# Involving No Significant Hazards Considerations > Briefs, arguments, decisions, and more. URL: https://www.frixlaw.com/law-library/documents/fr%3A05-1574 ## Record - **Collection:** Federal Register - **Document type:** Notice - **Published:** February 1, 2005 - **Citation:** 70 FR 5233 ## Text 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 January 7, 2005, through January 19, 2005. The last biweekly notice was published on January 18, 2005 (70 FR 2886). 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, Docket No. 50-461, Clinton Power Station, Unit 1, DeWitt County, Illinois Date of amendment request: September 15, 2004. Description of amendment request: The proposed amendment would delete requirements from the Technical Specifications (TSs) to maintain hydrogen recombiners and hydrogen and oxygen monitors. A notice of availability for this TS improvement using the consolidated line item improvement process was published in the Federal Register on September 25, 2003 (68 FR 55416). Licensees were generally required to implement upgrades as described in NUREG-0737, “Clarification of TMI [Three Mile Island] Action Plan Requirements,” and Regulatory Guide (RG) 1.97, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.” Implementation of these upgrades was an outcome of the lessons learned from the accident that occurred at TMI, Unit 2. Requirements related to combustible gas control were imposed by order for many facilities and were added to, or included, in the TSs for nuclear power reactors currently licensed to operate. The revised Title 10 of the Code of Federal Regulations (10 CFR) Section 50.44, “Standards for combustible gas control system in light-water-cooled power reactors,” eliminated the requirements for hydrogen recombiners and relaxed safety classifications and licensee commitments to certain design and qualification criteria for hydrogen and oxygen monitors. The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on September 25, 2003 (68 FR 55416). The licensee affirmed the applicability of the model NSHC determination in its application dated September 15, 2004. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of NSHC is presented below: Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated The revised 10 CFR 50.44 no longer defines a design-basis loss-of-coolant accident (LOCA) hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The NRC has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant accident sequences that could threaten containment integrity. With the elimination of the design-basis LOCA hydrogen release, hydrogen and oxygen monitors are no longer required to mitigate design-basis accidents and, therefore, the hydrogen monitors do not meet the definition of a safety-related component as defined in 10 CFR 50.2. RG 1.97 Category 1, is intended for key variables that most directly indicate the accomplishment of a safety function for design-basis accident events. The hydrogen and oxygen monitors no longer meet the definition of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 50.44 the NRC found that Category 3, as defined in RG 1.97, is an appropriate categorization for the hydrogen monitors because the monitors are required to diagnose the course of beyond design-basis accidents. Also, as part of the rulemaking to revise 10 CFR 50.44, the NRC found that Category 2, as defined in RG 1.97, is an appropriate categorization for the oxygen monitors, because the monitors are required to verify the status of the inert containment. The regulatory requirements for the hydrogen and oxygen monitors can be relaxed without degrading the plant emergency response. The emergency response, in this sense, refers to the methodologies used in ascertaining the condition of the reactor core, mitigating the consequences of an accident, assessing and projecting offsite releases of radioactivity, and establishing protective action recommendations to be communicated to offsite authorities. Classification of the hydrogen monitors as Category 3, [classification of the oxygen monitors as Category 2,] and removal of the hydrogen and oxygen monitors from TS will not prevent an accident management strategy through the use of the severe accident management guidelines, the emergency plan, the emergency operating procedures, and site survey monitoring that support modification of emergency plan protective action recommendations. Therefore, the elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, does not involve a significant increase in the probability or the consequences of any accident previously evaluated. Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, will not result in any failure mode not previously analyzed. The hydrogen recombiner and hydrogen and oxygen monitor equipment was intended to mitigate a design-basis hydrogen release. The hydrogen recombiner and hydrogen and oxygen monitor equipment are not considered accident precursors, nor does their existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radionuclides within the containment building. Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated. Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, in light of existing plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, results in a neutral impact to the margin of safety. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The NRC has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. Category 3 hydrogen monitors are adequate to provide rapid assessment of current reactor core conditions and the direction of degradation while effectively responding to the event in order to mitigate the consequences of the accident. The intent of the requirements established as a result of the TMI, Unit 2 accident can be adequately met without reliance on safety-related hydrogen monitors. Category 2 oxygen monitors are adequate to verify the status of an inerted containment. Therefore, this change does not involve a significant reduction in the margin of safety. The intent of the requirements established as a result of the TMI, Unit 2 accident can be adequately met without reliance on safety-related oxygen monitors. Removal of hydrogen and oxygen monitoring from TS will not result in a significant reduction in their functionality, reliability, and availability. The NRC staff proposes to determine that the amendment request involves NSHC. Attorney for licensee: Mr. Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60666. NRC Section Chief: Gene Y. Suh. Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Maricopa County, Arizona Date of amendments request: December 16, 2004. Description of amendments request: The requested change will delete Technical Specification (TS) 5.6.1, “Occupational Radiation Exposure Report,” and TS 5.6.4, “Monthly Operating Reports.” The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on June 23, 2004 (69 FR 35067). The licensee affirmed the applicability of the model NSHC determination in its application dated December 16, 2004. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration 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 eliminates the Technical Specifications (TSs) reporting requirements to provide a monthly operating letter report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the TS reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. 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. The proposed change does not involve a physical alteration of the plant, add any new equipment, or require any existing equipment to be operated in a manner different from the present design. Therefore, the proposed 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. This is an administrative change to reporting requirements of plant operating information and occupational radiation exposure data, and has no effect on plant equipment, operating practices or safety analyses assumptions. For these reasons, the proposed change does not involve a significant reduction in the margin of safety. Based upon the reasoning presented above, the requested change does not involve significance hazards consideration. Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix, Arizona 85072-2034. NRC Section Chief: Robert A. Gramm. 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: December 1, 2004. Description of amendments request: The requested change will delete Technical Specification (TS) 5.6.1, “Occupational Radiation Exposure Report,” and TS 5.6.4, “Monthly Operating Reports.” The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on June 23, 2004 (69 FR 35067). The licensee affirmed the applicability of the model NSHC determination in its application dated December 1, 2004. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration 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 eliminates the Technical Specifications (TSs) reporting requirements to provide a monthly operating report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the TS reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. 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. The proposed change does not involve a physical alteration of the plant, add any new equipment, or require any existing equipment to be operated in a manner different from the present design. Therefore, the proposed 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. This is an administrative change to reporting requirements of plant operating information and occupational radiation exposure data, and has no effect on plant equipment, operating practices or safety analyses assumptions. For these reasons, the proposed change does not involve a significant reduction in the margin of safety. Based upon the reasoning presented above, the requested change does not involve significance hazards consideration. Attorney for licensee: James M. Petro, Jr., Esquire, Counsel, Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor, Baltimore, MD 21202. NRC Section Chief: Richard J. Laufer. Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan Date of amendment request: December 6, 2004. Description of amendment request: The requested change will delete Technical Specification (TS) 5.6.1, “Occupational Radiation Exposure Report,” and TS 5.6.4, “Monthly Operating Reports.” The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on June 23, 2004 (69 FR 35067). The licensee affirmed the applicability of the model NSHC determination in its application dated December 6, 2004. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration 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 eliminates the TSs reporting requirements to provide a monthly operating report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the TS reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. 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. The proposed change does not involve a physical alteration of the plant, add any new equipment, or require any existing equipment to be operated in a manner different from the present design. Therefore, the proposed 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. This is an administrative change to reporting requirements of plant operating information and occupational radiation exposure data, and has no effect on plant equipment, operating practices or safety analyses assumptions. For these reasons, the proposed change does not involve a significant reduction in the margin of safety. Based upon the reasoning presented above, the requested change does not involve significant hazards consideration. Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279 . NRC Section Chief: M. Kotzalas (Acting). Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power Station, Unit No. 2, New London County, Connecticut Date of amendment request: June 6, 2004. Description of amendment request: The proposed change would modify the Millstone Power Station, Unit No. 2 Technical Specifications (TSs) to extend the 10-year test interval for the Integrated Leakage Rate Test program to 15 years from the last Type A test. Specifically, the proposed change would revise TS 6.19, “Containment Leakage Rate Testing [CLRT] Program,” and permit a one-time, 5-year extension of the 10-year performance-based Type A test interval. In addition, the testing would be in accordance with the CLRT Program, Regulatory Guide (RG) 1.163, “Performance-Based Containment Leak-Test Program” and surveillance testing requirements as proposed in Nuclear Energy Institute 94-01 for Type A testing. 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 license amendment involve a significant increase in the probability or consequences of an accident previously evaluated? The proposed extension to Type A testing cannot increase the probability of an accident previously evaluated since extension of the containment Type A testing is not a physical plant modification that could alter the probability of accident occurrence, nor is it an activity or modification that by itself could lead to equipment failure or accident initiation. The proposed one-time, five-year extension to Type A testing does not result in a significant increase in the consequences of an accident as documented in NUREG-1493. The NUREG notes that very few potential containment leakage paths are not identified by Type B and C tests. It concludes that even reducing the Type A (ILRT [integrated leak rate test]) testing frequency to once per twenty years leads to an imperceptible increase in risk. DNC (the licensee) provides a high degree of assurance through indirect testing and inspection that the containment will not degrade in a manner detectable only by Type A testing. The last two Type A tests identified containment leakage within acceptance criteria, indicating a very leak-tight containment. Inspections required by the ASME Code [American Society of Mechanical Engineers Boiler and Pressure Vessel Code] are also performed in order to identify indications of containment degradation that could affect leak-tightness. Separately, Type B and C testing required by Technical Specifications, identifies any containment opening from design penetrations, such as valves, that would otherwise be detected by a Type A test. These factors establish that a one-time, five-year extension to the Millstone Unit 2 Type A test interval will not represent a significant increase in the consequences of an accident. 2. Does the proposed license amendment create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed revision to the Technical Specifications adds a one-time extension to the current interval for Type A testing for Millstone Unit 2. The current test interval of ten years, based on past performance, would be extended on a one-time basis to fifteen years from the last Type A test. The proposed extension to Type A testing does not create the possibility of a new or different type of accident since there are no physical changes being made to the plant and there are no changes to the operation of the plant that could introduce a new failure. 3. Does the proposed amendment involve a significant reduction in a margin of safety? The proposed revision to Millstone Unit 2 Technical Specifications adds a one-time extension to the current interval for Type A testing. The current test interval of ten years, based on past performance, would be extended on a one-time basis to fifteen years from the last Type A test for Millstone Unit 2. RG 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines very small changes in risk as resulting in increases of CDF [core damage frequency] below 10 − 6 /yr and increases in LERF [large early release frequency] below 10 − 7 /yr. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF, resulting from a change in the Type A ILRT test interval from a once-per-ten-years to a once-per-fifteen-years is 0.83 × 10 − 8 /yr, based on internal events. Since guidance in Reg. Guide 1.174 defines very small changes in LERF as below 10 − 7 /yr, increasing the ILRT interval from ten to fifteen years is, therefore, considered non-risk significant and will not significantly reduce the margin of safety. The NUREG-1493 generic study of the effects of extending containment leakage testing found that a 20-year interval in Type A leakage testing resulted in an imperceptible increase in risk to the public. NUREG-1493 generically concludes that the design containment leakage rate contributes about 0.1 percent of the overall risk. Decreasing the Type A testing frequency would have a minimal effect on this risk since 95% of the Type A detectable leakage paths would already be detected by Type B and C testing. Given that the proposed change will continue to meet the current design basis, any reduction in a margin of safety would not be significant. 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: Lillian M. Cuoco, Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 06385. NRC Section Chief: Darrell J. Roberts. Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone Power Station, Unit No. 3, New London County, Connecticut Date of amendment request: December 16, 2004. Description of amendment request: The proposed amendment would revise the current fuel rod average licensing basis burnup limit for one lead test assembly (LTA) containing advanced zirconium based alloys to a limit not exceeding 71,000 MWD/MTU. 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. Involve a significant increase in the probability or consequences of an accident previously evaluated. The Westinghouse LTA is very similar in design to the Westinghouse fuel that comprises the remainder of the core. The reload core design for Millstone Unit 3 Cycle 12, where one LTA will operate to high burnup, will meet all applicable design criteria. The performance of the Emergency Core Cooling System will not be affected by the operation of the LTA and operation of the LTA to high burnup will not result in a change to the Millstone Unit 3 reload design and safety analysis limits. Operation of one Westinghouse LTA to high burnup will not result in a measurable impact on normal operating releases, and will not increase the predicted radiological consequences of accidents postulated in Chapter 15 of the Millstone FSAR [final safety analysis report]. Therefore, neither the probability of occurrence nor the consequences of any accident previously evaluated is significantly increased. 2. Create the possibility of a new or different kind of accident from any accident previously evaluated. The Westinghouse LTA is very similar in design (both mechanical and composition of materials) to the resident Westinghouse fuel. All design and performance criteria will continue to be met and no new single failure mechanisms will be created. The irradiation of one LTA to high burnup does not involve any alteration to plant equipment or procedures, which would introduce any new or unique operational modes or accident precursors. Therefore, the possibility for a new or different kind of accident from any accident previously evaluated is not created. 3. Involve a significant reduction in a margin of safety. The operation of one Westinghouse LTA to high burnup does not change the performance requirements of any system or component such that any design criteria will be exceeded. The normal limits on core operation defined in the Millstone Unit 3 Technical Specifications will remain applicable for the core in which the high burnup assembly is irradiated. Therefore, the margin of safety as defined in the Bases to the Millstone Unit 3 Technical Specifications is not significantly reduced. 