above, provide notice to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555, of the name, address, and telephone number of the employer or the entity where he is, or will be, involved in NRC-licensed activities. In the notification, Mr. Miller shall include a statement of his commitment to compliance with regulatory requirements and the basis why the Commission should have confidence that he will now comply with applicable NRC requirements. The Director, Office of Enforcement, may, in writing, relax or rescind any of the above conditions upon demonstration by Mr. Miller of good cause. VI In accordance with 10 CFR 2.202, Dale Miller 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, consideration may be given to extending the response time for submitting an answer as well as the time for requesting a hearing, for good cause shown. A request for extension of time must be made in writing to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 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 admit or deny each allegation or charge made in this Order and shall set forth the matters of fact and law on which Mr. Miller 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, U.S. Nuclear Regulatory Commission, Attn: Rulemakings and Adjudications Staff, Washington, DC 20555. Copies also shall be sent to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555, to the Assistant General Counsel for Materials Litigation and Enforcement at the same address, to the Regional Administrator, NRC Region III, 2443 Warrenville Road, Lisle, IL 60532-4352, and to Mr. Miller if the answer or hearing request is by a person other than Mr. Miller. Because of continuing 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 Office of the General Counsel either by means of facsimile transmission to 301-415-3725 or by e-mail to *OGCMailCenter@nrc.gov.* If a person other than the Mr. Miller 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 2.309. If a hearing is requested by Mr. Miller 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. Pursuant to 10 CFR 2.202(c)(2)(I), Mr. Miller, may, in addition to demanding a hearing, at the time the answer is filed or sooner, move the presiding officer to set aside the immediate effectiveness of the Order on the ground that the Order, including the need for immediate effectiveness, is not based on adequate evidence but on mere suspicion, unfounded allegations, or error. In the absence of any request for hearing, or written approval of an extension of time in which to request a hearing, the provisions specified in Section V above shall be effective immediately 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 V shall be final when the extension expires if a hearing request has not been received. Dated this 4th day of January 2006. For the Nuclear Regulatory Commission. Martin J. Virgilio, Deputy Executive Director for Materials, Research, State, and Compliance Programs, Office of the Executive Director for Operations. [FR Doc. E6-438 Filed 1-13-06; 8:45 am] BILLING CODE 7590-01-P NUCLEAR REGULATORY COMMISSION [IA-05-054] Steven Moffitt; Order Prohibiting Involvement in NRC-Licensed Activities (Effective Immediately) I Mr. Steven Moffitt was previously employed, at times relevant to this Order, as the Technical Services Director at the Davis-Besse Nuclear Power Station (Davis-Besse) operated by FirstEnergy Nuclear Operating Company (FENOC or licensee). The licensee holds License No. NPF-3 which was issued by the Nuclear Regulatory Commission (NRC or Commission) pursuant to 10 CFR Part 50 on April 22, 1977. The license authorizes the operation of Davis-Besse in accordance with the conditions specified therein. The facility is located on the Licensee's site near Oak Harbor, Ohio. II On August 3, 2001, the NRC issued Bulletin 2001-001, “Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles,” (Bulletin). In the Bulletin, the NRC requested that all holders of operating licenses for pressurized water nuclear power reactors (PWR), including FENOC for the Davis-Besse facility, provide information to the NRC relating to the structural integrity of the reactor pressure vessel

(RPV)head penetration nozzles at their respective facilities. The information requested from the licensees included the extent of RPV head penetration nozzle leakage and cracking that had been found to date, a description of the inspections and repairs undertaken to satisfy applicable regulatory requirements, and the basis for concluding that a licensee's plans for future inspections would ensure compliance with applicable regulatory requirements. The NRC also required that all the Bulletin addressees, including FENOC, submit a written response to the NRC in accordance with the provisions of 10 CFR 50.54(f). That regulation provides, in part, that upon request of the NRC, an NRC-licensee must submit written statements, signed under oath or affirmation, to enable the NRC to determine whether the license should be modified, suspended, or revoked. On September 4, October 17, and October 30, 2001, the licensee provided written responses to the Bulletin. Additionally, the licensee met with the NRC staff on numerous occasions during October and November of 2001 to provide clarifying information. Based, in part, on the information provided by FENOC in its written responses to the Bulletin and during meetings with the NRC staff, the NRC staff allowed the licensee to continue operation of the Davis-Besse facility until February 2002, rather than requiring FENOC to shut the unit down to perform inspections by December 31, 2001, as provided in the Bulletin. On February 16, 2002, FENOC shut down Davis-Besse for refueling and inspection of control rod drive mechanism

(RCS)pressure boundary. A smaller cavity was also found near CRDM Penetration Nozzle No. 2. The licensee conducted a root cause evaluation and determined that, contrary to the earlier information provided to the NRC, the cavities were caused by boric acid from the RCS released through cracks in the CRDM RPV head penetration nozzles. The root cause evaluation found that the licensee had previously conducted limited cleaning and inspections of the RPV head during the Twelfth Refueling Outage (12RFO) that ended on May 18, 2000. However, neither the limited RPV head cleaning nor the resultant inspections during 12RFO were sufficient to ensure that the significant boric acid deposits on the RPV head were only a result of CRDM flange leakage and were not a result of RCS pressure boundary leakage. On March 6 and March 10, 2002, the licensee provided information to the NRC concerning the identification of a large cavity in the RPV head adjacent to CRDM Penetration Nozzle No. 3. The NRC conducted an Augmented Inspection Team

(DB)the Reactor Vessel head inspection was performed in accordance with boron inspection walkdown as required by GL-88-05 and GL 97-01. Large boron leakage from a CRDM flange was observed. This leakage did not permit the detailed inspection of CRDM nozzles.” • On August 11, 2001, FENOC held a meeting to discuss its pending response to the Bulletin. Mr. Moffitt was listed as an attendee at the meeting, as documented in an E-mail from a design engineer that same day. As stated in the E-mail, “it was pointed out that we can not clean our head thru the mouse holes and a system engineer is requesting that three large holes be cut in the Service Structure for viewing [inspection] and cleaning.” • During a sworn, transcribed interview with OI, Mr. Moffitt stated that around the August 11, 2001, time frame he remembered talking to the engineer who had cleaned the RPV head regarding how much of the head was cleaned. Mr. Moffitt further stated that the engineer told him about 80 percent of the head was cleaned. • During September 2001, Mr. Moffitt hired a contractor employed by Piedmont Management and Technical Services, Inc. to review Davis-Besse's preparation for 13RFO with implementing the requirements of Bulletin 2001-001. On September 14, 2001, the contractor provided Mr. Moffitt a copy of the letter [report] containing his recommendations and approximately one week later verbally briefed Mr. Moffitt on the contents of the report. The report stated, in part: “It is noted that on completion of 12RFO, the Reactor Vessel head did have boric acid crystal deposits of considerable depth left in the center top area of the head, since cleaning of this area at that time was not successful in removing all the deposits (partly due to limited access).” • During a licensee interview of Mr. Moffitt on July 1, 2002, Mr. Moffitt indicated that he knew in the July to August 2001 time-frame that boric acid was left on the head in 12RFO and that the boric acid impeded a complete inspection of the head. The above information demonstrates that Mr. Moffitt had sufficient knowledge of the results of previous inspections of the RPV head and that he knew the licensee's written and oral responses to NRC Bulletin 2001-001 were incomplete and inaccurate. Several FENOC employees, including Mr. Steven Moffitt, were responsible for the information provided to the NRC by FENOC in response to the Bulletin. III Steven Moffitt was employed by FENOC as the Technical Services Director at Davis-Besse at the time the responses to the Bulletin were developed and transmitted to the NRC. Mr. Moffitt participated in an October 3, 2001, teleconference with the NRC staff and a presentation on October 11, 2001, to the NRC Commissioners' Technical Assistants. On October 17, 2001, Mr. Moffitt concurred in the issuance of the supplemental licensee response, dated October 17, 2001. On October 3, 2001, Mr. Moffitt was a senior Davis-Besse management official on a conference call with the NRC staff. Mr. Moffitt was also involved in preparatory meetings for the October 3rd conference call. The agenda for the conference call stated: “Video Inspection Review from RFO10, RFO11, and RFO12: Further Confirmation of no indication of leakage attributable to CRDM Nozzle leakage; clearly CRDM flange leakage.” During the conference call, Mr. Moffitt's direct subordinate informed the NRC that 100% of the RPV head had been inspected during the last outage (12RFO) but that some areas were precluded from inspection and that videotapes of the inspections conducted during 10RFO, 11RFO, and 12RFO had been reviewed. Mr. Moffitt was aware at the time of the October 3, 2001, meeting that the licensee did not conduct a 100% inspection of the RPV head during 12RFO due to the presence of boric acid on the head which obscured a significant number of the RPV head nozzles yet approved the misleading statements thereby causing the incomplete and inaccurate information to be submitted to the NRC. On October 10, 2001, Mr. Moffitt participated in a meeting with other FENOC officials for the purpose of finalizing presentation slides to be used during an October 11, 2001, meeting with the NRC Commissioner's Technical Assistants. Draft Presentation Slide 20 stated: “Reviewed video inspections of Reactor Vessel head taken during 11RFO (April 1998) and 12RFO (April 2000) and confirmed that Davis-Besse has not experienced boron leakage as seen at Oconee or Arkansas Nuclear.” Presentation Draft Slide 21 for the briefing stated: “Reviewed past 3 outages of Reactor Vessel Head inspection video tapes which were taken to satisfy Generic Letter 97-01: No telltale “popcorn” type boron deposits; During 12RFO (Spring 2000), Davis-Besse identified sources of boron that precluded the visual inspection of some CRDM penetrations, as five leaking flanges above the mirror insulation; Viewed past 3 outages of inspection video tapes of area masked by boron in 12 RFO did not have previous leakage.” On October 11, 2001, Mr. Moffitt and other licensee staff briefed the NRC Commissioners' Technical Assistants on FENOC's basis for concluding that Davis-Besse was safe to operate until the next refueling outage (March 2002). During the briefing, FENOC utilized the presentation slides that were finalized the previous day. Presentation Slide 6 stated, in part: “Conducted and recorded video inspections of the head during 11RFO (April 1998) and 12RFO (April 2000)—No head penetration leakage was identified.” Presentation Slide 7 stated, in part: “All CRDM [control rod drive mechanism] penetrations were verified to be free from “popcorn” type boron deposits using video recordings from 11RFO or 12RFO.” The licensee's October 11, 2001, presentation to the NRC Commissioners' Technical Assistants was materially incomplete and inaccurate in that the presentation slides did not state that the build-up of boric acid on the RPV head was so significant that the licensee could not inspect all of the RPV head penetration nozzles. Due to the significant amount of boric acid present on the RPV head, of which Mr. Moffitt was aware, the licensee also did not have a basis for stating that no visible evidence of RPV penetration nozzle leakage was detected. Mr. Moffitt knew the information was incomplete and inaccurate and allowed it to be submitted to the NRC. On October 17, 2001, the licensee provided a supplemental response to the Bulletin. The second paragraph under the section entitled, “Previous Inspection Results,” on Page 2 of Attachment 1 of the licensee's October 17, 2001, supplemental response stated, in part: “The inspections performed during the 10th, 11th, and 12th Refueling Outage (10RFO, conducted April 8 to June 2, 1996; 11RFO, conducted April 10 to May 23, 1998; and, 12RFO, conducted April 1 to May 18, 2000) consisted of a whole head visual inspection of the RPV head in accordance with the DBNPS Boric Acid Control Program pursuant to Generic Letter 88-05 “Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants.” The visual inspections were conducted by remote camera and included below insulation inspections of the RPV bare head such that the Control Rod Drive Mechanism

