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, August 20, 2004, through September 2, 2004. The last biweekly notice was published on August 31, 2004, (69 FR 53098).
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 email 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 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/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by email to pdr@nrc.gov.
AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New York
Date of amendment request: August 27, 2004.
Description of amendment request: The licensee proposed to amend the Oyster Creek Nuclear Generating Station (OCNGS) Technical Specifications (TSs) regarding the safety limit minimum critical power ratio (SLMCPR) to reflect the results of cycle-specific calculations performed for the next fuel cycle (i.e., Cycle 20), using Nuclear Regulatory Commission (NRC)-approved methodology documented in Topical Report NEDE-24011-P-A-14, “General Electric Standard Application for Reactor Fuel” (GESTAR II), updated to Amendment 25. Specifically, the licensee proposed to revise TS Section 2.1.A, changing the SLMCPR values from 1.10 to 1.12 for three-recirculation-loop operation, and from 1.09 to 1.10 for four- or five-recirculation-loop operation.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration. The NRC staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The NRC staff's review is presented below:
(1) Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
No. SLMCPR values, and their derivation using NRC-approved methods, do not change the design or operating procedures of OCNGS, and have no role on the occurrence of an initiating event of an accident or transient. The basis of the SLMCPR is to ensure no mechanistic fuel damage will occur if the limit is not violated. The new SLMCPR values will preserve the existing margin to transition boiling (i.e., in the event of an accident or transient, the amount of fuel damaged would not be increased as a result of the new SLMCPR values). Furthermore, the proposed new SLMCPR values do not lead to, nor do they arise as a result of, plant design or procedural changes. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
(2) Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
No. The new SLMCPR values for OCNGS Cycle 20 core have been calculated in accordance with the methods and procedures described in an NRC-approved topical report. The proposed new SLMCPR values do not lead to, nor do they arise as a result of, plant design or procedural changes. The changes do not involve any new method for operating the facility and do not involve any facility modifications. As a result, no new initiating events or transients could develop from the proposed changes. Therefore, the proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
(3) Does the proposed amendment involve a significant reduction in a margin of safety?
No. The margin of safety as defined in OCNGS's licensing basis will remain the same. The new cycle-specific SLMCPR values are calculated using NRC-approved methods and procedures that are in accordance with the current fuel design and licensing criteria. The SLMCPR values will remain high enough to ensure that greater than 99.9% of all fuel rods in the core are expected to avoid transition boiling if the limits are not violated, thereby preserving the fuel cladding integrity. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.
Based on the above review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the proposed amendment involves no significant hazards consideration.
Attorney for licensee: Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LCC, 4300 Winfield Road, Warrenville, IL 60555.
NRC Section Chief: Richard J. Laufer.
Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: May 27, 2004.
Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) based on the radiological dose analysis margins obtained by using an alternative source term consistent with 10 CFR 50.67. Specifically, the amendment would revise TS 3/4.7.7, “Control Room Emergency Air Filtration System,” surveillance requirements and delete TS 3/4.7.8, “Control Room Envelope Pressurization System.”
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 proposed amendment does not involve a significant increase in the probability or consequence of an accident previously analyzed. The Millstone Unit 3 Control Room Emergency Air Filtration System only functions following the initiation of a design basis radiological accident. Therefore, the change to the value used for methyl iodide penetration test acceptance criteria following a design basis accident will not increase the probability of any previously analyzed accident. The Millstone Unit 3 Control Room Envelope Pressurization System is no longer credited in the accident analyses described in the Alternative Source Term (AST) implementation analyses. In accordance with AST implementation analyses, the requirements contained in this Specification do not meet any of 10 CFR 50.36(c)(2)(ii) criteria on items for which Technical Specifications must be established. Deletion of this Technical Specification will not increase the probability of occurrence of any previously analyzed accident and does not impact the consequences of any evaluated accident since it is no longer analytically credited. The Millstone Unit 3 containment and the containment systems function to prevent or control the release of radioactive fission products following a postulated accident. Therefore, the change to the value used for the leakage rate acceptance criteria for all penetrations that are secondary containment bypass leakage paths following a design basis accident will not increase the probability of any previously analyzed accident and is limited to ensure it does not increase any accident consequence.