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: Lillian M. Cuoco, Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127. NRC Section Chief: Darrell Roberts. Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423, Millstone Power Station, Unit Nos. 2 and 3, New London County, Connecticut Date of amendment request: September 8, 2004. Description of amendment request: The proposed amendment deletes the requirements from the technical specifications (TSs) to maintain hydrogen recombiners and hydrogen monitors. Licensees were generally required to implement upgrades as described in NUREG-0737, “Clarification of TMI [Three Mile Island] Action Plan Requirements,” and Regulatory Guide (RG) 1.97, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.” Implementation of these upgrades was an outcome of the lessons learned from the accident that occurred at TMI Unit 2. Requirements related to combustible gas control were imposed by Order for many facilities and were added to or included in the TSs for nuclear power reactors currently licensed to operate. The revised 10 CFR 50.44, “Standards for Combustible Gas Control System in Light-Water-Cooled Power Reactors,” eliminated the requirements for hydrogen recombiners and relaxed safety classifications and licensee commitments to certain design and qualification criteria for hydrogen and oxygen monitors. The NRC staff issued a notice of availability of a model no significant hazards consideration determination for referencing in license amendment applications in the Federal Register on September 25, 2003 (68 FR 55416). The licensee affirmed the applicability of the model NSHC determination in its application dated September 8, 2004. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below: Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated The revised 10 CFR 50.44 no longer defines a design-basis loss-of-coolant accident (LOCA) hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant accident sequences that could threaten containment integrity. With the elimination of the design-basis LOCA hydrogen release, hydrogen monitors are no longer required to mitigate design-basis accidents and, therefore, the hydrogen monitors do not meet the definition of a safety-related component as defined in 10 CFR 50.2. Category 1 in RG 1.97 is intended for key variables that most directly indicate the accomplishment of a safety function for design-basis accident events. The hydrogen monitors no longer meet the definition of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 50.44 the Commission found that Category 3, as defined in RG 1.97, is an appropriate categorization for the hydrogen monitors because the monitors are required to diagnose the course of beyond design-basis accidents. The regulatory requirements for the hydrogen monitors can be relaxed without degrading the plant emergency response. The emergency response, in this sense, refers to the methodologies used in ascertaining the condition of the reactor core, mitigating the consequences of an accident, assessing and projecting offsite releases of radioactivity, and establishing protective action recommendations to be communicated to offsite authorities. Classification of the hydrogen monitors as Category 3, and removal of the hydrogen monitors from TS will not prevent an accident management strategy through the use of the severe accident management guidelines (SAMGs), the emergency plan (EP), the emergency operating procedures (EOP), and site survey monitoring that support modification of emergency plan protective action recommendations (PARs). Therefore, the elimination of the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TSs, does not involve a significant increase in the probability or the consequences of any accident previously evaluated. Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From any Previously Evaluated The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TSs, will not result in any failure mode not previously analyzed. The hydrogen recombiner and hydrogen monitor equipment was intended to mitigate a design-basis hydrogen release. The hydrogen recombiner and hydrogen monitor equipment are not considered accident precursors, nor does their existence or elimination have any adverse impact on the pre-accident state of the reactor core or post-accident confinement of radionuclides within the containment building. Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated. Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TSs, in light of existing plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, results in a neutral impact to the margin of safety. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. Category 3 hydrogen monitors are adequate to provide rapid assessment of current reactor core conditions and the direction of degradation while effectively responding to the event in order to mitigate the consequences of the accident. The intent of the requirements established as a result of the TMI Unit 2 accident can be adequately met without reliance on safety-related hydrogen monitors. Therefore, this change does not involve a significant reduction in the margin of safety. Removal of hydrogen monitoring from TSs will not result in a significant reduction in their functionality, reliability, and availability. The NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 06385. NRC Section Chief: Darrell J. Roberts. Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina; Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina Date of amendment request: September 20, 2004. Description of amendment request: The proposed amendment deletes the requirements from the technical specifications (TS) to maintain hydrogen recombiners (McGuire only) and hydrogen monitors (McGuire and Oconee). Licensees were generally required to implement upgrades as described in NUREG-0737, “Clarification of TMI [Three Mile Island] Action Plan Requirements,” and Regulatory Guide 1.97, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.” Implementation of these upgrades was an outcome of the lessons learned from the accident that occurred at TMI Unit 2. Requirements related to combustible gas control were imposed by Order for many facilities and were added to or included in the TS for nuclear power reactors currently licensed to operate. The revised 10 CFR 50.44, “Standards for Combustible Gas Control System in Light-Water-Cooled Power Reactors,” eliminated the requirements for hydrogen recombiners and relaxed safety classifications and licensee commitments to certain design and qualification criteria for hydrogen and oxygen monitors. The NRC staff issued a notice of availability of a model no significant hazards consideration determination for referencing in license amendment applications in the Federal Register on September 25, 2003 (68 FR 55416). The licensee affirmed the applicability of the model NSHC determination in its application dated September 20, 2004. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below: Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated The revised 10 CFR 50.44 no longer defines a design-basis loss-of-coolant accident (LOCA) hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant accident sequences that could threaten containment integrity. With the elimination of the design-basis LOCA hydrogen release, hydrogen monitors are no longer required to mitigate design-basis accidents and, therefore, the hydrogen monitors do not meet the definition of a safety-related component as defined in 10 CFR 50.2. Category 1 in [Regulatory Guide] RG 1.97 is intended for key variables that most directly indicate the accomplishment of a safety function for design-basis accident events. The hydrogen monitors no longer meet the definition of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 50.44 the Commission found that Category 3, as defined in RG 1.97, is an appropriate categorization for the hydrogen monitors because the monitors are required to diagnose the course of beyond design-basis accidents. The regulatory requirements for the hydrogen monitors can be relaxed without degrading the plant emergency response. The emergency response, in this sense, refers to the methodologies used in ascertaining the condition of the reactor core, mitigating the consequences of an accident, assessing and projecting offsite releases of radioactivity, and establishing protective action recommendations to be communicated to offsite authorities. Classification of the hydrogen monitors as Category 3, and removal of the hydrogen monitors from [Technical Specification] TS will not prevent an accident management strategy through the use of the severe accident management guidelines (SAMGs), the emergency plan (EP), the emergency operating procedures (EOP), and site survey monitoring that support modification of emergency plan protective action recommendations (PARs). Therefore, the elimination of the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, does not involve a significant increase in the probability or the consequences of any accident previously evaluated. Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From any Previously Evaluated The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, will not result in any failure mode not previously analyzed. The hydrogen recombiner and hydrogen monitor equipment was intended to mitigate a design-basis hydrogen release. The hydrogen recombiner and hydrogen monitor equipment are not considered accident precursors, nor does their existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radionuclides within the containment building. Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated. Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, in light of existing plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, results in a neutral impact to the margin of safety. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. Category 3 hydrogen monitors are adequate to provide rapid assessment of current reactor core conditions and the direction of degradation while effectively responding to the event in order to mitigate the consequences of the accident. The intent of the requirements established as a result of the TMI Unit 2 accident can be adequately met without reliance on safety-related hydrogen monitors. Therefore, this change does not involve a significant reduction in the margin of safety. Removal of hydrogen monitoring from TS will not result in a significant reduction in their functionality, reliability, and availability. The NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation, 422 South Church Street, Charlotte, North Carolina 28201-1006. NRC Section Chief: John A. Nakoski. Entergy Nuclear Operations, Docket Nos. 50-247 and 50-286, Indian Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York Date of amendment request: October 22, 2004. Description of amendment request: The proposed amendments would delete the requirements from the Technical Specifications (TSs) to maintain hydrogen recombiners and hydrogen monitors. Licensees were generally required to implement upgrades as described in NUREG-0737, “Clarification of TMI [Three Mile Island] Action Plan Requirements,” and Regulatory Guide (RG) 1.97, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.” Implementation of these upgrades was an outcome of the lessons learned from the accident that occurred at TMI Unit 2. Requirements related to combustible gas control were imposed by Order for many facilities and were added to or included in the TSs for nuclear power reactors currently licensed to operate. The revised 10 CFR 50.44, “Standards for Combustible Gas Control System in Light-Water-Cooled Power Reactors,” eliminated the requirements for hydrogen recombiners and relaxed safety classifications and licensee commitments to certain design and qualification criteria for hydrogen and oxygen monitors. The Nuclear Regulatory Commission (NRC) staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on September 25, 2003 (68 FR 55416). The licensee affirmed the applicability of the model NSHC determination in its application dated October 22, 2004. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below: Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated The revised 10 CFR 50.44 no longer defines a design-basis loss-of-coolant accident (LOCA) hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant accident sequences that could threaten containment integrity. With the elimination of the design-basis LOCA hydrogen release, hydrogen monitors are no longer required to mitigate design-basis accidents and, therefore, the hydrogen monitors do not meet the definition of a safety-related component as defined in 10 CFR 50.2. Category 1 in RG 1.97 is intended for key variables that most directly indicate the accomplishment of a safety function for design-basis accident events. The hydrogen monitors no longer meet the definition of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 50.44 the Commission found that Category 3, as defined in RG 1.97, is an appropriate categorization for the hydrogen monitors because the monitors are required to diagnose the course of beyond design-basis accidents. The regulatory requirements for the hydrogen monitors can be relaxed without degrading the plant emergency response. The emergency response, in this sense, refers to the methodologies used in ascertaining the condition of the reactor core, mitigating the consequences of an accident, assessing and projecting offsite releases of radioactivity, and establishing protective action recommendations to be communicated to offsite authorities. Classification of the hydrogen monitors as Category 3, and removal of the hydrogen monitors from TS will not prevent an accident management strategy through the use of the severe accident management guidelines (SAMGs), the emergency plan (EP), the emergency operating procedures (EOP), and site survey monitoring that support modification of emergency plan protective action recommendations (PARs). Therefore, the elimination of the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, does not involve a significant increase in the probability or the consequences of any accident previously evaluated. Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, will not result in any failure mode not previously analyzed. The hydrogen recombiner and hydrogen monitor equipment was intended to mitigate a design-basis hydrogen release. The hydrogen recombiner and hydrogen monitor equipment are not considered accident precursors, nor does their existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radionuclides within the containment building. Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated. Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, in light of existing plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, results in a neutral impact to the margin of safety. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. Category 3 hydrogen monitors are adequate to provide rapid assessment of current reactor core conditions and the direction of degradation while effectively responding to the event in order to mitigate the consequences of the accident. The intent of the requirements established as a result of the TMI Unit 2 accident can be adequately met without reliance on safety-related hydrogen monitors. Therefore, this change does not involve a significant reduction in the margin of safety. Removal of hydrogen monitoring from TS will not result in a significant reduction in their functionality, reliability, and availability. The NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Mr. John Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601. NRC Section Chief: Richard J. Laufer. Entergy Nuclear Operations, Docket Nos. 50-247 and 50-286, Indian Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York Date of amendment request: October 25, 2004. Description of amendment request: The requested change will delete Technical Specification (TS) 5.6.1, “Occupational Radiation Exposure Report,” and TS 5.6.4, “Monthly Operating Reports.” The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on June 23, 2004 (69 FR 35067). The licensee affirmed the applicability of the model NSHC determination in its application dated October 25, 2004. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration 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 eliminates the Technical Specifications (TSs) reporting requirements to provide a monthly operating report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the TS reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. 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. The proposed change does not involve a physical alteration of the plant, add any new equipment, or require any existing equipment to be operated in a manner different from the present design. Therefore, the proposed 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. This is an administrative change to reporting requirements of plant operating information and occupational radiation exposure data, and has no effect on plant equipment, operating practices or safety analyses assumptions. For these reasons, the proposed change does not involve a significant reduction in [a] margin of safety. Based upon the reasoning presented above, the requested change does not involve significance hazards consideration. Attorney for licensee: Mr. John Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601. NRC Section Chief: Richard J. Laufer. Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York Date of amendment request: December 30, 2004. Description of amendment request: The proposed amendment would revise a Technical Specification (TS) surveillance requirement (SR) in TS 3.1.4, “Control Rod Scram Times.” Specifically, the proposed change would revise the frequency for SR 3.1.4.2, “Control Rod Scram Time Testing,” from “120 days cumulative operation in MODE 1” to “200 days cumulative operation in MODE 1.” The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in licensing amendment applications in the Federal Register on August 23, 2004 (69 FR 51864). The licensee affirmed the applicability of the model NSHC determination in its application dated December 30, 2004. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below: 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change extends the frequency for testing control rod scram time testing from every 120 days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 operation. The frequency of surveillance testing is not an initiator of any accident previously evaluated. The frequency of surveillance testing does not affect the ability to mitigate any accident previously evaluated, as the tested component is still required to be operable. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change extends the frequency for testing control rod scram time testing from every 120 days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 operation. The proposed change does not result in any new or different modes of plant operation. 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 proposed change extends the frequency for testing control rod scram time testing from every 120 days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 operation. The proposed change continues to test the control rod scram time to ensure the assumptions in the safety analysis are protected. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Mr. John Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601. NRC Section Chief: Richard J. Laufer. Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas Date of amendment request: December 20, 2004. Description of amendment request: The proposed amendment would increase the lifting tripod's rating from 150 tons to 190 tons. This would allow for additional flexibility when lifting the new reactor vessel head during refueling outages. 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 ANO-1 [Arkansas Nuclear One, Unit 1] Tripod does not perform a safety function required by 10 CFR [Part] 50. The Tripod serves to perform heavy load movements during refueling outages[,] including [movement of] the reactor vessel head. Safe load paths have been established in accordance with NUREG-0612[, “Control of Heavy Loads at Nuclear Power Plants,”] to ensure that the fuel and safety[-]related equipment required to be inservice are protected. Use of actual Tripod eyelet Certified Material Test Reports (CMTRs) demonstrates that a safety factor of 3 to yield is maintained and that the lifting devices will perform their design function under maximum lifted loads. The Tripod does not serve any mitigative functions to lessen accidents. Therefore, the proposed change does not affect the probability or consequences of any ANO-1 analyzed accidents. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The only time that the Tripod is performing heavy loads movements is during Refueling operations. Safe load paths and load drop analyses have been performed to assure that heavy loads movements will not cause fuel damage or cause safety[-]related equipment to become inoperable. The proposed use of CMTRs instead of minimum yield strength of the material still assures that the Tripod will perform its required function to not create an accident. In addition, there is no change to the operation of the Tripod that would create a new failure mode or possible accident. Therefore, the proposed change does not create the possibility of a new or different type of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The design margin for the Tripod is established by NUREG-0612 and ANSI [American National Standards Institute] N14.6-1978[, “Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More for Nuclear Materials”]. A factor of safety of 3 for yield strength and 5 for ultimate strength for both the static and dynamic load factors is required to be met. These factors of safety provide sufficient margin to assure that the Tripod will perform its design function of maximum lifted loads. In addition, the use [of] a dynamic load factor of 1.15 above the static load is well above the actual dynamic factor to be experienced from the design lift speed of the polar crane. The use of CMTRs does not result in a significant reduction in the margin of safety of the Tripod. In addition, the Tripod will be load tested to 150% [percent] of its design static and dynamic loading which will further assure adequate safety margin. Therefore, the margin of safety is not changed by the proposed change to the ANO-1 SAR [Safety Analysis Report]. 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: Nicholas S. Reynolds, Esquire, Winston and Strawn, 1400 L Street, NW., Washington, DC 20005-3502. NRC Section Chief: Robert A. Gramm. Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi Date of amendment request: December 17, 2004. Description of amendment request: The proposed change will revise the air lock surveillance test acceptance criteria to be consistent with the NRC approved Industry Technical Specification Task Force (TSTF) change to the Standard Technical Specifications (STS), TSTF-52, entitled “Implement 10 CFR [Part] 50, Appendix J, Option B.” By letter dated April 6, 1998, the NRC Staff issued amendment number 135 to the GGNS license permitting the implementation of the containment leak rate testing provisions of 10 CFR Part 50, Appendix J, Option 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 or consequences of an accident previously evaluated? Response: No. Primary containment air lock leak rate testing can have no effect on the probability of any postulated accident. The proposed change will increase the allowed containment air lock leakage rate and convert it from an absolute leakage rate to a percentage of the overall primary containment leakage rate. No change to the overall leakage rate of the containment is being proposed, therefore there is no change to the consequences of any postulated accident. The change in air lock leakage rate will not impact the design or operation of any plant system or component nor will they affect initiation or mitigation of any accidents previously analyzed. 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. The primary containment air locks form part of the primary containment pressure boundary. The periodic containment air lock leakage rate tests specified in SR 3.6.1.2.1 verifies that the air lock leakage does not exceed the allowed fraction of the overall primary containment leakage rate. This request involves a change in the allowable leakage rate of the primary containment air locks without increasing the overall allowed leakage rate of the containment. Changing the allowable leakage rate has no influence on, nor does it contribute in any way to, the possibility of a new or different kind of accident or malfunction from those previously analyzed. There will be no effect on the types and amounts of overall leakage from the primary containment boundary. The proposed amendment will not produce any changes to the design or operation of the plant. The method of performing the test is not changed. No new accident modes are created by changing the allowable leakage in this manner. No safety-related equipment or safety functions are altered as a result of this change. 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. Air lock integrity and leak tightness are essential for maintaining primary containment leakage rate to within limits in the event of a design basis accident. The periodic containment air lock leakage rate tests verify that the air lock leakage does not exceed the allowed fraction of the overall primary containment leakage rate. Since no changes are proposed to the maximum allowable primary containment leakage rate, the design basis radiological analysis is not impacted by this change. The license amendment request removes unnecessary conservatism from the testing program and allows consistency with current industry practice. Since no changes are proposed to the maximum allowable primary containment leakage rate, the design basis radiological analysis is not impacted by this change. 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: Nicholas S. Reynolds, Esquire, Winston and Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502. NRC Section Chief: Michael K. Webb. Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois; Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois; Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois; Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Date of amendment request: September 15, 2004. Description of amendment request: The proposed amendment would delete requirements from the Technical Specifications (TSs) to maintain hydrogen recombiners and hydrogen and oxygen monitors. Licensees were generally required to implement upgrades as described in NUREG-0737, “Clarification of TMI [Three Mile Island] Action Plan Requirements,” and Regulatory Guide (RG) 1.97, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.” Implementation of these upgrades was an outcome of the lessons learned from the accident that occurred at TMI, Unit 2. Requirements related to combustible gas control were imposed by order for many facilities and were added to, or included, in the TSs for nuclear power reactors currently licensed to operate. The revised Title 10 of the Code of Federal Regulations (10 CFR) Section 50.44, “ Combustible gas control for nuclear power reactors,” eliminated the requirements for hydrogen recombiners and relaxed safety classifications and licensee commitments to certain design and qualification criteria for hydrogen and oxygen monitors. The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on September 25, 2003 (68 FR 55416). The licensee affirmed the applicability of the model NSHC determination in its application dated September 15, 2004. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of NSHC is presented below: Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated The revised 10 CFR 50.44 no longer defines a design-basis loss-of-coolant accident (LOCA) hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant accident sequences that could threaten containment integrity. With the elimination of the design-basis LOCA hydrogen release, hydrogen and oxygen monitors are no longer required to mitigate design-basis accidents and, therefore, the hydrogen monitors do not meet the definition of a safety-related component as defined in 10 CFR 50.2. RG 1.97 Category 1, is intended for key variables that most directly indicate the accomplishment of a safety function for design-basis accident events. The hydrogen and oxygen monitors no longer meet the definition of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 50.44 the Commission found that Category 3, as defined in RG 1.97, is an appropriate categorization for the hydrogen monitors because the monitors are required to diagnose the course of beyond design-basis accidents. Also, as part of the rulemaking to revise 10 CFR 50.44, the Commission found that Category 2, as defined in RG 1.97, is an appropriate categorization for the oxygen monitors, because the monitors are required to verify the status of the inert containment. The regulatory requirements for the hydrogen and oxygen monitors can be relaxed without degrading the plant emergency response. The emergency response, in this sense, refers to the methodologies used in ascertaining the condition of the reactor core, mitigating the consequences of an accident, assessing and projecting offsite releases of radioactivity, and establishing protective action recommendations to be communicated to offsite authorities. Classification of the hydrogen monitors as Category 3, classification of the oxygen monitors as Category 2, and removal of the hydrogen and oxygen monitors from TS will not prevent an accident management strategy through the use of the SAMGs [severe accident management guidelines], the emergency plan (EP), the emergency operating procedures (EOPs), and site survey monitoring that support modification of emergency plan protective action recommendations (PARs). Therefore, the elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, does not involve a significant increase in the probability or the consequences of any accident previously evaluated. Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, will not result in any failure mode not previously analyzed. The hydrogen recombiner and hydrogen and oxygen monitor equipment was intended to mitigate a design-basis hydrogen release. The hydrogen recombiner and hydrogen and oxygen monitor equipment are not considered accident precursors, nor does their existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radionuclides within the containment building. Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated. Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, in light of existing plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, results in a neutral impact to the margin of safety. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. Category 3 hydrogen monitors are adequate to provide rapid assessment of current reactor core conditions and the direction of degradation while effectively responding to the event in order to mitigate the consequences of the accident. The intent of the requirements established as a result of the TMI, Unit 2 accident can be adequately met without reliance on safety-related hydrogen monitors. Category 2 oxygen monitors are adequate to verify the status of an inerted containment. Therefore, this change does not involve a significant reduction in the margin of safety. The intent of the requirements established as a result of the TMI, Unit 2 accident can be adequately met without reliance on safety-related oxygen monitors. Removal of hydrogen and oxygen monitoring from TS will not result in a significant reduction in their functionality, reliability, and availability. Based on the reasoning presented above and the previous discussion of the amendment request, the requested change does not involve a significant hazards consideration. Attorney for licensee: Mr. Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Section Chief: Gene Y. Suh. Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois; Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, Rock Island County, Illinois Date of amendment request: November 4, 2004. Description of amendment request: The proposed amendments would revise the plant technical specification (TS) pressure and temperature (P/T) limit curves for 54 effective full power years (EFPY) to support a 20-year license extension for both DNPS and QCNPS to 60 years ( i.e. , 54 EFPY), and resolves a non-conservative condition for TS Section 3.4.9, Figure 3.4.9-2, “Non-Nuclear Heatup/Cooldown Curve,” for QCNPS. Basis for proposed no significant hazards consideration determination: As required by Title 10 of the Code of Federal Regulations (10 CFR) section 50.91(a), Exelon Generation Company (EGC) has provided its analysis of the issue of no significant hazards consideration (NSHC), 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. In support of this determination, an evaluation of each of the three criteria set forth in 10 CFR 50.92 is provided below regarding the proposed license amendment. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? The proposed changes request that, for DNPS, Units 2 and 3 and QCNPS, Units 1 and 2, P/T limit curves in TS 3.4.9, “RCS Pressure and Temperature (P/T) Limits,” be revised. 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 non-ductile failure of the reactor coolant pressure boundary, which is an unanalyzed condition. The methodology used to determine the P/T limits has been approved by the NRC [Nuclear Regulatory Commission] and thus is an acceptable method for determining these limits. Therefore, the proposed changes do not affect the probability of an accident previously evaluated. There is no specific accident that postulates a non-ductile failure of the reactor coolant pressure (RCP) boundary. The loss of coolant accident analyzed for the plant assumes a 4.281 square feet complete break of the recirculation pump suction line. The revision to the P/T limits does not change this assumption. Thus, the radiological consequences of any 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. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed changes do not change the response of plant equipment to transient conditions. The proposed changes do not introduce any new equipment, modes of system operation, or failure mechanisms. Non-ductile failure of the RCP boundary is not an analyzed accident. The proposed changes to the P/T limits were developed using an NRC-approved methodology, and thus the revised limits will continue to provide protection against non-ductile failure of the RCP boundary. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. Does the change involve a significant reduction in a margin of safety? The margin of safety related to the proposed changes is the margin between the proposed P/T limits and the pressures and temperatures that would produce nonductile failure of the RCP boundary. NRC requirements to protect the integrity of the reactor coolant pressure boundary in nuclear power plants is established in 10 CFR 50, Appendix G, “Fracture Toughness Requirements,” which requires that the P/T limits for an operating plant be at least as conservative as those that would be generated if the methods of American Society of Mechanical Engineers, Section XI, Appendix G, were applied. The use of an NRC-approved methodology, together with conservatively chosen plant-specific input parameters, provides an acceptable margin of safety. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. Based upon the above responses, EGC concluded that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92 and, accordingly, a finding of no significant hazards consideration is justified. The NRC staff has reviewed EGC'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 NSHC. Attorney for licensee: Mr. Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Section Chief: Gene Y. Suh. 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: September 15, 2004. Description of amendment request: The proposed amendment would delete requirements from the Technical Specifications (TSs) to maintain containment hydrogen and oxygen monitors. A notice of availability for this technical specification improvement using the consolidated line item improvement process (CLIIP) was published in the Federal Register on September 25, 2003 (68 FR 55416). Licensees were generally required to implement upgrades as described in NUREG-0737, “Clarification of TMI [Three Mile Island] Action Plan Requirements,” and Regulatory Guide 1.97, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.” Implementation of these upgrades was an outcome of the lessons learned from the accident that occurred at TMI, Unit 2. Requirements related to combustible gas control were imposed by Order for many facilities and were added to or included in the TSs for nuclear power reactors currently licensed to operate. The revised 10 CFR 50.44, “Standards for combustible gas control system in light-water-cooled power reactors,” eliminated the requirements for hydrogen recombiners and relaxed safety classifications and licensee commitments to certain design and qualification criteria for hydrogen and oxygen monitors. The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on September 25, 2003 (68 FR 55416). The licensee affirmed the applicability of the relevant portions of the model NSHC determination (hydrogen and oxygen monitors only) in its application dated September 15, 2004. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below: Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated The revised 10 CFR 50.44 no longer defines a design-basis loss-of-coolant accident (LOCA) hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant accident sequences that could threaten containment integrity. With the elimination of the design-basis LOCA hydrogen release, hydrogen and oxygen monitors are no longer required to mitigate design-basis accidents and, therefore, the hydrogen monitors do not meet the definition of a safety-related component as defined in 10 CFR 50.2. RG [Regulatory Guide] 1.97 Category 1, is intended for key variables that most directly indicate the accomplishment of a safety function for design-basis accident events. The hydrogen and oxygen monitors no longer meet the definition of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 50.44 the Commission found that Category 3, as defined in RG 1.97, is an appropriate categorization for the hydrogen monitors because the monitors are required to diagnose the course of beyond design-basis accidents. Also, as part of the rulemaking to revise 10 CFR 50.44, the Commission found that Category 2, as defined in RG 1.97, is an appropriate categorization for the oxygen monitors, because the monitors are required to verify the status of the inert containment. The regulatory requirements for the hydrogen and oxygen monitors can be relaxed without degrading the plant emergency response. The emergency response, in this sense, refers to the methodologies used in ascertaining the condition of the reactor core, mitigating the consequences of an accident, assessing and projecting offsite releases of radioactivity, and establishing protective action recommendations to be communicated to offsite authorities. Classification of the hydrogen monitors as Category 3, [classification of the oxygen monitors as Category 2,] and removal of the hydrogen and oxygen monitors from TS will not prevent an accident management strategy through the use of the severe accident management guidelines (SAMGs), the emergency plan (EP), the emergency operating procedures (EOPs), and site survey monitoring that support modification of emergency plan protective action recommendations (PARs). Therefore, the elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, does not involve a significant increase in the probability or the consequences of any accident previously evaluated. Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, will not result in any failure mode not previously analyzed. The hydrogen recombiner and hydrogen and oxygen monitor equipment was intended to mitigate a design-basis hydrogen release. The hydrogen recombiner and hydrogen and oxygen monitor equipment are not considered accident precursors, nor does their existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radionuclides within the containment building. Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated. Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, in light of existing plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, results in a neutral impact to the margin of safety. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. Category 3 hydrogen monitors are adequate to provide rapid assessment of current reactor core conditions and the direction of degradation while effectively responding to the event in order to mitigate the consequences of the accident. The intent of the requirements established as a result of the TMI, Unit 2 accident can be adequately met without reliance on safety-related hydrogen monitors. Category 2 oxygen monitors are adequate to verify the status of an inerted containment. Therefore, this change does not involve a significant reduction in the margin of safety. The intent of the requirements established as a result of the TMI, Unit 2 accident can be adequately met without reliance on safety-related oxygen monitors. Removal of hydrogen and oxygen monitoring from TS will not result in a significant reduction in their functionality, reliability, and availability. The NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for Licensee: Thomas S. O'Neill, Associate and General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Section Chief: Darrell Roberts. Florida Power Corporation, et al., Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant, Citrus County, Florida Date of amendment request: September 21, 2004. Description of amendment request: The proposed amendment deletes the requirements from the technical specifications (TS) to maintain containment hydrogen monitors. Licensees were generally required to implement upgrades as described in NUREG-0737, “Clarification of TMI [Three Mile Island] Action Plan Requirements,” and Regulatory Guide (RG) 1.97, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.” Implementation of these upgrades was an outcome of the lessons learned from the accident that occurred at TMI Unit 2. Requirements related to combustible gas control were imposed by Order for many facilities and were added to or included in the TS for nuclear power reactors currently licensed to operate. The revised 10 CFR 50.44, “Standards for Combustible Gas Control System in Light-Water-Cooled Power Reactors,” eliminated the requirements for hydrogen recombiners and relaxed safety classifications and licensee commitments to certain design and qualification criteria for hydrogen and oxygen monitors. The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on September 25, 2003 (68 FR 55416). The licensee affirmed the applicability of the relevant portions of the model NSHC determination (hydrogen monitors only) in its application dated September 21, 2004. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below: Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated The revised 10 CFR 50.44 no longer defines a design-basis loss-of-coolant accident (LOCA) hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant accident sequences that could threaten containment integrity. With the elimination of the design-basis LOCA hydrogen release, hydrogen monitors are no longer required to mitigate design-basis accidents and, therefore, the hydrogen monitors do not meet the definition of a safety-related component as defined in 10 CFR 50.2. Category 1 in RG 1.97 is intended for key variables that most directly indicate the accomplishment of a safety function for design-basis accident events. The hydrogen monitors no longer meet the definition of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 50.44 the Commission found that Category 3, as defined in RG 1.97, is an appropriate categorization for the hydrogen monitors because the monitors are required to diagnose the course of beyond design-basis accidents. The regulatory requirements for the hydrogen monitors can be relaxed without degrading the plant emergency response. The emergency response, in this sense, refers to the methodologies used in ascertaining the condition of the reactor core, mitigating the consequences of an accident, assessing and projecting offsite releases of radioactivity, and establishing protective action recommendations to be communicated to offsite authorities. Classification of the hydrogen monitors as Category 3, and removal of the hydrogen monitors from TS will not prevent an accident management strategy through the use of the severe accident management guidelines (SAMGs), the emergency plan (EP), the emergency operating procedures (EOPs), and site survey monitoring that support modification of emergency plan protective action recommendations (PARs). Therefore, the elimination of the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, does not involve a significant increase in the probability or the consequences of any accident previously evaluated. Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, will not result in any failure mode not previously analyzed. The hydrogen recombiner and hydrogen monitor equipment was intended to mitigate a design-basis hydrogen release. The hydrogen recombiner and hydrogen monitor equipment are not considered accident precursors, nor does their existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radionuclides within the containment building. Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated. Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, in light of existing plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, results in a neutral impact to the margin of safety. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. Category 3 hydrogen monitors are adequate to provide rapid assessment of current reactor core conditions and the direction of degradation while effectively responding to the event in order to mitigate the consequences of the accident. The intent of the requirements established as a result of the TMI Unit 2 accident can be adequately met without reliance on safety-related hydrogen monitors. Therefore, this change does not involve a significant reduction in the margin of safety. Removal of hydrogen monitoring from TS will not result in a significant reduction in their functionality, reliability, and availability. The NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Steven R. Carr, Associate General Counsel—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602. NRC Section Chief: Michael L. Marshall. Florida Power Corporation, et al. , Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant, Citrus County, Florida Date of amendment request: January 13, 2005. Description of amendment request: The proposed change would allow a one-time extended allowed outage time (AOT) change to Improved Technical Specifications (ITS) 3.5.2, Emergency Core Cooling Systems (ECCS)—Operating; 3.6.6, Reactor Building Spray and Containment Cooling Systems; 3.7.8, Decay Heat Closed Cycle Cooling Water System (DC); and 3.7.10, Decay Heat Seawater System to allow the refurbishment of Decay Heat Seawater System Pump RWP-3B online. 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: This request has been evaluated against the standards in 10 CFR 50.92, and has been determined to not involve a significant hazards consideration. In support of this conclusion, the following analysis is provided: 1. Does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed license amendment extends, on a one-time basis, the Completion Time for the systems described above from 72 hours to 10 days. These Systems are designed to provide cooling for components essential to the mitigation of plant transients and accidents. The systems are not initiators of design basis accidents. The proposed ITS changes have been evaluated to assess their impact on normal operation of the systems affected and to ensure that their design basis safety functions are preserved. A Probabilistic Safety Assessment (PSA) has been performed to assess the risk impact of an increase in Completion Time from 72 hours to 10 days. Although the proposed one-time change results in an increase in Core Damage Frequency (CDF) and Large Early Release Frequency (LERF), the value of these increases are considered as small (CDF) and very small (LERF) in the current regulatory guidance. Therefore, granting this LAR [License Amendment Request] does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does not create the possibility of a new or different type of accident from any accident previously evaluated. The proposed license amendment extends, on a one-time basis, the Completion Time for the systems described above from 72 hours to 10 days. The proposed LAR will not result in changes to the design, physical configuration of the plant or the assumptions made in the safety analysis. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does not involve a significant reduction in the margin of safety. The proposed license amendment extends, on a one-time basis, the Completion Time for the systems described above from 72 hours to 10 days. The proposed change will allow online repair of Decay Heat Seawater pump RWP-3B to restore the pump to full qualification which will improve its reliability and useful lifetime, thus increasing the long term margin of safety of the system. The proposed LAR will reduce the probability (and associated risk) of a plant shutdown to repair a Decay Heat Services Seawater pump. To ensure defense-in-depth capabilities and the assumptions in the risk assessment are maintained during the proposed one-time extended Completion Time, CR-3 will continue the performance of 10 CFR 50.65(a)(4) assessments before performing maintenance or surveillance activities and no maintenance activities of other risk sensitive equipment beyond that required for the refurbishment activity will be scheduled concurrent with the repair activity. Other compensatory actions that will be implemented include: operator attention to the importance of protecting the operable redundant train and support systems will be increased, selection of beneficial Makeup Pump configurations, no elective maintenance will be scheduled in the switchyard, and the establishment of fire watches. As described above in Item 1, a PSA has been performed to assess the risk impact of an increase in Completion Time. Although the proposed one-time change results in an increase in Core Damage Frequency (CDF), and Large Early Release Frequency (LERF), the value of these increases is considered as small (CDF) and very small (LERF) in the current regulatory guidance. Therefore, granting this LAR does not involve a significant reduction in the margin of safety. Based on the above, Progress Energy Florida, Inc. (PEF) concludes that the proposed LAR presents a 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 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 T. Conley, Associate General Counsel II—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602. NRC Section Chief: Michael L. Marshall. Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa Date of amendment request: October 29, 2004. Description of amendment request: The proposed amendment would revise Technical Specification 3.1.8, “Scram Discharge Volume (SDV) Vent and Drain Valves,” to allow a vent or drain line with one inoperable valve to be isolated instead of requiring the valve to be restored to Operable status within 7 days. The NRC staff issued a notice of opportunity for comment in the Federal Register on February 24, 2003 (68 FR 8637), on possible amendments to revise the action for one or more SDV vent or drain lines with an inoperable valve, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line-item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 15, 2003 (68 FR 18294). The licensee affirmed the applicability of the model NSHC determination in its application dated October 29, 2004. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below: Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated A change is proposed to allow the affected SDV vent and drain line to be isolated when there are one or more SDV vent or drain lines with one valve inoperable instead of requiring the valve to be restored to operable status within 7 days. With one SDV vent or drain valve inoperable in one or more lines, the isolation function would be maintained since the redundant valve in the affected line would perform its safety function of isolating the SDV. Following the completion of the required action, the isolation function is fulfilled since the associated line is isolated. The ability to vent and drain the SDV is maintained and controlled through administrative controls. This requirement assures the reactor protection system is not adversely affected by the inoperable valves. With the safety functions of the valves being maintained, the probability or consequences of an accident previously evaluated are not significantly increased. Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. Thus, this change does not create the possibility of a new or different kind of accident from any previously evaluated. Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety The proposed change ensures that the safety functions of the SDV vent and drain valves are fulfilled. The isolation function is maintained by redundant valves and by the required action to isolate the affected line. The ability to vent and drain the SDV is maintained through administrative controls. In addition, the reactor protection system will prevent filling of the SDV to the point that it has insufficient volume to accept a full scram. Maintaining the safety functions related to isolation of the SDV and insertion of control rods ensures that the proposed change does not involve a significant reduction in the margin of safety. 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 Section Chief: M. Kotzalas (Acting). Southern California Edison Company, et al., Docket Nos. 50-361 and 50-362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego County, California Date of amendment requests: December 27, 2004. Description of amendment requests: The requested change will delete Technical Specification (TS) 5.7.1.1.a, “Occupational Radiation Exposure Report,” and TS 5.7.1.4, “Monthly Operating Reports.” The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on June 23, 2004 (69 FR 35067). The licensee affirmed the applicability of the model NSHC determination in its application dated December 27, 2004. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration 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 eliminates the Technical Specifications (TSs) reporting requirements to provide a monthly operating letter report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the TS reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. 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. The proposed change does not involve a physical alteration of the plant, add any new equipment, or require any existing equipment to be operated in a manner different from the present design. Therefore, the proposed 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. This is an administrative change to reporting requirements of plant operating information and occupational radiation exposure data, and has no effect on plant equipment, operating practices or safety analyses assumptions. For these reasons, the proposed change does not involve a significant reduction in the margin of safety. Based upon the reasoning presented above, the requested change does not involve significance hazards consideration. Attorney for licensee: Douglas K. Porter, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770. NRC Section Chief: Robert A. Gramm. Southern California Edison Company, et al., Docket Nos. 50-361 and 50-362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego County, California Date of amendment requests: December 27, 2004. Description of amendment requests: The proposed amendments would revise the San Onofre Nuclear Generating Station (SONGS), Units 2 and 3 accident source term used in the design basis radiological consequences analyses. These license amendments are requested in accordance with the requirements of 10 CFR 50.67, which addresses the use of an Alternative Source Term (AST) at operating reactors, and relevant guidance of Regulatory Guide 1.183. These license amendments represent full-scope implementation of the AST described in Regulatory Guide 1.183. 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 to the Facility Operating Licenses for San Onofre Units 2 and 3 credit an Alternative Source Term (AST) for the design basis radiological site boundary and control room dose analyses. This change represents full scope implementation of the AST as described in Regulatory Guide 1.183. The proposed changes to the Facility Operating Licenses also expand the allowed use of fuel failure estimates by Departure from Nucleate Boiling (DNB) statistical convolution methodology from only the reactor coolant pump sheared shaft event to the Updated Final Safety Analysis Report (UFSAR) Chapter 15 non-Loss-of-Coolant-Accident (LOCA) events that assume a loss of flow ( i.e. , a loss of AC power) and that fail fuel. The proposed changes reflect the parameters used in the radiological consequences calculations for the LOCA, Fuel Handling Accident inside containment (FHA-IC), Fuel Handling Accident in the Fuel Handling Building (FHA-FHB) and pre-trip Steam Line Break Outside Containment (SLB-OC). The purpose of this proposed change is to change the design requirements for the Control Room Envelope (CRE). This proposed change will allow an increase in the assumed amount of unfiltered air inleakage through the CRE. Currently, design basis radiological consequence analyses assume CRE inleakage of 0 cfm, plus an assumed 10 cubic feet per minute (cfm) inleakage due to ingress and egress into the Control Room. Analyses to support this change demonstrate acceptable post-accident dose conseq [Text truncated at 120,000 characters. The full text is on the page linked above.] --- Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A05-1574. Public record. Not legal advice.