(CRDM)nozzle penetrations were viewed. During 10RFO, 65 of 69 nozzles were viewed, during 11RFO, 50 of 69 nozzles were viewed, and during 12RFO, 45 of 69 nozzles were viewed. It should be noted that 19 of the obscured nozzles in 12RFO were also those obscured in 11RFO.” Information included under Column 6 of Attachment 2 of the licensee's October 17, 2001, response stated, in part, that 24 nozzles have a “flange leak evident.” Note 1 on the same table stated, in part: “In 1996 during 10 RFO, the entire RPV head was inspected. Since the video was void of head orientation narration, each specific nozzle view could not be correlated.” The licensee's October 17, 2001, supplemental response was materially incomplete and inaccurate, in that the licensee did not view the stated number of RPV head penetration nozzles during the referenced outages, and the licensee believed that only five RPV head control rod drive mechanism flanges were leaking instead of the 24 RPV head control rod drive mechanism flanges noted in the response. Specifically, during 12RFO the licensee did not clean all of the RPV head; therefore, the licensee could not have viewed each of the RPV head penetration nozzles and determined that the observed boric acid accumulation was not a result of RPV nozzle leakage. Mr. Moffitt knew the information was incomplete and inaccurate but nonetheless, concurred on the response, thereby allowing the information to be submitted to the NRC. Based on the above information, the NRC concludes that Mr. Moffitt had knowledge of the condition of the RPV head and the limitations experienced during RPV head inspections, and he deliberately failed to ensure that information that was developed for and presented during an October 3, 2001, teleconference with the NRC; was developed during an October 10, 2001, meeting and presented during an October 11, 2001, meeting with the NRC; and was included in the licensee's October 17, 2001, supplemental response to the NRC Bulletin 2001-001 was materially complete and accurate. The information presented to the NRC and provided in the licensee's October 17, 2001, supplemental response was material to the NRC because the information gave the impression to the NRC staff that the Davis-Besse RPV head had been completely inspected for evidence of nozzle cracks, when this was not the case at the time the information was provided or the supplemental response was submitted. In addition, information provided during the October 3 and October 11, 2001, meetings and in the licensee's October 17, 2001, supplemental response to the NRC was material to the NRC because the NRC used the information, in part, to allow FENOC to operate Davis-Besse until February 2002 rather than requiring the plant to shut down by December 31, 2001, to conduct inspections of the RPV head as discussed in Item 3.v.1 of the Bulletin. Based on the above, Mr. Steven Moffitt, while employed by the licensee, engaged in deliberate misconduct by providing FENOC and the NRC information that he knew was not complete and accurate in all material respects to the NRC, a violation of 10 CFR 50.5(a)(2). Mr. Moffitt's actions also placed FENOC in violation of 10 CFR 50.9. The NRC determined that these violations were of very high safety and regulatory significance because they demonstrated a pattern of deliberate inaccurate or incomplete documentation of information that was required to be submitted to the NRC. Had the NRC been aware of this incomplete and inaccurate information, the NRC would likely have taken immediate regulatory action to shut down the plant and require the licensee to implement appropriate corrective actions. IV The NRC must be able to rely on the licensee and its employees to comply with NRC requirements, including the requirement to provide information and maintain records that are complete and accurate in all material respects. Mr. Moffitt's deliberate actions raise serious doubt as to whether he can be relied upon to comply with NRC requirements and to provide complete and accurate information to the NRC. Consequently, I lack the requisite reasonable assurance that licensed activities can be conducted in compliance with the Commission's requirements and that the health and safety of the public will be protected if Mr. Moffitt is permitted to be involved in NRC-licensed activities. Therefore, the public health, safety and interest require that Mr. Moffitt be prohibited from any involvement in NRC-licensed activities for a period of five years from the date of this Order. Additionally, Mr. Moffitt is required to notify the NRC of his first employment in NRC-licensed activities for a period of five years following the prohibition period. V Accordingly, pursuant to sections 103, 104, 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, 10 CFR 50.5, and 10 CFR 150.20, *It is hereby ordered* that effective immediately: 1. Mr. Steven Moffitt is prohibited for five years from the date of this Order from engaging in NRC-licensed activities. The NRC considers NRC-licensed activities to be those activities that are conducted pursuant to a specific or general license issued by the NRC, including those activities of Agreement State licensees conducted pursuant to the authority granted by 10 CFR 150.20. 2. If Mr. Moffitt is currently involved with another licensee in NRC-licensed activities, he must immediately cease those activities, and inform the NRC of the name, address and telephone number of the employer, and provide a copy of this Order to the employer. 3. For a period of five years after the five-year period of prohibition has expired, Mr. Moffitt shall, within 20 days of acceptance of his first employment offer involving NRC-licensed activities or his becoming involved in NRC-licensed activities, as defined in Paragraph IV.1 above, provide notice to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555, of the name, address, and telephone number of the employer or the entity where he is, or will be, involved in NRC-licensed activities. In the notification, Mr. Moffitt shall include a statement of his commitment to compliance with regulatory requirements and the basis why the Commission should have confidence that he will now comply with applicable NRC requirements. The Director, Office of Enforcement, may, in writing, relax or rescind any of the above conditions upon demonstration by Mr. Moffitt of good cause. VI In accordance with 10 CFR 2.202, Steven Moffitt 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. However, since this enforcement action is being proposed prior to the U.S. Department of Justice completing its review of the OI investigation results, consideration may be given to extending the response time for submitting an answer as well as the time for requesting a hearing, for good cause shown. A request for extension of time must be made in writing to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 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 admit or deny each allegation or charge made in this Order and shall set forth the matters of fact and law on which Mr. Moffitt 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, U.S. Nuclear Regulatory Commission, Attn: Rulemakings and Adjudications Staff, Washington, DC 20555. Copies also shall be sent to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555, to the Assistant General Counsel for Materials Litigation and Enforcement at the same address, to the Regional Administrator, NRC Region III, 2443 Warrenville Road, Lisle, IL 60532-4352, and to Mr. Moffitt if the answer or hearing request is by a person other than Mr. Moffitt. Because of continuing 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 Office of the General Counsel either by means of facsimile transmission to 301-415-3725 or by e-mail to *OGCMailCenter@nrc.gov.* If a person other than Mr. Moffitt 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 2.309. If a hearing is requested by Mr. Moffitt 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. Pursuant to 10 CFR 2.202(c)(2)(I), Mr. Moffitt, may, in addition to demanding a hearing, at the time the answer is filed or sooner, move the presiding officer to set aside the immediate effectiveness of the Order on the ground that the Order, including the need for immediate effectiveness, is not based on adequate evidence but on mere suspicion, unfounded allegations, or error. In the absence of any request for hearing, or written approval of an extension of time in which to request a hearing, the provisions specified in Section V above shall be effective immediately 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 V shall be final when the extension expires if a hearing request has not been received. Dated this 4th day of January 2006. For the Nuclear Regulatory Commission. Martin J. Virgilio, Deputy Executive Director for Materials, Research, State, and Compliance Programs, Office of the Executive Director for Operations. [FR Doc. E6-416 Filed 1-13-06; 8:45 am] BILLING CODE 7590-01-P NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-321 and 50-366] Southern Nuclear Operating Company, Inc., Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2 Exemption 1.0 Background The Southern Nuclear Operating Company, Inc. (SNC, or the licensee), is the holder of Facility Operating License Nos. DPR-57 and NPF-5 which authorizes operation of the Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2 (Hatch 1 and 2), respectively. The license provides, among other things, that the facility is subject to all rules, regulations, and orders of the Nuclear Regulatory Commission (NRC, Commission) now or hereafter in effect. The facility consists of two boiling water reactors located in Appling County, Georgia. 2.0 Request/Action Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(b)(2)(ix), states the requirements for the examination of metal containments and liners of concrete containments. In particular, Section 50.55a(b)(2)(ix)(G) requires, in part, that a VT-3 examination method be used to conduct examinations of Item E.20 of Table IWE-2500-1 of Section IX of the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code (ASME Code). By letter dated March 30, 2005, as supplemented by letters dated August 2 and 24, 2005, the licensee submitted a request for an exemption from the requirements of Section 50.55a(b)(2)(ix)(G). The exemption request would allow the licensee to perform an alternative examination of the accessible surface areas of the containment vessel pressure retaining boundary vent system, in lieu of the VT-3 examination required by the rule. The licensee stated that the alternate examination method is currently in use at Hatch 1 and 2 and has proven to be sufficient to maintain the structural integrity and leak-tightness of the containment surfaces, and, therefore, serves the underlying purpose of the rule. The licensee is currently in its 3rd 10-year inservice inspection

(ISI)interval. The licensee's code of record for the 3rd 10-year ISI interval is the 1992 edition through the 1992 addenda of the ASME Code. The code of record contains the requirement to perform a VT-3 examination of the accessible surface areas of the vent system. In Relief Request RR-MC-9 submitted by letter dated July 19, 2000, the licensee requested relief from the requirement to perform a VT-3 examination on nonsubmerged, accessible pressure boundary surfaces, including the vent system, at the end of the 3rd 10-year ISI interval. The licensee explained that the proposed alternative to perform a general visual examination was sufficient to detect the types of corrosion expected in the components covered by the relief. On October 4, 2000, this request was approved by the NRC staff. The licensee's 4th 10-year ISI interval is scheduled to begin in 2006. The licensee's code of record for this interval will be the 2001 edition through the 2003 addenda of the ASME Code. Modifications to the ASME Code and 10 CFR 50.55a since the beginning of the 3rd 10-year ISI interval have relocated the requirement to perform the subject VT-3 examination from the ASME Code to 10 CFR 50.55a(b)(2)(ix). As a result, licensees wanting relief from the requirement to perform a VT-3 examination for the subject structures must now request an exemption from the requirements of 10 CFR 50.55a(b)(2)(ix)(G). The licensee stated in its August 24, 2005, letter that the examination provisions previously authorized through Relief Request RR-MC-9 have proven to be sufficient to maintain the structural integrity and leak-tightness of the containment surfaces, and, therefore, serve the underlying purpose of the rule. As an alternative to the VT-3 examination, SNC is proposing the examination on all nonsubmerged, accessible pressure boundary surfaces of the vent system. This general visual-type examination will be performed in accordance with the Hatch 1 and 2 Qualified