These systems are not initiators of any design bases accident. Revised dose calculations, which take into account the changes proposed by this amendment and the use of the alternative source term, have been performed for the Millstone Unit 3 design basis radiological accidents. The results of these revised calculations indicate that public and control room doses will not exceed the limits specified in 10 CFR 50.67 and Regulatory Guide 1.183. There is not a significant increase in predicted dose consequences for any of the analyzed accidents. Therefore, the proposed changes do not involve a significant increase in the consequences of any previously analyzed accident.
2. Create the possibility of a new or different kind of accident from any accident previously evaluated.
The implementation of the proposed changes does not create the possibility of an accident of a different type than was previously evaluated in the UFSAR [updated final safety report]. Although the proposed changes could affect the operation of the Control Room Emergency Air Filtration System, and containment and the containment systems following a design basis radiological accident, none of these changes can initiate a new or different kind of accident since they are only related to system capabilities that provide protection from accidents that have already occurred. These changes do not alter the nature of events postulated in the UFSAR nor do they introduce any unique precursor mechanisms. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from those previously analyzed.
3. Involve a significant reduction in the margin of safety.
The implementation of the proposed changes does not reduce the margin of safety. The proposed changes for the Control Room Emergency Air Filtration System, and containment and the containment systems do not affect the ability of these systems to perform their intended safety functions to maintain dose less than the required limits during design basis radiological events. The revised dose calculations also indicate that the change to the containment depressurization times will continue to maintain the dose to the public and control room operators less than the required limits. The radiological analysis results, when compared with the revised TEDE acceptance criteria, meet the applicable limits. These acceptance criteria have been developed for application to analyses performed with alternative source terms. These acceptance criteria have been developed for the purpose of use in design basis accident analyses such that meeting the stated limits demonstrates adequate protection of public health and safety. It is thus concluded that the margin of safety will not be reduced by the implementation of the changes.
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: James W. Clifford.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: March 22, 2004.
Description of amendment request: The amendments would revise the Catawba Nuclear Station Facility Operating Licenses and Technical Specifications (TSs) to change the surveillance frequency on selected Engineered Safety Features Actuation System (ESFAS) slave relays from 92 days to 18 months.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
(1) The proposed license amendments do not involve a significant increase in the probability or consequences of an accident previously evaluated.
This change to the TS does not result in a condition where the design, material, and construction standards that were applicable prior to the change are altered. Only the slave relay test interval is changed. The proposed change will not modify any system interface and could not increase the likelihood of an accident since these events are independent of this change. The proposed activity will not change, degrade, or prevent actions or alter any assumptions previously made in evaluating the radiological consequences of an accident described in the UFSAR [Updated Final Safety Analysis Report]. Therefore, the proposed amendments do not result in any increase in the probability or consequences of an accident previously evaluated.
(2) The proposed license amendments do not create the possibility of a new or different kind of accident from any accident previously evaluated.
This change does not alter the performance of the affected systems. The slave relays will still be tested every 18 months. Changing the surveillance frequency for the slave relays will not create any new accident initiators or scenarios. Periodic surveillance of these instruments will detect significant degradation in the channel characteristic. Implementation of the proposed amendments does not create the possibility of a new or different kind of accident from any accident previously evaluated.
(3) The proposed license amendments do not involve a significant reduction in a margin of safety.
The surveillance test frequency is relaxed for certain slave relays because of demonstrated high reliability of the relay and its insensitivity to any short term wear or aging effects. Based on the above, it is concluded that the proposed license amendment request does not result in a reduction in a margin with respect to plant 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: Ms. Lisa F. Vaughn, Legal Department (PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, North Carolina 28201-1006.
NRC Section Chief: Mary Jane Ross-Lee, Acting.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: April 6, 2004.
Description of amendment request: The amendments would revise the Catawba Nuclear Station Technical Specifications (TSs) to allow a diesel generator battery to remain operable with no more than one cell less than 1.36 Volts DC on float charge.
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 change involve a significant increase in the probability or consequences of an accident previously evaluated?
No. The DC electrical power system provides normal and emergency DC electrical power for the diesel generators, emergency auxiliaries, and control and switching during all modes of operation. This change will not affect or degrade the ability of the DC Electrical Power Systems to perform their specified safety function.
The only effect on systems, structures and components (SSCs) by this change is that one DG battery with one cell less than 1.36 volts the system will still be considered operable. With one or more DG batteries with one or more battery cell(s) not within limits of level or temperature, sufficient capacity to supply the required load for the DG is not assumed, and the corresponding DC electrical power subsystem must be declared inoperable immediately. With one or more DG batteries with two or more battery cells not within limits of voltage, sufficient capacity to supply the required load for the DG is not assumed, and the corresponding DC electrical power subsystem must be declared inoperable immediately.