(N)Coatings Program. The licensee indicated that the details of this program were provided in the October 19, 1998, response to NRC Generic Letter 98-04, “Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System after a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment.” The procedures and personnel qualifications applicable for the coatings program implementation are in compliance with Regulatory Guide 1.54 (1973), and the implementation is based on the following documents:

(1)ANSI N 101.2-1972, “Protective Coatings (Plants) for Light Water Nuclear Reactor Containment Facilities;”

(2)ANSI N101.4-1972, “Quality Assurance for Protective Coatings Applied to Nuclear Facilities;” and

(3)EPRI Report TR-109937, “Guideline on Nuclear Safety-Related Coatings.” This program was approved by the NRC staff in a letter dated November 19, 1999. The licensee further noted that the Qualified

(N)Coatings program examination frequency is equivalent to the requirements of Section XI to the ASME Code, and the program requires that when evidence of degradation is detected, a detailed examination and evaluation be performed. The detailed visual examination would be performed in accordance with the provisions of ASME Code, Section XI, paragraph IWE-2310(c). The exterior surfaces of the vent system that connects the drywell to the suppression pool are located in the reactor building. The reactor building environment does not pose adverse conditions that would promote rapid degradation of the outside pressure boundary surfaces of the vent system. The interior surfaces of the vent system that connect the drywell to the suppression pool and the portions of the vent system located inside the suppression pool are maintained in a nitrogen inerted environment during normal power operation in accordance with technical specification requirements. Operational experience and previous examinations have indicated that this environment does not promote rapid degradation of the surfaces. The licensee stated that the requirements specified for a VT-3 examination were developed for detecting flaws in metal components and are more stringent than those required for detecting corrosion-related degradation. Since corrosion of base metal is the primary issue of concern for containment pressure boundary surface areas, a general visual-type examination, in accordance with the Hatch 1 and 2 Qualified

(N)Coatings Program, is sufficient to inspect the subject surface areas of the containment and will provide an acceptable level of quality and safety. In summary, the licensee is proposing an exemption from the requirements of Section 50.55a(b)(2)(ix)(G) to use an alternate examination method to examine Item E.20 of Table IWE-2500-1 of ASME Code, Section XI, pursuant to 10 CFR 50.12(a)(1) and 10 CFR 50.12(a)(2)(ii). The licensee stated in its application that compliance with the visual examination requirements of Section 50.55a(b)(2)(ix)(G) is not necessary for accessible surface areas of the containment vessel pressure retaining boundary Vent System to achieve the underlying purpose of the rule. 3.0 Discussion Pursuant to 10 CFR 50.12, the Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of 10 CFR Part 50 when:

(1)The exemptions are authorized by law, will not present an undue risk to public health or safety, and are consistent with the common defense and security; and

(2)when special circumstances are present. Special circumstances are present whenever, in accordance with 10 CFR Part 50.12(a)(2)(ii), “Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule * * *.” Therefore, in determining the acceptability of the licensee's exemption request, the NRC staff has performed the following evaluation to satisfy the requirements of 10 CFR 50.12 for granting the exemption. The underlying purpose of 10 CFR 50.55a(b)(2)(ix)(G), as it applies to Item E1.20 of Table IWE-2500-1, is to ensure that an examination of the metal containment or the metal liner of a concrete containment is performed to identify corrosion or other degradation that could affect the structural or leak-tight integrity of the structure. The NRC staff examined the licensee's rationale to support the exemption request and concluded that maintaining the integrity of the coating system applied to the Hatch 1 and 2 containment vent system components is a preventive measure that would protect against corrosion of the coated components. As the licensee emphasizes the effectiveness of its coating program, the NRC staff believes that the general visual examination performed as part of maintaining the integrity of the coating system is a proactive action and will ensure the integrity of the coated vent system components. The proposed alternative will provide the quality and safety level similar to the one intended by the use of VT-3 examination of the vent system components, and would meet the underlying purpose of 10 CFR Section 50.55a(b)(2)(ix)(G). Based on a consideration of proposed alternatives contained in the licensee's letters dated March 20, and August 2 and 24, 2005, the NRC staff concludes that degradation of the containment structure would be detected using the proposed alternative, thus meeting the underlying purpose of the rule. Therefore, the NRC staff concludes that the proposed exemption from 10 CFR Section 50.55a(b)(2)(ix)(G) is acceptable. 4.0 Conclusion Accordingly, the Commission has determined that, pursuant to 10 CFR 50.12, the exemption is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security. Also, special circumstances are present. Therefore, the Commission hereby grants SNC an exemption from the requirement of 10 CFR Section 50.55a(b)(2)(ix)(G) to perform a VT-3 examination for Item E1.2 of Table IWE-2500-1, for Hatch 1 and 2, for the 4th 10-year ISI interval. Pursuant to 10 CFR 51.32, the Commission has determined that the granting of this exemption will not have a significant effect on the quality of the human environment (70 FR 76082). This exemption is effective upon issuance. Dated at Rockville, Maryland, this 6th day of January 2006. For the Nuclear Regulatory Commission. Catherine Haney, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. E6-415 Filed 1-13-06; 8:45 am] BILLING CODE 7590-01-P NUCLEAR REGULATORY COMMISSION Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations I. Background Pursuant to section 189a.

(2)of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. This biweekly notice includes all notices of amendments issued, or proposed to be issued from December 22, 2005 to January 5, 2006. The last biweekly notice was published on January 3, 2006 (71 FR 145). Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not

(1)involve a significant increase in the probability or consequences of an accident previously evaluated; or

(2)create the possibility of a new or different kind of accident from any accident previously evaluated; or

(3)involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period should circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. Should the Commission take action prior to the expiration of either the comment period or the notice period, it will publish in the **Federal Register** a notice of issuance. Should the Commission make a final No Significant Hazards Consideration Determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently. Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this **Federal Register** notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below. Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, *http://www.nrc.gov/reading-rm/doc-collections/cfr/.* If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order. As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements:

(1)The name, address, and telephone number of the requestor or petitioner;

(2)the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding;

(3)the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and

(4)the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also set forth the specific contentions which the petitioner/requestor seeks to have litigated at the proceeding. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner/requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner/requestor intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner/requestor to relief. A petitioner/requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party. Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing. If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment. A request for a hearing or a petition for leave to intervene must be filed by:

(1)First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff;

(2)courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff;

(3)E-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, *HearingDocket@nrc.gov* ; or

(4)facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at

(301)415-1101, verification number is

(301)415-1966. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and it is requested that copies be transmitted either by means of facsimile transmission to

(301)415-3725 or by e- mail to *OGCMailCenter@nrc.gov.* A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee. Nontimely requests and/or petitions and contentions will not be entertained absent a determination by the Commission or the presiding officer of the Atomic Safety and Licensing Board that the petition, request and/or the contentions should be granted based on a balancing of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii). For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the ADAMS Public Electronic Reading Room on the Internet at the NRC Web site, *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 PDR Reference staff at 1

(800)397-4209,

(301)415-4737 or by e-mail to *pdr@nrc.gov.* AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek Nuclear Generating Station (OCNGS), Ocean County, New Jersey *Date of amendment request:* December 2, 2005. *Description of amendment request:* The amendment would revise the Technical Specifications to increase the allowable as-found main steam safety valve code safety function lift setpoint tolerance from ±1% to ±3%. *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Will operation of the facility in accordance with the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed TS changes allow for an increase in the as-found Main Steam Safety Valve

(MSSV)setpoint tolerance from ±1% to ±3%. The proposed changes do not alter the MSSV nominal lift setpoints or MSSV lift setpoint test frequency. The proposed TS changes have been evaluated on both a generic and plant specific basis. The NRC has approved the general approach of this change; however, implementation is contingent on several plant specific evaluations. The required plant specific analyses and evaluations included transient analysis of the anticipated operational transients (AOTs); analysis of the design basis overpressurization event; evaluation of the performance of high pressure systems, and evaluation of the containment response during Loss-of-Coolant Accident

(LOCA)and hydrodynamic loads on the MSSV discharge lines and containment. These analyses and evaluations demonstrate that there is adequate margin to the design core thermal limits and reactor vessel pressure limits using the ±3% MSSV as-found setpoint tolerance. The analyses and evaluations also demonstrate that the operation of high-pressure safety systems will not be adversely affected and that the containment response during a LOCA will be acceptable. Evaluations of the impact of the proposed change on the equipment important to safety have been performed and no adverse conditions were identified. The reactor pressure vessel and attached systems and piping have been evaluated for the impact of this proposed TS change. A plant specific analysis has been performed which indicates that the ASME Code upset limits for the reactor pressure vessel will not be exceeded for the limiting event, i.e., Main Steam Isolation Valve

(MSIV)closure with flux Scram. The reactor pressure vessel and attached piping design values will not be exceeded. Therefore, the probability of a malfunction of the reactor pressure vessel and attached systems and piping is not increased and the consequences of such an accident remain acceptable. The nuclear fuel has been evaluated for the impact of the proposed change. Plant specific analyses were performed which indicate that for all abnormal operational transients adequate margin to the fuel thermal limit parameters, i.e., Minimum Critical Power Ratio

(MCPR)and thermal-mechanical limits, is maintained. Emergency Core Cooling System (ECCS)/LOCA performance is maintained adequate to meet the requirements of 10 CFR 50.46. Therefore, the consequences of these accidents remain acceptable and the probability of the malfunction of the nuclear fuel is not increased. The Containment response during a LOCA has been evaluated for the impact of the proposed change. The major factor in the Containment pressure response to a LOCA is the rate of reactor vessel water inventory loss due to a DBA LOCA. The rate of reactor vessel water inventory loss is mainly dependent on the initial reactor pressure, which is not affected by the proposed setpoint tolerance change. The major factor in the Containment temperature response to a LOCA is the integrated steam inventory loss due to Main Steamline Break. The rate of reactor vessel steam inventory loss is mainly dependent on the reactor decay heat, which is not affected by the proposed setpoint tolerance change. Therefore, the consequences of these accidents remain acceptable and the probability of the malfunction of Containment is not increased. The Control Rod Drive

(CRD)system has been evaluated for the impact of the proposed change. The CRD system capability of controlling reactor power during normal plant operation and rapidly inserting control rod blades (Scram) during abnormal plant conditions is not impacted by the proposed change. Therefore, the probability of a malfunction of the CRD system is not increased. The Reactor Vessel Instrumentation System has been evaluated for the impact of the proposed change. The Reactor Vessel Instrumentation System will continue to be operated within the current design pressure/temperature requirements; therefore, the probability of a malfunction of the Reactor Vessel Instrumentation System is not increased. An administrative change is also being proposed to correct the reference to “IWV-3510 of Section XI of the ASME Boiler and Pressure Vessel Code” in TS 4.3.E because the stated ASME section no longer exists. The TS is being changed to reference specification 4.3.C for MSSV testing. This is an administrative change and does not affect previously evaluated accidents. Therefore, the proposed TS changes do not significantly increase the probability or consequences of an accident previously evaluated. 2. Will operation of the facility in accordance of the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed TS changes allow for an increase in the as-found MSSV setpoint tolerance from ±1% to ±3%. Generic and plant specific analyses and evaluations indicate that the plant response to any previously evaluated event will remain acceptable. All plant systems, structures, and components will continue to be capable of performing their required safety function as required by event analysis guidance. The proposed TS changes do not alter the MSSV nominal lift setpoints or MSSV lift setpoint test frequency. The operation and response of the affected equipment important to safety is unchanged. All systems, structures, and components will continue to be operated within acceptable operating and/or design parameters. No system, structure, or component will be subjected to a condition that has not been evaluated and determined to be acceptable using the guidance required for specific event analysis. The change to correct the reference to “IWV-3510 of Section XI of the ASME Boiler and Pressure Vessel Code” in TS 4.3.E is an administrative change and does not affect the possibility of a new or different kind of accident. Therefore, the proposed TS changes do not create the possibility of a new or different kind of accident from any previously identified. 3. Will operation of the facility in accordance with the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed TS changes allow for an increase in the as-found MSSV setpoint tolerance from ±1% to ±3%. The proposed TS changes do not alter the MSSV nominal lift setpoints or MSSV lift setpoint test frequency. The operation and response of the affected equipment important to safety is unchanged. All systems, structures, and components will continue to be operated within acceptable operating and/or design parameters. While the calculated peak reactor vessel pressure for the ASME overpressure event is higher than that calculated without the increase in setpoint tolerance, it is still within the respective licensing acceptance limits associated with this event. These licensing acceptance limits have been determined by the NRC to provide a sufficient margin of safety. The increase in MSSV steam flow and reactor vessel pressure does not reduce the margin of safety associated with the MSSVs and associated components and structures since the increased MSSV steam flow rate and reactor vessel pressure are bounded by the current design analysis. The margin of safety for fuel thermal limits and 10 CFR 50.46 limits are unaffected by the proposed change. The margin of safety for the Containment is unaffected by the proposed change. The capability of the SLC system and the CRD system to perform their safety functions during all required events, using the required guidance for event analysis, is maintained. Therefore, the proposed changes do not reduce the margin of safety provided by the SLC and CRD systems. The change to correct the reference to “IWV-3510 of Section XI of the ASME Boiler and Pressure Vessel Code” in TS 4.3.E is an administrative change and does not affect the margin of safety. Therefore, these proposed TS changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee* : Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LCC, 4300 Winfield Road, Warrenville, IL 60555. *NRC Branch Chief:* Darrell J. Roberts. Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland *Date of amendments request:* November 3, 2005. *Description of amendments request:* The proposed amendments would revise the accident source term in the design-basis radiological consequences analyses and the associated Technical Specifications (TSs), pursuant to section 50.67 of part 50 of Title 10 of the *Code of Federal Regulations* (10 CFR 50.67). The proposed amendments would provide for the full implementation of the alternate source term

(AST)in accordance with the guidance in Regulatory Guide 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.” The proposed amendments would also increase the flow rate for the control room emergency ventilation system (CREVS) from 2000 to 10000 cubic feet per minute in TS 5.5.11, “Ventilation Filter Testing Program,” by means of a modification to the CREVS. In addition, automatic isolation dampers and radiation monitors will also be installed at access control heating, ventilating, and air conditioning

(HVAC)unit no. RTU-1 and access control air conditioning unit no. 13. *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. The results of the applicable radiological design basis accidents

(DBAs)re-evaluation demonstrated that, with the requested changes, the dose consequences of these limiting events are within the regulatory limits and guidance provided by the Nuclear Regulatory Commission in 10 CFR 50.67 and Regulatory Guide 1.183 for AST methodology. The AST is an input to calculations used to evaluate the consequences of an accident and does not by itself affect the plant response or the actual pathway of the activity released from the fuel. It does, however, better represent the physical characteristics of the release such that appropriate mitigation techniques may be applied. The change from the original source term to the new proposed AST is a change in the analysis method and assumptions and has no effect on accident initiators or causal factors that contribute to the probability of occurrence of previously analyzed accidents. Use of an AST to analyze the dose effect of DBAs shows that regulatory acceptance criteria for the new methodology continues to be met. Changing the analysis methodology does not change the sequence or progression of the accident scenario. The proposed Technical Specification changes reflect the plant configuration that will either support implementation of the AST analyses or eliminate requirements that are no longer needed as a result of the revised DBA analyses. The equipment affected by the proposed changes is mitigative in nature and relied upon after an accident has been initiated. The operation of various filtration systems have been considered in the evaluations for these proposed changes. While the operation of some systems does change with the implementation of an AST, the affected systems are not accident initiators; and application of the AST methodology, itself, is not an initiator of a DBA. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. As described in Item 1 above, the changes proposed in this license amendment request involve the use of a new analysis methodology and related regulatory acceptance criteria. The proposed Technical Specification changes reflect the plant configuration that will either support implementation of the new methodology or eliminate requirements that are no longer needed as a result of the new methodology. No new or different accidents result from utilizing the proposed changes. Although the proposed changes require modification to the Control Room emergency ventilation system and installation of automatic isolation dampers and radiation monitors at Access Control HVAC Unit RTU-1 and Access Control Air Conditioning Unit 13 on the Auxiliary Building roof, none of these changes can initiate a new or different kind of accident since they are only related to system capabilities that provide protection from accidents that have already occurred. As a result, no new failure modes are being introduced that could lead to different accidents. These changes do not alter the nature of events postulated in the Updated Final Safety Analysis Report nor do they introduce any unique precursor mechanisms. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety. As described in Item 1 above, the changes proposed in this license amendment request involve the use of a new analysis methodology and related regulatory acceptance criteria. The proposed Technical Specification changes reflect the plant configuration that will either support implementation of the new methodology or eliminate requirements that are no longer needed as a result of the new methodology. Safety margins and analytical conservatisms have been evaluated and have been found acceptable. The analyzed events have been carefully selected and, with plant modification, margin has been retained to ensure that the analyses adequately bound postulated event scenarios. The analyses have been performed using conservative methodologies, as specified in Regulatory Guide 1.183. The dose consequences of these DBAs remain within the acceptance criteria presented in 10 CFR 50.67, “Accident Source Term,” and Regulatory Guide 1.183. The proposed changes continue to ensure that the doses at the exclusion area boundary and low population zone boundary, as well as the Control Room, are within corresponding regulatory limits. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendments request involves no significant hazards consideration. *Attorney for licensee:* Carey Fleming, Sr. Counsel—Nuclear Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 17th floor, Baltimore, MD 21202. *NRC Branch Chief:* Richard J. Laufer. Exelon Generation Company, LLC, Docket No. 50-352, Limerick Generating Station, Unit 1, Montgomery County, Pennsylvania *Date of amendment request:* December 14, 2005. *Description of amendment request:* The proposed amendment modifies the Technical Specifications

(TSs)to incorporate a revised Single Loop Operation Safety Limit Minimum Critical Power Ratio (SLO SLMCPR) due to the cycle-specific analysis. *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The derivation of the cycle specific Single Loop Operation Safety Limit Minimum Critical Power Ratio (SLO SLMCPR) for incorporation into the Technical Specifications (TS), and its use to determine cycle-specific thermal limits, has been performed using the methodology discussed in “General Electric Standard Application for Reactor Fuel,” NEDE-24011-P-A-15 (GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-15-US, September, 2005, which includes Amendment 25. Amendment 25 was approved by the NRC in a March 11, 1999 safety evaluation report. The basis of the SLO SLMCPR calculation is to ensure that greater than 99.9% of all fuel rods in the core avoid transition boiling if the limit is not violated. The new SLO SLMCPR preserves the existing margin to transition boiling. The GE-14 fuel is in compliance with Amendment 22 to “General Electric Standard Application for Reactor Fuel,” NEDE-24011-P-A-15 (GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-15-US, September 2005, which provides the fuel licensing acceptance criteria. The probability of fuel damage will not be increased as a result of this change. Therefore, the proposed TS change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The SLO SLMCPR is a TS numerical value, calculated to ensure that transition boiling does not occur in 99.9% of all fuel rods in the core if the limit is not violated. The new SLO SLMCPR is calculated using NRC approved methodology discussed in “General Electric Standard Application for Reactor Fuel,” NEDE-24011-P-A-15 (GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-15-US, September 2005, which includes Amendment 25. Additionally, the GE-14 fuel is in compliance with Amendment 22 to “General Electric Standard Application for Reactor Fuel,” NEDE-24011-P-A-15 (GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-15-US, September, 2005, which provides the fuel licensing acceptance criteria. The SLO SLMCPR is not an accident initiator, and its revision will not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. There is no significant reduction in the margin of safety previously approved by the NRC as a result of the proposed change to the SLO SLMCPR, which includes the use of GE-14 fuel. The new SLO SLMCPR is calculated using methodology discussed in “General Electric Standard Application for Reactor Fuel,” NEDE-24011-P-A-15 (GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-15-US, September, 2005, which includes Amendment 25. The SLO SLMCPR ensures that greater than 99.9% of all fuel rods in the core will avoid transition boiling if the limit is not violated when all uncertainties are considered, thereby preserving the fuel cladding integrity. Therefore, the proposed TS change will not involve a significant reduction in [a] margin of safety previously approved by the NRC. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:* Mr. Brad Fewell, Assistant General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 19348. *NRC Branch Chief:* Darrell J. Roberts. Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania *Date of amendment request:* December 21, 2005. *Description of amendment request:* The proposed amendment revises the Technical Specifications by relocating the Pressure Isolation Valve