Surveillance (SR) 3.8.4.2 is being relocated to TS 3.8.6 as a new surveillance and the wording of the Bases section is being revised for clarity as follows: “For this surveillance, a minimum of two cells shall be tested every seven days. The cells selected for testing shall be rotated on a monthly basis.” The new SR 3.8.6.5 will check the DG battery cell voltage on selected cells to ensure they are greater than or equal to 1.36 volts on a seven day frequency. This test will continue to assure that the batteries are available to perform their design functions.
This amendment will not change any previously evaluated accidents such as “Loss of Non-Emergency AC Power to Station Auxiliaries (Blackout)”, “Loss of Coolant Accident (LOCA),” and “LOCA/Blackout.” The prevention and mitigation of these accidents is also not affected by this change.
The likelihood of a malfunction of the batteries is not increased by this change in the surveillances. The systems will continue to be able to perform their design functions of supplying emergency power during the evaluated accidents listed above. Therefore, the changes do 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?
No. This change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The change does not alter assumptions made in the safety analysis or licensing basis. This change will not affect or degrade the ability of the DC Electrical Power Systems to perform their specified safety function. Therefore, the change does not create the possibility of a new or different kind of credible accident from any accident previously evaluated.
(3) Does the proposed change involve a significant reduction in a margin of safety?
No. Assuming that one cell in a 94-cell battery is at a full-reverse voltage of −1.80V, the remaining cells would be required to supply 106.80V, or 1.1484V/cell, in order to maintain a minimum battery terminal voltage of 105.0V. The manufacturer has extrapolated new sizing factors for an end-voltage of 1.1484V and used the new sizing factors to recalculate the battery capacity required to satisfy the design basis requirements. The load profile data and sizing methodology was taken from 125 Vdc Diesel Auxiliary Power Battery Sizing Calculations. Considering all possible loading scenarios, the minimum capacity margin available with one cell assumed to be in full reversal (−1.80V) was calculated to be 34%. This assumes the battery is at an end-of-life capacity of 80%, the electrolyte temperature is at the design-minimum of 60 °F, and that no cells are jumpered out.
Based on the discussion above and the results of the battery sizing calculations, a DG battery remains operable and fully capable of satisfying its design requirements with one cell < 1.36V on an indefinite basis. Therefore, the proposed changes listed above do not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department (PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, North Carolina 28201-1006.
NRC Section Chief: Mary Jane Ross-Lee, Acting.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: September 29, 2003.
Description of amendment request: The proposed amendment would revise the Technical Specification (TS) 3.7.15 spent fuel pool (SFP) storage criteria based upon fuel type, fuel enrichment, burnup, cooling time and partial credit for soluble boron in the SFP. This amendment allows for the safe storage of fuel assemblies with a nominal enrichment of Uranium-235 up to 5.00 weight percent. In addition, this amendment decreases the required soluble boron credit, which provides an acceptable margin of subcriticality in the McGuire Nuclear Station (McGuire), Units 1 and 2, spent fuel storage pools.
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
This license amendment transitions the McGuire SFP from conformance with a temporary exemption to 10 CFR 70.24 to compliance with 10 CFR 50.68(b). This regulation requires that the SFP remain subcritical if flooded with unborated water and remain 5 percent subcritical with credit for soluble boron. The SFP will be maintained with a minimum TS required soluble boron concentration that would provide substantial margin to criticality. The criticality analysis takes into consideration fuel type, fuel enrichment, fuel burnup, spent fuel cooling time and partial credit for soluble boron.
There is no significant increase in the probability or consequence of a fuel assembly drop accident in the SFP as a result of this amendment. The method of handling fuel assemblies in the SFP is not affected by the changes made to the criticality analysis for the SFP or by the TS changes. The handling of fuel assemblies during normal operation is unchanged, since the same equipment and procedures will be used.
There is no significant increase in the probability or consequence of the accidental misloading of spent fuel assemblies. Fuel assembly placement and storage will be controlled in accordance with approved fuel handling procedures and other approved processes to ensure compliance with the TS requirements. Analyses demonstrate that the pool will remain subcritical following an accidental misloading because the SFP contains an adequate margin of soluble boron concentration.
The mitigating actions as the result of a loss of SFP cooling are not changed. The heat up rate in the SFP is a nearly linear function of the fuel decay heat load. The fuel decay heat load will not be significantly affected since the number of fuel assemblies and the fuel burnups are unchanged. In the unlikely event that all pool cooling is lost, sufficient time will still be available for the operators to provide alternate means of cooling before the onset of pool boiling.
A decrease in pool water temperature from a large emergency makeup would cause an increase in water density, increasing fuel bundle reactivity. However, the margin provided by the TS required minimum boron concentration, above the concentration required to maintain 5 percent subcritical, will compensate for the increased fuel bundle reactivity which could result from a decrease in SFP water temperature.
Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated
This license amendment regarding fuel storage requirements, nominal fuel enrichment, and the credit for soluble boron in the SFP specified by TS 4.3 will have no effect on normal pool operations and maintenance. There are no changes in equipment design or in plant configuration.
Criticality and other SFP accidents have been analyzed in the McGuire's Updated Final Safety Analysis Report and Criticality Analysis reports. Specific accidents considered and evaluated include fuel assembly drop, accidental misloading, and significant changes in SFP water temperature. Region 1 of the SFP for both units had previously been updated with new replacement in-kind fuel racks utilizing boral neutron poison. As a result of this amendment no credit will be taken for the degrading boraflex neutron poison in Region 2 of the SFP.
Therefore, the proposed amendment will not result in the possibility of a new or different kind of accident previously evaluated.
Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety
The proposed TS changes and the resulting spent fuel storage operating limits will provide adequate safety margin to ensure that the stored fuel assembly array will always remain subcritical. Those limits are based on a plant-specific criticality analysis. This methodology takes partial credit for soluble boron in the SFP and requires conformance with 10 CFR 50.68(b).
Therefore, the proposed changes in this license amendment will not result in a significant reduction in the facility's margin of safety.
Therefore, 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: Mary Jane Ross-Lee, Acting.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas
Date of amendment request: June 24, 2004.
Description of amendment request: The proposed amendment would modify the Safety Analysis Report (SAR) by increasing the maximum hypothetical accident (MHA) doses to the control room operators, due to an increase in the allowable unfiltered in-leakage into the control room envelope. However, the new MHA doses would still be within NRC-approved guidance.
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 adopt new dose acceptance criteria in Regulatory Guide 1.195 for calculating radiological consequences of design basis accidents. The proposed change increases the allowable unfiltered inleakage to 52 scfm [standard cubic feet per minute] which increases the licensing basis thyroid doses for ANO [Arkansas Nuclear One] operators to 49.9 rem for the ANO-1 [Arkansas Nuclear One, Unit 1] Safety Analysis Report MHA. The new MHA doses are within NRC approved guidance. The proposed change does not impact the probability of an accident previously evaluated in the SAR.
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 accident analysis performed in establishing [the] new control room unfiltered inleakage value of 52 scfm were primarily performed using the existing licensing basis for the ANO-1 SAR. However, a new thyroid dose acceptance criterion of 50 rem was used per Regulatory Guide 1.195 instead of the previous Standard Review Plan thyroid dose limit of 30 rem. Dose consequences of non-LOCA [non-loss-of-coolant accident] events (except for the Fuel Handling Accident) were not historically calculated in the ANO-1 SAR. The doses had been assumed to be a fraction of the doses resulting from the MHA. New analyses of these control room doses confirmed them to be bounded by the revised MHA control room doses.
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.
Even though the ANO-1 SAR reported doses for the MHA are being increased in the proposed change, they are still within the NRC acceptance criteria of Regulatory Guide 1.195. Other assumptions are consistent with the current ANO-1 licensing basis or previously NRC approved assumptions within the industry. The increase in allowable in leakage by the proposed change maintains the operator doses within GDC [General Design Criteria] 19 limits with no compensatory measures to reduce thyroid uptake.
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., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: June 28, 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 (NSHC) 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 June 28, 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: M. S. Ross, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
Acting NRC Section Chief: Daniel S. Collins.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey
Date of amendment request: June 7, 2004.
Description of amendment request: The proposed changes would reflect an expanded operating domain resulting from implementation of Average Power Range Monitor/Rod Block Monitor/Technical Specifications/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA). The average power range monitor (APRM) flow-biased flux scram setpoint and the APRM and rod block monitor (RBM) flow-biased rod block trip setpoints would be revised to permit operation in the MELLLA region. In addition, the APRM scram and rod block trip setdown requirement would be replaced by more direct power and flow-dependent thermal limits to reduce the need for APRM gain adjustments and to allow more direct thermal limits administration during operation at other than rated conditions. The amendment would also change the methods used to evaluate annulus pressurization and jet loads resulting from the postulated recirculation suction line break.