(PIV)tables to the Technical Requirements Manual (TRM). *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed relocation of Technical Specification Table 3.4.3.2-1 does not alter the requirements for pressure isolation valve operability or surveillance currently in the Technical Specifications. The proposed change to remove the pressure isolation valve table from TS and relocate the information to an administratively controlled document, and to revise the wording in TS to reflect this change, will have no impact on any safety related structures, systems or components. The probability of occurrence of a previously evaluated accident is not increased because this change does not introduce any new potential accident initiating conditions. The consequences of accidents previously evaluated in the UFSAR [Updated Final Safety Analysis Report] are not affected because the ability of the PIVs to limit leakage through these valves in amounts that do not compromise safety is not affected. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes are administrative in nature and do not result in physical alterations or changes in the method by which any safety related system performs its intended function(s). The proposed changes do not impact any safety analysis assumptions. The proposed changes do not create any new accident initiators or involve an activity that could be an initiator of an accident of a different type. All PIVs and alarm instrumentation will continue to be tested to the same rigorous requirements as defined in the Technical Specification Surveillance Requirements. The proposed revision does not make changes in any method of testing or how any safety related system performs its safety functions. Therefore, the possibility of an accident of a different type than any previously evaluated in the UFSAR is not created. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The administrative change to relocate Technical Specification Table 3.4.3.2-1 to the Technical Requirements Manual does not alter the basic regulatory requirement for Reactor Coolant System pressure isolation and will not affect the isolation capability for credible accident scenarios. Future revisions to the Technical Requirements Manual Table will be subject to evaluation pursuant to 10 CFR 50.59. Additionally, the proposed relocation does not alter the requirements for pressure isolation valve and alarm instrumentation operability currently in the Technical Specifications. The LCO [limiting condition for operation] and Surveillance Requirements will be retained in the revised Technical Specifications. The proposed change will not affect the meaning, application, and function of the current Technical Specification requirements for the valves in Table 3.4.3.2-1. Therefore, the proposed changes do not result in a significant reduction in [a] margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:* Mr. Brad Fewell, Assistant General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 19348. *NRC Branch Chief:* Darrell J. Roberts. Exelon Generation Company, LLC (EGC, licensee), Docket No. 50-265, Quad Cities Nuclear Power Station (QCNPS), Unit 2, Rock Island County, Illinois *Date of amendment request:* December 15, 2005. *Description of amendment request:* The proposed change revises the values of the safety limit minimum critical power ratio (SLMCPR) in Technical Specification

(TS)section 2.1.1, “Reactor Core SLs.” Specifically, the proposed change would require that for Unit 2, the minimum critical power ratio

(MCPR)for Global Nuclear Fuel

(GNF)fuel shall be ≥1.09 for two recirculation loop operation, or ≥1.10 for single recirculation loop operation. Additionally, the proposed change would require that MCPR for Westinghouse fuel shall be ≥1.11 for two recirculation loop operation, or ≥1.13 for single recirculation loop operation. *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: According to 10 CFR 50.92, “Issuance of amendment,” paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

(1)Involve a significant increase in the probability or consequences of an accident previously evaluated; or

(2)Create the possibility of a new or different kind of accident from any accident previously evaluated; or

(3)Involve a significant reduction in a margin of safety. EGC has evaluated the proposed change to the TS for QCNPS, Unit 2, using the criteria in 10 CFR 50.92, and has determined that the proposed change does not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration. 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences. Limits have been established consistent with NRC-approved methods to ensure that fuel performance during normal, transient, and accident conditions is acceptable. The proposed change conservatively establishes the SLMCPR for QCNPS, Unit 2, Cycle 19 such that the fuel is protected during normal operation and during plant transients or anticipated operational occurrences (AOOs). Changing the SLMCPR does not increase the probability of an evaluated accident. The change does not require any physical plant modifications, physically affect any plant components, or entail changes in plant operation. Therefore, no individual precursors of an accident are affected. The proposed change revises the SLMCPR to protect the fuel during normal operation as well as during plant transients or AOOs. Operational limits will be established based on the proposed SLMCPR to ensure that the SLMCPR is not violated. This will ensure that the fuel design safety criterion (i.e., that at least 99.9% of the fuel rods do not experience transition boiling during normal operation and AOOs) is met. Since the proposed change does not affect operability of plant systems designed to mitigate any consequences of accidents, the consequences of an accident previously evaluated are not expected to increase. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. Creation of the possibility of a new or different kind of accident would require creating one or more new accident precursors. New accident precursors may be created by modifications of plant configuration, including changes in allowable modes of operation. The proposed change does not involve any plant configuration modifications or changes to allowable modes of operation. The proposed change to the SLMCPR assures that safety criteria are maintained for QCNPS, Unit 2, Cycle 19. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The SLMCPR provides a margin of safety by ensuring that at least 99.9% of the fuel rods do not experience transition boiling during normal operation and AOOs if the MCPR limit is not violated. The proposed change will ensure the appropriate level of fuel protection by continuing to ensure that at least 99.9% of the fuel rods do not experience transition boiling during normal operation and AOOs if the MCPR limit is not violated. Additionally, operational limits will be established based on the proposed SLMCPR to ensure that the SLMCPR is not violated. This will ensure that the fuel design safety criteria (i.e., that no more than 0.1% of the rods are expected to be in boiling transition if the MCPR limit is not violated) are met. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based upon the above, EGC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the requested amendments involve no significant hazards consideration. *Attorney for licensee:* Mr. Brad Fewell, Assistant General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. *NRC Acting Branch Chief:* Mindy S. Landau. First Energy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant, Unit 1 (PNPP), Lake County, Ohio *Date of amendment request:* November 21, 2005. *Description of amendment request:* The proposed amendment would revise the acceptance criteria of Technical Specification

(TS)Surveillance Requirements

(SRs)associated with TS 3.8.1, “AC Sources—Operating,” to modify the Emergency Diesel Generator

(EDG)start tests to provide minimum voltage and frequency limits and clarify other limits as steady state parameters. Specifically, the amendment would revise SRs 3.8.1.2, 3.8.1.7, 3.8.1.12, 3.8.1.15 and 3.8.1.20. This change is consistent with the approved Technical Specification Task Force Traveler

(TSTF)163, Revision 2. *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration which is presented below: 1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change is a LAR (license amendment request) that modifies the acceptance criteria for the PNPP TS SRs pertaining to the EDGs. The EDGs mitigate the consequences of previously evaluated accidents involving a loss of offsite power. The EDGs are used to support mitigation of the consequences of an accident, but they are not considered as the initiator of any previously analyzed accident. The proposed LAR does not change the manner in which the EDGs are operated and when implemented will continue to ensure the EDGs perform their function when called upon. The proposed revision to the TS SRs will continue to ensure that minimum frequency and voltage are attained within the required time. The SRs will continue to ensure that proper steady state voltage and frequency are attained consistent with proper EDG governor and voltage regulator performance. The proposed LAR does not affect the design of the EDGs, the operational characteristics of the EDGs, the interfaces between the EDGs and other plant systems, the function, or reliability of the EDGs. Thus, the EDGs will be capable of performing their accident mitigation function and there is no impact to the radiological consequences of any accident analysis. As such, the proposed change continues to provide adequate assurance of operable EDGs and does not involve any increase to the probability or consequences of an accident previously evaluated. 2. The proposed change would not create the possibility of a new or different kind of accident from any previously evaluated. The proposed LAR introduces no new mode of plant operation and it does not involve physical modification to the plant. New equipment is not installed with the proposed LAR, nor does the proposed LAR cause existing equipment to be operated in a new or different manner. Since the proposed changes do not involve a change to the plant design or operation, no new system interactions are created by this change. The proposed LAR does not produce any parameters or conditions that could contribute to the initiation of accidents different from those already evaluated in the Updated Safety Analysis Report. The changes to the affected TS SRs do not affect the assumed accident performance of the EDGs, nor any plant structure, system or component previously evaluated. Therefore, the proposed LAR does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. The proposed change will not involve a significant reduction in the margin of safety. The proposed change is a LAR that does not impact EDG performance, including the capability for each EDG to attain and maintain required voltage and frequency for accepting and supporting plant safety loads within the required time, as assumed in the plant safety analysis. The proposed LAR does not involve a significant reduction in a margin of safety since the operability of the EDGs continues to be determined as required to support the capability of the EDGs to provide emergency power to plant equipment that mitigate the consequences of an accident. The proposed LAR does not introduce changes to setpoints or limits established or assumed by the accident analysis. Therefore, implementation of the proposed LAR does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:* David W. Jenkins, Attorney, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308. *NRC Branch Chief:* Mindy Landau, Acting. FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire *Date of amendment request:* December 6, 2005. *Description of amendment request:* The proposed amendment would revise the Seabrook Station, Unit No. 1 Technical Specification 3.8.3.1, “Onsite Power Distribution,” to extend the allowed outage time for balance-of-plant vital inverters 1-EDE-I-1E and 1-EDE-I-1F from 24 hours to 7 days. *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change extends the allowed outage time

(AOT)for the balance-of-plant

(BOP)instrument bus inverters from 24 hours to 7 days. The BOP instrument bus inverters do not solely support any risk-significant functions. The failure of an inverter is not an initiator of any analyzed event and does not increase the frequency of an initiating event. Consequently, extending the AOT will not have an impact on the frequency of occurrence of any event previously analyzed. The proposed change does not alter the design, configuration, operation, or function of any plant system, structure, or component. As a result, the outcomes of previously evaluated accidents are unaffected. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. The proposed change does not challenge the performance or integrity of any safety-related system. The proposed change neither installs nor removes any plant equipment, not alters the design, physical configuration, or mode of operation of any plant structure, system, or component. Installed equipment will not be operated in a new or different manner. No physical changes are being made to the plant, so no new accident causal mechanisms are being introduced. Procedures that ensure the unit operates within analyzed limits and procedures that respond to off-normal and emergency conditions are not altered with this proposed change. Therefore, the proposed change does not create the possibility of a new or different accident from any previously evaluated. 3. The proposed changes do not involve a significant reduction in [a] margin of safety. The margin of safety associated with the acceptance criteria of any accident is unchanged. The proposed change does not alter the design, configuration, operation, or function of any plant system, structure, or component. The ability of any operable structure, system, or component to perform its designated safety function is unaffected by this change. Operation with one instrument bus inverter inoperable and the associated instrument bus aligned to its maintenance supply does not result in a significant reduction in [a] margin of safety. Surveillance testing of the emergency diesel generators

(EDGs)and the electrical distribution system provides confidence that the EDGs will energize the emergency AC buses following a loss of power. Therefore, the proposed change does not involve a significant reduction in [a] margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:* M. S. Ross, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420. *NRC Branch Chief:* Darrell J. Roberts. Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin *Date of amendment request:* November 12, 2004. *Description of amendment request:* The proposed amendment would revise Technical Specification