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. The NRC staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The staff's review is presented below:
1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Proposed Change of APRM/RBM Setpoints
The APRM and RBM are not involved in the initiation of any accident and the APRM flow-biased simulated thermal power scram and rod block functions are not credited in any Hope Creek Generating Station safety analyses. The revised evaluation of the rod withdrawal error event will continue to demonstrate acceptable results without crediting operation of the RBM. Therefore, the proposed change would have no effect on the probability of an accident previously evaluated, and the increase in consequences of a previously-evaluated accident, if any, would not be significant.
Proposed Replacement of APRM Scram and Rod Block Trip Setdown Requirements by More Direct Power and Flow Dependent Thermal Limits
Neither the APRM scram and rod block setdown requirements, nor the power and flow-dependent thermal limits have any impact on accident initiating mechanisms. Adjustments to the thermal limits will be made using NRC-approved methods such that the fuel thermal and mechanical design bases will be maintained. Therefore, the proposed change will have no effect on the probability of an accident previously evaluated, and because the design bases will be maintained, an increase in the consequences of a previously-evaluated accident, if any, would not be significant.
Proposed Change in the Methods Used To Evaluate Annulus Pressurization and Jet Loads Resulting From the Postulated Recirculation Suction Line Break
The proposed change would modify the method of accident analysis for selected scenarios, and as such could have no impact on the probability of an accident previously evaluated. Since the loads resulting from the recirculation suction line break are demonstrated to be bounded by the current licensing basis, the increase in consequences of a previously-evaluated accident, if any, would not be significant.
2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Proposed Change of APRM/RBM Setpoints
Changing the formulation of the flow-biased APRM rod block and scram trip setpoints and the RBM flow biased rod block trip setpoint would not change their respective functions and manner of operation. The change would not introduce a sequence of events or introduce a new failure mode that would create a new or different type of accident. Operating within the expanded power flow map would not require any systems, structures or components to function differently. Therefore, the proposed change would not create the possibility of a new or different kind of accident from any accident previously evaluated.
Proposed Replacement of APRM Scram and Rod Block Trip Setdown Requirements by More Direct Power and Flow Dependent Thermal Limits
The replacement of the APRM scram and rod block trip setdown requirements by power and flow dependent thermal limits will continue to maintain the mechanical and thermal fuel design bases. Given that these design bases will be maintained, the proposed change would not create the possibility of a new or different kind of accident from any previously evaluated.
Proposed Change in the Methods Used To Evaluate Annulus Pressurization and Jet Loads Resulting From the Postulated Recirculation Suction Line Break
The proposed change to the methods of analysis does not change the design function or operation of any plant equipment. Therefore, the proposed change would 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?
Proposed Change of APRM/RBM Setpoints
The minimum critical power ratio (MCPR) and maximum average planar linear heat generation rate (MAPLHGR) thermal limits will be developed to ensure that the fuel thermal and mechanical design bases shall be maintained. Operation in the expanded operating domain would not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. Given that the proposed change will continue to meet the current design basis, any reduction in a margin of safety would not be significant.
Proposed Replacement of APRM Scram and Rod Block Trip Setdown Requirements by More Direct Power and Flow Dependent Thermal Limits
Replacement of the APRM setpoint requirements with power- and flow-dependent adjustments to the MCPR and MAPLHGR or LHGR thermal limits will continue to ensure that margins to the fuel cladding safety limit are preserved during operation at other than rated conditions. The fuel cladding safety limit will continue to be bounding for any anticipated operational occurrence. The flow and power dependent adjustments will continue to ensure that all fuel thermal and mechanical design bases shall remain bounding. The 10 CFR 50.46 acceptance criteria for the performance of the emergency core cooling system following postulated loss-of-coolant accidents will continue to be met. Therefore, any reduction in a margin of safety would not be significant.
Proposed Change in the Methods Used To Evaluate Annulus Pressurization and Jet Loads Resulting From the Postulated Recirculation Suction Line Break
The proposed change in methods shows that the loads from a postulated recirculation suction line break would be bounded by the current design basis loads. Therefore, any reduction in a margin of safety would not be significant.
Based on this review, it appears that the three standards of 10 CFR 50.92”) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama
Date of amendment request: July 8, 2004.
Description of amendment request: The proposed amendment would delete requirements from the Technical Specifications (TS) to maintain 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 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 [not installed at Browns Ferry and, therefore, not addressed by this proposed amendment] 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 model NSHC determination in its application dated July 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 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 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 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 and oxygen monitor equipment was intended to mitigate a design-basis hydrogen release. The 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 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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Acting Section Chief: Michael L. Marshall, Jr.
Previously Published Notices of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan
Date of amendment request: July 30, 2004.
Brief description of amendment request: The proposed amendment would (1) add License Condition 2.C.(22) requiring an integrated tracer gas test of the control room envelope using methods described in American Society for Testing and Materials E741-00, “Standard Test Method for Determining Air Change in a Single Zone by Means of a Tracer Gas Dilution,” and (2) delete Surveillance Requirement 3.7.3.6, which requires verification that unfiltered inleakage from control room emergency filtration system duct work outside the control room envelope is within limits.
Date of publication of individual notice in Federal Register: August 13, 2004 (68 FR 50217).
Expiration date of individual notice: October 12, 2004.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for A Hearing in connection with these actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated.
For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html . If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or by email to pdr@nrc.gov.
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: December 23, 2003.
Brief description of amendment: The amendment clarified the requirements for inoperable core spray (CS) system components, rendered inoperable CS component verification requirements consistent with each other, and modified the location requirement of stored water during periods of CS system inoperability.
Date of Issuance: August 19, 2004.
Effective date: August 20, 2004, and shall be implemented within 60 days of issuance.
Amendment No.: 247.
Facility Operating License No. DPR-16: Amendment revised the Technical Specifications.
Date of initial notice in Federal Register: January 20, 2004 (69 FR 2738).
The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated August 19, 2004.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina
Date of application for amendment: February 4, 2004, as supplemented by letter dated June 9, 2004.
Brief description of amendment: This amendment revises Technical Specification (TS) Surveillance Requirement 4.4.1.3.2, “Reactor Coolant System Hot Shutdown Surveillance Requirements,” and Limiting Condition for Operation 3.4.1.4.1.b, “Reactor Coolant System Cold Shutdown—Loops Filled Limiting Condition For Operation,” by eliminating a requirement that the wide-range instrumentation be inoperable before the narrow-range instrumentation can be used for confirmation of the minimum steam generator secondary side water level. The amendment also revises the TS Index to restore consistency with other sections of the TS.
Date of issuance: August 16, 2004.
Effective date: August 16, 2004.
Amendment No.: 116.
Facility Operating License No. NPF-63 . Amendment revises the Technical Specifications.
Date of initial notice in Federal Register: March 16, 2004 (69 FR 12365). The June 9, 2004, supplement provided clarifying information only and did not change the initial no proposed significant hazards consideration determination or expand the scope of the initial application.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 16, 2004.
No significant hazards consideration comments received: No.
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.
Date of application for amendments: June 27, 2003, as supplemented by letters dated January 29, 2004, March 3, 2004, June 4, 2004, and August 11, 2004.
Brief description of amendments: The amendments revise TS 3.4.10, “Pressurizer Safety Valves,” by changing the existing pressurizer safety valve lift settings from “≥2460 psig and ≤2510 psig,” to “≥2411 psig and ≤2509 psig.”
Date of issuance: August 26, 2004.
Effective date: As of the date of issuance and shall be implemented within 30 days.
Amendment Nos.: 138/138, 131/131.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 30, 2003 (68 FR 56343).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 26, 2004.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska
Date of amendment request: December 29, 2003, as supplemented by letters dated March 8 and June 8, 2004.
Brief description of amendment: The amendment revises the following: (1) Incorporates into the Updated Safety Analysis Report (USAR) the overall main steam isolation valve leakage pathway configuration (including the post-accident manual actions necessary to establish that configuration), (2) incorporates into the Cooper Nuclear Station licensing basis the loss-of-coolant accident (LOCA) dose calculation methodology (previously approved on an interim basis), and (3) deletes License Condition 2.C.(6), eliminating the commitment to provide potassium iodide to the control room personnel during LOCA conditions with core damage.
Date of issuance: September 1, 2004.
Effective date: As of the date of issuance and shall be implemented within 60 days of issuance.
Amendment No.: 206.
Facility Operating License No. DPR-46: Amendment revises the USAR and Operating License.