(TS)5.5.7, “Inservice Testing Program,” and TS 5.5.8, “Steam Generator

(SG)Tube Surveillance Program,” to update references to the American Society of Mechanical Engineers

(ASME)*Boiler and Pressure Vessel Code*

(Code)and certain associated periodicities for inservice testing activities consistent with the requirements of Title 10 of the *Code of Federal Regulations* (10 CFR) section 50.55a, “Codes and standards.” The proposed amendment would also correct a typographical error contained in TS 5.5.8.b.2. *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration which is presented below: 1. Operation of the Point Beach Nuclear Plant in accordance with the proposed amendments does not result in a significant increase in the probability or consequences of any accident previously evaluated. The proposed change revises Technical Specifications for consistency with the requirements of 10 CFR 50.55a(f)(4) and 10 CFR 50.55a(g)(4). The proposed change incorporates revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves. The proposed change does not involve any hardware changes, nor does it affect the probability of any event initiators. There will be no change to normal plant operating parameters, engineered safety feature actuation setpoints, accident mitigation capabilities, or accident analysis assumptions or inputs. Therefore, the probability or consequences of any accident previously evaluated will not be significantly increased as a result of the proposed change. 2. Operation of the Point Beach Nuclear Plant in accordance with the proposed amendments does not result in a new or different kind of accident from any accident previously evaluated. The proposed change incorporates revisions to the ASME Code that result in a net improvement in the measures for testing. The proposed change does not involve a modification to the physical configuration of the plant (i.e., no new equipment will be installed) or change in the methods governing normal plant operation. The proposed change will not impose any new or different requirements or introduce a new accident initiator, accident precursor, or malfunction mechanism. Additionally, there is no change in the types or increases in the amounts of any effluent that may be released off-site and there is no increase in individual or cumulative occupational exposure. Equipment important to safety will continue to operate as designed. The changes do not result in any event previously deemed incredible been made credible. The changes do not result in adverse conditions or result in any increase in the challenges to safety systems. Therefore, operation of the Point Beach Nuclear Plant in accordance with the proposed amendment will not create the possibility of a new or different type of accident from any accident previously evaluated. 3. Operation of the Point Beach Nuclear Plant in accordance with the proposed amendments does not result in a significant reduction in a margin of safety. The proposed change incorporates revisions to the ASME Code that result in a net improvement in the measures for testing. The safety function of the affected components will be maintained. There are no new or significant changes to the initial conditions contributing to accident severity or consequences. The proposed amendment will not otherwise affect the plant protective boundaries, will not cause a release of fission products to the public, nor will it degrade the performance of any other structures, systems or components

(SSCs)important to safety. Therefore, the requested change will not result in a significant reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:* Jonathan Rogoff, Esquire, Vice President, Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, Hudson, WI 54016. *NRC Branch Chief:* L. Raghavan. PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey *Date of amendment request:* October 11, 2005. *Description of amendment request:* The proposed amendment would revise certain 18-month Technical Specification

(TS)Surveillance Requirements

(SRs)to eliminate the condition that testing be conducted during shutdown. *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes permit PSEG to evaluate the conditions required to safely perform a TS SR. These surveillance tests verify that equipment will perform its intended safety function of mitigating an accident. No analyzed accident scenario is being revised. The initiating conditions and assumptions for accidents described in the Hope Creek Generating Station Updated Final Safety Analysis Report (UFSAR) remain as previously analyzed. The proposed changes do not reduce the ability of the mitigating equipment to perform its safety function. The TS will continue to require the surveillance tests to be performed on an eighteen-month periodicity to verify operability. As a result, the ability of the mitigating equipment to perform its safety function is unaffected by the proposed change. The capitalization change is proposed to improve readability and does not alter any requirement. Based upon the above, the proposed changes will not involve a significant increase in the probability or consequences of an accident previously analyzed. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated in the UFSAR. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. Specifically, no new hardware is being added to the plant as part of the proposed change, no existing equipment is being modified, and no significant changes in operations are being introduced. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed changes will not alter any assumptions, initial conditions, or results of any accident analyses. The proposed changes to remove the requirement to perform certain testing during shutdown conditions allows PSEG to evaluate the conditions needed to safely perform the required testing. There is no change to the frequency of testing or in the testing that is required. There is no change in the responsibility of PSEG to perform tests in a safe and responsible manner. Any changes to procedures will have to be individually evaluated to ensure that they do not reduce the margin of safety. The changes do not affect the ability of systems, structures or components to perform their safety related functions. In addition, the proposed changes do not affect the ability of the safety systems to ensure that the facility can be maintained in a shutdown or refueling condition for extended periods of time. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:* Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038. *NRC Branch Chief:* Darrell J. Roberts. PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey *Date of amendment request:* August 31, 2005; as supplemented December 8, 2005. *Description of amendment request:* The proposed amendment would relocate the containment high range accident monitors from the radiation monitoring instrumentation technical specification

(TS)to the accident monitoring TS and correct a typographical error contained in a previous amendment. *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change presents no change in the probability of a previously evaluated accident. The proposed change presents no change in the consequence of an accident, since the containment high range accident monitors are used post-accident to determine the amount of core damage and status of the fission product barriers. The containment high range accident monitors are used post accident to assess the conditions inside containment. They have an automatic function to switch the subcooling margin monitor

(SCMM)to “adverse” mode (i.e., it displays a more conservative indication of the amount of subcooling in the RCS) [reactor coolant system]. Additionally, the containment high range accident monitors provide an indication that is used post accident in determining the status of the fission product barriers. There will be no change in the operation or use of the containment high range accident monitors. The remaining change is editorial in nature and does not impact the accident analysis in any manner. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated. Response: No. The proposed change is a minor change that is administrative in nature. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. No new hardware is added, existing hardware is not modified and no significant changes in operations are implemented. Post accident monitoring instrumentation is not associated with the initiation of an accident. 3. Does the proposed change involve a significant reduction in [a] margin of safety? Response: No. The proposed change does not alter the manner in which safety limits, limiting safety systems settings or limiting conditions for operation are determined. The proposed change will not alter any assumptions, initial conditions or results specified in any accident analysis. There is no change in the containment high range accident monitor high level alarm setpoint. The ECS [electronic check source] is functionally equivalent to the TS definition of SOURCE CHECK. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:* Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038. *NRC Branch Chief:* Darrell J. Roberts. PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station, Unit No. 2, Salem County, New Jersey *Date of amendment request:* September 21, 2005. *Description of amendment request:* The amendment would change the scope of steam generator

(SG)tube inspections required in the SG tubesheet region. *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. Of the various accidents previously evaluated, the proposed changes only affect the steam generator tube rupture

(SGTR)event evaluation and the postulated steam line break

(SLB)accident evaluation. Loss-of-coolant accident

(LOCA)conditions cause a compressive axial load to act on the tube. Therefore, since the LOCA tends to force the tube into the tubesheet rather than pull it out, it is not a factor in this amendment request. Another faulted load consideration is a safe shutdown earthquake (SSE); however, the seismic analysis of Westinghouse 51 Series SGs has shown that axial loading of the tubes is negligible during an SSE. PSEG's amendment request takes credit for how the tubesheet enhances the tube integrity in the Westinghouse Electric Company explosive tube expansion (WEXTEX) region by precluding tube deformation beyond its initial expanded outside diameter. For the SGTR and SLB events, the required structural margins of the SG tubes will be maintained due to the presence of the tubesheet. Tube rupture is precluded for axial cracks in the WEXTEX region due to the constraint provided by the tubesheet. Therefore, the normal operating 3 <sup>Δ</sup> P margin and the postulated accident 1.43 <sup>Δ</sup> P margin against burst are maintained. The W* length supplies the necessary resistive force to preclude pullout loads under both normal operating and accident conditions. The contact pressure results from the WEXTEX expansion process, thermal expansion mismatch between the tube and tubesheet, and from the differential pressure between the primary and secondary side. Therefore, the proposed change results in no significant increase in the probability or the occurrence of an SGTR or SLB accident. The proposed changes do not affect other systems, structures, components or operational features. Therefore, based on the above evaluation, the proposed changes do not involve a significant increase in the probability of an accident previously evaluated. The consequences of an SGTR event are primarily affected by the primary-to-secondary flow rate and the time duration of the primary-to-secondary flow during the event. Primary-to-secondary flow rate through a postulated ruptured tube (i.e., complete severance of a single SG tube) is not affected by the proposed change since the flow rate is based on the inside diameter of a[n] SG tube and the pressure differential. PSEG's amendment request does not change either of these. The duration of primary-to-secondary leakage is based on the time required for an operator to determine that a[n] SGTR has occurred, the time to identify and isolate the faulted SG, and ensure termination of radioactive release to the atmosphere from the faulted SG. PSEG's amendment request does not affect the duration of the primary-to-secondary leakage because it does not change the control room indicators with which an operator would determine that an SGTR has occurred. The consequences of an SGTR are secondarily affected by primary-to-secondary leakage, which could occur due to axial cracks remaining in service in the WEXTEX region in a non-faulted SG. During a[n] SGTR, the primary-to-secondary differential pressure is less than or equal to the normal operating differential pressure; therefore, the primary-to-secondary leakage due to axial cracks in the WEXTEX region of a non-faulted SG during a[n] SGTR would be less than or equal to the primary-to-secondary leakage experienced during normal operation. Primary-to-secondary leakage is considered in the calculation determining the consequences of a[n] SGTR and the value is bounding. The postulated SLB has the greatest primary-to-secondary pressure differential, and therefore could experience the greatest primary-to-secondary leakage. PSEG's amendment request requires the aggregate leakage, (i.e., the combined leakage for the tubes with service induced degradation inside the tubesheet) to remain below the maximum allowable SLB primary-to-secondary leakage rate limit such that the doses are maintained to less than the 10 CFR [Part] 100 limits and also less than the GDC-[General Design Criterion]19 limits. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. PSEG's amendment request does not introduce any physical changes to the Salem Unit 2 SGs. PSEG's amendment request takes credit for how the tubesheet enhances the SG tube integrity in the WEXTEX region. Because degradation detected within the W* distance are required to be plugged, it is highly unlikely that a tube would fail as a result of a circumferential defect. Therefore a tube severance, which would strike neighboring tubes and create a multiple tube rupture, is not credible. The proposed change does not introduce any new equipment or any change to existing equipment. No new effects on existing equipment are created. Based on the above evaluation, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The amendment request maintains the structural margins of the SG tubes for both normal and accident conditions that are required by Regulatory Guide 1.121. For cracking located within the tubesheet, tube burst is precluded due to the presence of the tubesheet. WCAP-14797, Revision 2 defines a length W* of degradation free expanded tubing, that provides the necessary resistance to tube pullout due to the pressure induced forces (with applicable safety factor applied). Application of the W* methodology will preclude unacceptable primary-to-secondary leakage during all plant conditions. The methodology for determining leakage provides for large margins between calculated and actual leakage values in the W* criteria. Based on the above, it is concluded that the proposed changes do not result in a significant reduction of margin with respect to plant safety as defined in the Updated Final Analysis Report or Technical Specifications. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:* Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038. *NRC Branch Chief:* Darrell J. Roberts. PPL Susquehanna, LLC, Docket No. 50-387, Susquehanna Steam Electric Station, Unit 1 (SSES 1), Luzerne County, Pennsylvania *Date of amendment request:* December 1, 2005. *Description of amendment request:* The proposed amendment would change the SSES-1 Technical Specifications