Date of initial notice in Federal Register: March 2, 2004 (69 FR 9861).
The March 8 and June 8, 2004, supplemental letters provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 1, 2004.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-282, Prairie Island Nuclear Generating Plant, Unit 1, Goodhue County, Minnesota
Date of application for amendment: August 27, 2003, as supplemented December 16, 2003, March 22, 2004, and July 19, 2004.
Brief description of amendment: The amendment revises Technical Specification 5.5.14 to allow the licensee to perform post-modification testing of the containment pressure boundary following steam generator replacement in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI, instead of 10 CFR Part 50, Appendix J, Option B. The steam generator replacement is scheduled for fall 2004.
Date of issuance: August 20, 2004.
Effective date: As of the date of issuance and shall be implemented within 30 days.
Amendment No.: 165.
Facility Operating License No. DPR-42: Amendment revises the Technical Specifications.
Date of initial notice in Federal Register: January 20, 2004 (69 FR 2744).
The March 22 and July 19, 2004, supplemental letters provided clarifying information that was within the scope of the original amendment request and did not change the Nuclear Regulatory Commission staff's initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 20, 2004
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit No. 1, Fairfield County, South Carolina
Date of application for amendment: July 23, 2003.
Brief description of amendment: Revised the near end-of-life Moderator Temperature Coefficient (MTC) Surveillance Requirement 4.1.1.3.b by placing a set of conditions on core operation, which if met, would allow exemption from the required MTC measurement. The conditional exemption is determined on a cycle-specific basis by considering the margin predicted to the surveillance requirement MTC limit and the performance of other core parameters, such as beginning of life MTC measurements and the critical boron concentration as a function of cycle life.
Date of issuance: July 21, 2004.
Effective date: As of the date of issuance and shall be implemented within 30 days.
Amendment No.: 169.
Renewed Facility Operating License No. NPF-12: Amendment revises the Technical Specifications.
Date of initial notice in Federal Register: September 30, 2003 (68 FR 56346).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 21, 2004.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: November 4, 2003, as supplemented by letter dated June 29, 2004.
Brief description of amendments: The amendments revise the South Texas Project, Units 1 and 2 Technical Specifications for the Remote Shutdown System to reflect requirements consistent with those in NUREG-1431, “Standard Technical Specifications—Westinghouse Plants.” The changes increase the allowed outage time for inoperable Remote Shutdown System components to a time that is more consistent with their safety significance and relocate the description of the required components to the Bases where it will be directly controlled by the licensee.
Date of issuance: August 20, 2004.
Effective date: As of the date of issuance and shall be implemented within 60 days of issuance.
Amendment Nos.: Unit 1-163; Unit 2-152.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: November 25, 2003 (68 FR 66140). The supplement dated June 29, 2004, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 20, 2004.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket No. 50-338, North Anna Power Station, Unit 1, Louisa County, Virginia
Date of application for amendment: March 28, 2002, as supplemented by letters dated May 13, June 19, July 9, July 25, August 2, August 16, and November 15, 2002, May 6, May 9, May 27, June 11 (2 letters), July 18, August 20, August 26, September 4, September 5, September 22, September 26 (2 letters), November 10, December 8, and December 17, 2003, and January 6, January 22 (2 letters), February 12, February 13, March 1, June 16, and June 18 (2 letters), 2004. The November 15, 2002, submittal replaced the submittals dated July 9, July 25, and August 16, 2002.
Brief description of amendment: This amendment revises Improved Technical Specification Sections 2.1, 4.2, and 5.6.5 in order to allow Virginia Electric and Power Company to implement Framatome ANP Advanced Mark-BW fuel at North Anna Power Station, Unit 1.
Date of issuance: August 20, 2004.
Effective date: As of the date of issuance and shall be implemented prior to the initiation of core onload during Refueling Outage 17 (Fall 2004).
Amendment No.: 237.
Renewed Facility Operating License No. NPF-4: Amendment changes the Technical Specifications.
Date of initial notice in Federal Register: July 22, 2003 (68 FR 43397). The supplements dated July 18, August 20, August 26, September 4, September 5, September 22, September 26 (2 letters), November 10, December 8, and December 17, 2003, and January 6, January 22 (2 letters), February 12, February 13, March 1, June 16, and June 18 (2 letters), 2004, contained clarifying information only and did not change the initial no significant hazards consideration determination or expand the scope of the initial application.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 20, 2004.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and Final Determination of No Significant Hazards Consideration and Opportunity for a Hearing (Exigent Public Announcement or Emergency Circumstances)
During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application for the amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the date the amendment was needed, there was not time for the Commission to publish, for public comment before issuance, its usual Notice of Consideration of Issuance of Amendment, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a Federal Register notice providing opportunity for public comment or has used local media to provide notice to the public in the area surrounding a licensee's facility of the licensee's application and of the Commission's proposed determination of no significant hazards consideration. The Commission has provided a reasonable opportunity for the public to comment, using its best efforts to make available to the public means of communication for the public to respond quickly, and in the case of telephone comments, the comments have been recorded or transcribed as appropriate and the licensee has been informed of the public comments.
In circumstances where failure to act in a timely way would have resulted, for example, in derating or shutdown of a nuclear power plant or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level, the Commission may not have had an opportunity to provide for public comment on its no significant hazards consideration determination. In such case, the license amendment has been issued without opportunity for comment. If there has been some time for public comment but less than 30 days, the Commission may provide an opportunity for public comment. If comments have been requested, it is so stated. In either event, the State has been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an amendment immediately effective, notwithstanding the pendency before it of a request for a hearing from any person, in advance of the holding and completion of any required hearing, where it has determined that no significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated.
For further details with respect to the action see (1) the application for amendment, (2) the amendment to Facility Operating License, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment, as indicated. All of these items are available for public inspection at the Commission's Public Document Room, 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/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to pdr@nrc.gov.
The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are problems in accessing the document, contact the PDR Reference staff at 1-800-397-4209, 301-415-4737, or by e-mail to pdr@nrc.gov. If a request for a hearing or petition for leave to intervene is filed by the above date, 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 identify 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 intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner 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 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.
To the extent that the applications contain attachments and supporting documents that are not publicly available because they are asserted to contain safeguards or proprietary information, petitioners desiring access to this information should contact the applicant or applicant's counsel and discuss the need for a protective order.
Each contention shall be given a separate numeric or alpha designation within one of the following groups:
1. Technical—primarily concerns/issues relating to technical and/or health and safety matters discussed or referenced in the applications.
2. Environmental—primarily concerns/issues relating to matters discussed or referenced in the environmental analysis for the applications.
3. Miscellaneous—does not fall into one of the categories outlined above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors seek to co-sponsor a contention, the petitioners/requestors shall jointly designate a representative who shall have the authority to act for the petitioners/requestors with respect to that contention. If a petitioner/requestor seeks to adopt the contention of another sponsoring petitioner/requestor, the petitioner/requestor who seeks to adopt the contention must either agree that the sponsoring petitioner/requestor shall act as the representative with respect to that contention, or jointly designate with the sponsoring petitioner/requestor a representative who shall have the authority to act for the petitioners/requestors with respect to that contention.
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. Since the Commission has made a final determination that the amendment involves no significant hazards consideration, if a hearing is requested, it will not stay the effectiveness of the amendment. Any hearing held would take place while the amendment is in effect.
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 or 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).
Florida Power and Light Company, Docket No. 251, Turkey Point Plant, Unit 4, Miami-Dade County, Florida
Date of amendment request: July 28, 2004, as supplemented in a letter dated August 5, 2004.
Description of amendment request: The amendment revised Technical Specifications 3/4.1.3.1, 3/4.1.3.2 and 3/4.1.3.5 to allow the use of an alternate method of determining rod position for the control rod F-8, until the end of Cycle 22 or until repairs can be conducted on the Analog Rod Indication System at the next outage of sufficient duration, whichever comes first.
Date of issuance: August 20, 2004.
Effective date: As of the date of issuance and shall be implemented within 60 days of issuance.
Amendment No.: 221.
Facility Operating License No. (DPR-41): Amendment revised the Technical Specifications.
Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. August 5, 2004 (69 FR 47467). The licensee's August 5, 2004 submittal of supplemental information did not affect the original no significant hazards consideration determination, and did not expand the scope of the request as noticed on August 5, 2004. The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments have been received. The notice also provided an opportunity to request a hearing by August 19, 2004, but indicated that if the Commission makes a final NSHC determination, any such hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of exigent circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated August 20, 2004.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Michael L. Marshall, Jr. (Acting).
Dated at Rockville, Maryland, this 3rd day of September 2004.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 04-20497 Filed 9-13-04; 8:45 am]
BILLING CODE 7590-01-P