(TSs)by revising the Unit 1 Cycle 15 (U1C15) minimum critical power ratio

(MCPR)safety limit for single loop operation in section 2.1.1.2 and references listed in TS 5.6.5.b. *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated? Response: No. The proposed change to the single-loop MCPR Safety Limit does not directly or indirectly affect any plant system, equipment, component, or change the processes used to operate the plant. Further, the proposed U1C15 MCPR Safety Limit was generated using NRC approved methodology and meets the applicable acceptance criteria. Thus, this proposed amendment does not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated. Prior to the startup of U1C15, licensing analyses are performed (using NRC approved methodology referenced in Technical Specification Section 5.6.5.b) to determine changes in the critical power ratio as a result of anticipated operational occurrences. These results are added to the MCPR Safety Limit values to generate the MCPR operating limits in the U1C15 COLR [core operating limits report]. These limits could be different from those specified for the current Unit 1 COLR. The COLR operating limits thus assure that the MCPR Safety Limit will not be exceeded during normal operation or anticipated operational occurrences. Postulated accidents are also analyzed prior to the startup of U1C15 and the results shown to be within the NRC approved criteria. The changes to the references in Section 5.6.5.b were made to properly reflect the NRC approved methodology used to generate the U1C15 core operating limits. The use of this approved methodology does not increase the probability of occurrence or consequences of an accident previously evaluated. Therefore, the proposed amendment does not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The change to the single-loop MCPR Safety Limit does not directly or indirectly affect any plant system, equipment, or component and therefore does not affect the failure modes of any of these items. Thus, the proposed change does not create the possibility of a previously unevaluated operator error or a new single failure. The changes to the references in Section 5.6.5.b were made to properly reflect the NRC approved methodology used to generate the U1C15 core operating limits. The use of this approved methodology does not create the possibility of a new or different kind of accident. Therefore, this proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. Since the proposed changes do not alter any plant system, equipment, component, or the processes used to operate the plant, the proposed change will not jeopardize or degrade the function or operation of any plant system or component governed by Technical Specifications. The proposed single-loop MCPR Safety Limit does not involve a significant reduction in the margin of safety as currently defined in the Bases of the applicable Technical Specification sections, because the MCPR Safety Limits calculated for U1C15 preserve the required margin of safety. The changes to the references in section 5.6.5.b were made to properly reflect the NRC approved methodology used to generate the U1C15 core operating limits. This approved methodology is used to demonstrate that all applicable criteria are met, thus, demonstrating that there is no reduction in the margin of safety. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:* Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101-1179. *NRC Branch Chief:* Richard J. Laufer. PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania *Date of amendment request:* October 5, 2005. *Description of amendment request:* The proposed amendment would revise the SSES 1 and 2 Technical Specifications

(TSs)3.4.10, “RCS [reactor coolant system] Pressure and Temperature (P/T) Limits,” to remove valid P/T curve limit date and replacing it with the effective full-power years

(EFPY)of radiation exposure on each of the P/T limit curves for SSES 1 and 2. The new P/T limit would be 35.7 EFPY for SSES 1 and 30.2 EFPY for SSES 2. *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? No. The proposed changes request that the P/T limits curves in TS 3.4.10, “RCS Pressure and Temperature (P/T) Limits” be revised by removing the valid date and replacing it with the Effective Full Power Years of radiation exposure limit on each of the P/T curves for SSES Units 1 and 2. The P/T limits are prescribed during all operational conditions to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate, resulting in nonductile failure of the reactor coolant pressure boundary, an unanalyzed condition. Therefore, the proposed changes do not have any effect on the probability of an accident previously evaluated. The P/T curves are used as operational limits during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region. The P/T curves provide assurance that station operation is consistent with previously evaluated accidents. Thus, the radiological consequences of an accident previously evaluated are not increased. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? No. The proposed changes do not change the response of any plant equipment to transient conditions. The proposed changes do not introduce any new equipment, modes of system operation, or failure mechanisms. Therefore, there are no new types of failures or new or different kinds of accidents or transients that could be created by these changes. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? No. The consequences of a previously evaluated accident are not increased by these proposed changes, since the Loss of Coolant Accident analyzed in the FSAR [Final Safety Analysis Report] assumes a complete break of the reactor coolant pressure boundary. The changes to the P/T limits curves do not change this assumption. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:* Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101-1179. *NRC Branch Chief:* Richard J. Laufer. PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania *Date of amendment request:* November 9, 2004, as supplemented December 15, 2005. This notice supersedes the original notice published on April 26, 2005 (70 FR 21463), which was based upon the licensee's application dated November 9, 2004. *Description of amendment request:* The proposed amendments would change the SSES 1 and 2 Technical Specifications

(TSs)3.8.4, “DC Sources— Operating,” 3.8.5, “DC Sources—Shutdown,” 3.8.6, “Battery Cell Parameters,” and add a new TS section, 5.5.13, “Battery Monitoring and Maintenance Program.” These changes are consistent with Technical Specification Change Traveler

(TSTF)360, Revision 1. *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? No. The proposed changes restructure the Technical Specifications

(TSs)for the DC Electrical Power Systems. The proposed changes consist of the relocation of several surveillance requirements that perform preventive maintenance on the safety related batteries, to a new license controlled program. The DC electrical power systems, including associated battery chargers, are not initiators to any accident sequence analyzed in the Final Safety Analysis Report (FSAR). Operation in accordance with the proposed TS ensures that the DC electrical power systems are capable of performing functions as described in the FSAR. Therefore, the mitigative functions supported by the DC Power Systems will continue to provide the protection assumed by the analysis. The relocation of preventive maintenance surveillance, and certain operating limits and actions to a newly created, licensee-controlled TS 5.5.13, “Battery Monitoring and Maintenance Program,” will not challenge the ability of the DC electrical power systems to perform their design functions. The maintenance and monitoring required by current TS, which are based on industry standards, will continue to be performed. In addition, the DC Power Systems are within the scope of 10 CFR 50.65, “Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants,” which will ensure the control of maintenance activities associated with the DC electrical power systems. The integrity of fission product barriers, plant configuration, and operating procedures as described in the FSAR will not be affected by the proposed changes. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? No. The proposed changes involve restructuring the TS for the DC electrical power systems. These changes will rely on a new license controlled program to monitor battery parameters for operability. The DC electrical power systems, which include the associated battery chargers, are not initiators to any accident sequence analyzed in the FSAR. Rather, the DC electrical power systems are used to supply equipment used to mitigate an accident. These mitigative functions, supported by the DC electrical power systems are not affected by these changes and they will continue to provide the protection assumed by the safety analysis described in the FSAR. There are no new types of failures or new or different kinds of accidents or transients that could be created by these changes. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? No. The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed changes will not adversely affect operation of plant equipment. These changes will not result in a change to the setpoints at which protective actions are initiated. Sufficient DC electrical system capacity is ensured to support operation of mitigation equipment. The changes associated with the new Battery Maintenance and Monitoring Program will ensure that the station batteries are maintained in a highly reliable state. The equipment fed by the DC electrical sources will continue to provide adequate power to safety related loads in accordance with analysis assumptions. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:* Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101-1179. Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri *Date of application request:* October 26, 2005. *Description of amendment request:* The amendment would revise Technical Specification

(TS)3.6.6, “Containment Spray and Cooling Systems,” to change Required Action D.1 that currently allows 72 hours of operation with both containment cooling trains out of service as long as both containment spray trains are operable. The required action would be revised to impose the more stringent requirement of requiring plant shutdown if both containment cooling trains are out of service instead of allowing the 72 hours to restore an inoperable train. There are also changes to other required actions in TS 3.6.6 to reflect the revision to Required Action D.1. In addition, the required action for two inoperable containment spray trains is being revised. *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

(1)Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No The proposed change in the required action when two containment cooling trains are inoperable to require plant shutdown is more restrictive than the current required action that allows 72 hours of operation [to restore one containment cooling train to operable status]. Also the proposed change to the required action [F.1 for] when two containment cooling trains are inoperable to be in MODE 3 within 6 hours and MODE 5 within 36 hours [are the same as in the current Required Actions E.1 and E.2 for when the two containment cooling trains are inoperable. The proposed change to the required action for two containment spray trains being inoperable] is more restrictive than the current required action to enter LCO [Limiting Condition for Operation] 3.0.3 immediately [because] LCO 3.0.3 requires the plant to be in MODE 3 within 7 hours. The more stringent requirements are imposed to ensure process variables, structures, systems and components are maintained consistently with the safety analysis and licensing basis [for Callaway]. All of these proposed changes have been reviewed to ensure no previously evaluated accident has been adversely affected. [The proposed changes are not accident initiators.] Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

(2)Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in controlling [plant] parameters. The proposed change does impose different requirements. However, these changes are consistent with [the] assumptions made in the safety analysis and licensing basis [for Callaway]. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3)Does the proposed change involve a significant reduction in a margin of safety? Response: No The imposition of more stringent requirements has no impact on or will increase the margin of safety. The change in the required action when two containment cooling trains are out of service will increase the margin of safety by decreasing the allowed restoration time [to restore an inoperable containment cooling train to operable status]. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:* John O'Neill, Esq., Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037. *NRC Branch Chief:* David Terao. Notice of Issuance of Amendments to Facility Operating Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment. Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for A Hearing in connection with these actions was published in the **Federal Register** as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see

(1)the applications for amendment,

(2)the amendment, and

(3)the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, *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 PDR Reference staff at 1

(800)397-4209,

(301)415-4737 or by e-mail to *pdr@nrc.gov.* AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek Nuclear Generating Station (OCNGS), Ocean County, New Jersey *Date of application for amendment:* December 17, 2004. *Brief description of amendment:* The amendment revised Appendix B, Environmental Technical Specifications, of the OCNGS Facility Operating License, principally by deleting redundant reporting requirements, aligning various requirements with regulations and accepted guidance documents, and correcting administrative errors. *Date of Issuance:* January 4, 2006. *Effective date:* As of the date of issuance and shall be implemented within 60 days of issuance. *Amendment No.:* 257. *Facility Operating License No. DPR-16:* The amendment revised the Environmental Technical Specifications. *Date of initial notice in* Federal Register: April 12, 2005 (70 FR 19113). The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated January 4, 2006. No significant hazards consideration comments received: No. Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power Station, Unit No. 2, New London County, Connecticut *Date of application for amendment:* February 25, 2005, as supplemented by letter dated August 4, 2005. *Brief description of amendment:* The amendment revised the Millstone Power Station, Unit No. 2, Technical Specifications Surveillance Requirement for trisodium phosphate to remove the granularity term and chemical detail. In addition, the proposed change will increase the allowed outage time from 48 to 72 hours. *Date of issuance:* January 3, 2006. *Effective date:* As of the date of issuance and shall be implemented within 60 days from the date of issuance. *Amendment No.:* 290. *Facility Operating License No. DPR-65:* The amendment revised the Technical Specifications. *Date of initial notice in* Federal Register: July 19, 2005 (70 FR 41444). The additional information provided in the supplemental letter dated August 4, 2005, did not expand the scope of the application as noticed and did not change the NRC staff's original proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 3, 2006. No significant hazards consideration comments received: No. Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423, Millstone Power Station, Unit No. 3, New London County, Connecticut *Date of application for amendment:* December 16, 2004, as supplemented on October 5, 2005. *Brief description of amendment:* The amendment revised the current fuel rod average licensing basis burnup limit for one lead test assembly containing advanced zirconium based alloys to a limit not exceeding 71,000 megawatt-days per metric ton of uranium. *Date of issuance:* December 30, 2005. *Effective date:* As of the date of issuance and shall be implemented within 60 days from the date of issuance. *Amendment No.:* 228 *Facility Operating License No. NPF-49:* The amendment revised the design basis. *Date of initial notice in* Federal Register: February 1, 2005 (70 FR 5238). The October 5, 2005, supplement provided clarifying information and did not change the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 30, 2005. No significant hazards consideration comments received: No. Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana *Date of amendment request:* July 20, 2005, as supplemented by letter dated September 14, 2005. *Brief description of amendment:* The amendment approves the transfer of Facility Operating License and Materials License No. NPF-38, held by Entergy Louisiana, Inc.

(ELI)and Entergy Operatings, Inc. (EOI), for the Waterford Steam Electric Station, Unit 3 (Waterford 3). The transfer is associated with the restructuring of ELI from a Louisiana corporation to a Texas limited liability company, Entergy Louisiana, LLC (ELL). EOI will continue to operate Waterford 3, and the restructuring will not affect the technical or financial qualifications of ELL or EOI. *Date of issuance:* December 31, 2005. *Effective date:* At the time the transfer is completed. *Amendment No.:* 203. *Facility Operating License No. NPF-38:* The amendment revised the Facility Operating License and Materials License. *Date of initial notice in* Federal Register: October 17, 2005 (70 FR 60374). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 2, 2005. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania *Date of application for amendments:* December 17, 2004. *Brief description of amendments:* The amendments revised the Appendix B, Environmental Technical Specifications. *Date of issuance:* January 3, 2006. *Effective date:* As of the date of issuance, to be implemented within 60 days. *Amendments Nos.:* 257 and 260. *Renewed Facility Operating License Nos. DPR-44 and DPR-56:* The amendments revised the Environmental Technical Specifications. *Date of initial notice in* Federal Register: April 12, 2005 (70 FR 19112). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 3, 2006. No significant hazards consideration comments received: No. FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire *Date of amendment request:* March 28, 2005, as supplemented September 23, 2005. *Description of amendment request:* The amendment extended the expiration of Facility Operating License

(FOL)NPF-86 for Seabrook Station, Unit No. 1, by approximately 3.4 years. The extension sets the date of expiration of the FOL to occur 40 years from the date of issuance of the full-power operating license. Specifically, the FOL, with a previous expiration date of October 17, 2026, now expires March 15, 2030. This change allows the recapture of zero-power and low-power testing time in accordance with SECY-98-296, “Agency Policy Regarding Licensee Recapture of Low-Power Testing or Shutdown Time for Nuclear Power Plants,” dated December 21, 1998. *Date of issuance:* December 28, 2005. *Effective date:* As of its date of issuance, and shall be implemented within 30 days. *Amendment No.:* 105. *Facility Operating License No. NPF-86:* The amendment revised the License. *Date of initial notice in* Federal Register: May 24, 2005 (70 FR 29797). The licensee's September 23, 2005 supplement provided clarifying information that did not change the scope of the proposed amendment as described in the original notice of proposed action published in the **Federal Register** , and did not change the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 28, 2005. No significant hazards consideration comments received: No. Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie Plant, Unit No. 2, St. Lucie County, Florida *Date of application for amendment:* March 31, 2005, as supplemented November 9, 2005. *Brief description of amendment:* This amendment extended the date for the next Appendix J, Type A test at St. Lucie Unit 2 until the end of the SL2-17 refueling outage. *Date of Issuance:* December 23, 2005. *Effective Date:* As of the date of issuance and shall be implemented within 60 days. *Amendment No.:* 140. *Renewed Facility Operating License No. NPF-16:* Amendment revised the Technical Specifications. *Date of initial notice in* Federal Register: June 7, 2005 (70 FR 33215). The November 9, 2005, supplement did not affect the original proposed no significant hazards determination, or expand the scope of the request as noticed in the **Federal Register** . The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 23, 2005. No significant hazards consideration comments received: No. Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, California *Date of application for amendments:* October 29, 2004, as supplemented by letters dated May 6 and October 31, 2005. *Brief description of amendments:* The amendments revised the Technical Specification

(TS)requirements for the handling of irradiated fuel in the containment and fuel building, and certain specifications related to performing core alterations. These changes are based on analysis of the postulated fuel handling and core alteration accidents and transients for Diablo Canyon Nuclear Power Plant, Units 1 and 2. The amendments are consistent with the NRC-approved Industry/Technical Specification Task Force

(TSTF)Standard Technical Specifications Change Traveler, TSTF-51, Revision 2, “Revise containment requirements during handling irradiated fuel and core alterations.” In addition, the amendments made editorial corrections to TS 3.1.7, “Rod Position Indication,” TS 3.3.1, “Reactor Trip System

(RTS)Instrumentation,” TS 3.4.16, “RCS Specific Activity,” TS 3.7.3, “Main Feedwater Isolation Valve (MFIVs), Main Feedwater Regulating Valves (MFRVs), MFRV Bypass Valves, and Main Feedwater Pump

(MFWP)Turbine Stop Valves,” and TS 3.7.13, “Fuel Handling Building Ventilation System (FHBVS).” *Date of issuance:* January 3, 2006. *Effective date:* January 3, 2006, and shall be implemented within 90 days of issuance. *Amendment Nos.:* Unit 1—184; Unit 2—86. *Facility Operating License Nos. DPR-80 and DPR-82:* The amendments revised the Technical Specifications. *Date of initial notice in* Federal Register: January 4, 2005 (70 FR 403) The supplements dated May 6 and October 31, 2005, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 3, 2006. No significant hazards consideration comments received: No. Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, Georgia *Date of application for amendments:* November 12, 2004, as supplemented by letters dated September 2 and September 16, 2005. *Brief description of amendments:* The amendments revised the Technical Specifications

(TS)3.1.7, “Standby Liquid Control

(SLC)System,” for Hatch, Units 1 and 2. The amendments update Figure 3.1.7-1 and 3.1.7-2 of the Units 1 and 2 TS to reflect the increased concentration of Boron-10 in the solution. Conforming revisions to Bases B3.1.7, are also included. *Date of issuance:* January 5, 2006. *Effective date:* As of the date of issuance and shall be implemented within 30 days from the date of issuance. *Amendment Nos.:* 247/191. *Renewed Facility Operating License Nos. DPR-57 and NPF-5:* Amendments revised the Technical Specifications. *Date of initial notice in* Federal Register: February 1, 2005 (70 FR 5249). The supplemental letter dated September 2, 2005, contained clarifying information only and did not change the initial proposed no significant hazards consideration determination or expand the scope of the original **Federal Register** notice. The supplemental letter dated September 16, 2005, contained information that expanded the scope of the original **Federal Register** notice. The proposed amendment was re-noticed on October 25, 2005 (70 FR 61662). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 5, 2006. No significant hazards consideration comments received: No. Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee *Date of application for amendments:* April 27, 2005, as supplemented by letter dated November 17, 2005. *Brief description of amendments:* The amendments relocate several Technical Specification

(TS)requirements to the Sequoyah Technical Requirements Manual (TRM). Specifically, the amendments relocate the provisions for TS 3.3.2 (Movable Incore Detectors), TS 3.3.3.4 (Meteorological Instrumentation), TS 3.4.7 (Reactor Coolant System Chemistry), TS 3.4.11 (Reactor Coolant System Head Vents), TS 3.7.2 (Steam Generator Pressure and Temperature Limitations), TS 3.7.10 (Sealed Source Contamination), TS 3.9.5 (Refueling Operations Communications), and TS 3.9.6 (Manipulator Crane) to the TRM. These changes are consistent with the latest version of NUREG-1431, Revision 3, “Standard Technical Specifications for Westinghouse Plants,” and do not diminish the level of safety found in the current TSs. *Date of issuance:* December 28, 2005. *Effective date:* As of the date of issuance and shall be implemented within 45 days. *Amendment Nos.:* 305, 295. *Facility Operating License Nos. DPR-77 and DPR-79:* Amendments revised the technical specifications. *Date of initial notice in* Federal Register: July 5, 2005 (70 FR 38723). The supplemental letter of November 17, 2005, provided clarifying information that did not change the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 28, 2005. No significant hazards consideration comments received: No. Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia *Date of application for amendments:* December 17, 2004. *Brief Description of amendments:* These amendments revised the reactor coolant pressure and temperature limits, low-temperature overpressure protection system (LTOPS) setpoint values, and LTOPS enable temperatures that are valid for up to 47.6 effective full-power years

(EFPY)and 48.1 EFPY of operation at Surry Power Station, Unit Nos. 1 and 2, respectively. *Date of issuance:* January 3, 2006. *Effective date:* As of the date of issuance and shall be implemented within 180 days from the date of issuance. *Amendment Nos.:* 245/244. *Renewed Facility Operating License Nos. DPR-32 and DPR-37:* Amendments change the Technical Specifications. *Date of initial notice in* Federal Register: March 1, 2005 (70 FR 9999). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 3, 2006. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 9th day of January 2006. For the Nuclear Regulatory Commission. Edwin M. Hackett, Deputy Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 06-320 Filed 1-13-06; 8:45 am]

Connectionstraces to 17

Traces to 17 documents

CFR