# Biweekly Notice Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations > Briefs, arguments, decisions, and more. URL: https://www.frixlaw.com/law-library/documents/fr%3AX94-10119 ## Record - **Collection:** Federal Register - **Document type:** Uncategorized Document - **Published:** January 19, 1994 ## Text NUCLEAR REGULATORY COMMISSION Biweekly Notice Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations I. Background Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. Public Law 97-415 revised section 189 of the Atomic Energy Act of 1954, as amended (the Act), to require the Commission to publish notice of any amendments issued, or proposed to be issued, under a new provision of section 189 of the Act. This provision grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. This biweekly notice includes all notices of amendments issued, or proposed to be issued from December 23, 1993, through January 6, 1994. The last biweekly notice was published on January 5, 1994 (59 FR 615). 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. Normally, the Commission will not issue the amendment until the expiration of the 30-day notice period. However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30-day notice period, provided that its final determination is that the amendment involves no significant hazards consideration. The final determination will consider all public and State comments received before action is taken. Should the Commission take this action, it will publish in the Federal Register a notice of issuance and provide for opportunity for a hearing 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 Rules Review and Directives Branch, Division of Freedom of Information and Publications Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room P-223, Phillips Building, 7920 Norfolk Avenue, Bethesda, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the NRC Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555. The filing of requests for a hearing and petitions for leave to intervene is discussed below. By February 18, 1994, 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 2.714 which is available at the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555 and at the local public document room for the particular facility involved. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order. As required by 10 CFR 2.714, 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 factors: (1) the nature of the petitioner's right under the Act to be made a party to the proceeding; (2) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (3) the possible effect of any order which may be entered in the proceeding on the petitioner's interest. The petition should also identify the specific aspect(s) of the subject matter of the proceeding as to which petitioner wishes to intervene. Any person who has filed a petition for leave to intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to 15 days prior to the first prehearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above. Not later than 15 days prior to the first prehearing conference scheduled in the proceeding, a petitioner shall file a supplement to the petition to intervene which must include a list of the contentions which are sought to be litigated in the matter. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner shall provide a brief explanation of the bases of 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. Petitioner must provide 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 who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross- examine witnesses. If a hearing is requested, 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 with the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Docketing and Services Branch, or may be delivered to the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the above date. Where petitions are filed during the last 10 days of the notice period, it is requested that the petitioner promptly so inform the Commission by a toll-free telephone call to Western Union at 1- (800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator should be given Datagram Identification Number N1023 and the following message addressed to (Project Director): petitioner's name and telephone number, date petition was mailed, plant name, and publication date and page number of this Federal Register notice. A copy of the petition should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and to the attorney for the licensee. Nontimely filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for a hearing will not be entertained absent a determination by the Commission, the presiding officer or the Atomic Safety and Licensing Board that the petition and/or request should be granted based upon a balancing of factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d). For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555, and at the local public document room for the particular facility involved. Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 1, 2, and 3, Maricopa County, Arizona Date of amendment requests: October 26, 1993 Description of amendment requests: The proposed amendment would change the maximum nominal enrichment of the fuel allowed to be used in the reactor core. Specifically, in Technical Specification 5.3.1, ``Fuel Assemblies,'' the fuel enrichment would change from ``a maximum enrichment of 4.05 weight percent U-235'' to ``a maximum radially averaged enrichment of 4.30 weight percent U-235 at any axial location.'' 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: Standard 1 -- Involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The adequacy of a given core design must be demonstrated for each core prior to core reloading. The fuel enrichment is only one of the factors that must be considered in this determination. The fuel enrichment does not directly impact the results of the plant safety analysis. The Unit 1, 2, and 3 fuel and storage areas have been analyzed for a maximum radially averaged enrichment of any axial enrichment zone with a fuel assembly of 4.30 w/o U-235. The criticality analysis for Palo Verde's spent fuel pool is presented in Section D.3, Criticality Safety Analysis. The results of these analyses indicate that handling and storage of such fuel assemblies do not involve an unreviewed safety question. The results of these analyses are within the acceptance criteria defined in TS 5.6.1, ``Criticality.'' The applicable codes, standards and regulations of criticality safety for spent fuel and new fuel storage include the following: - General Design Criterion 62 - Prevention of Criticality in Fuel Storage and Handling. - NUREG -0800, USNRC Standard Review Plan, Section 9.1.2, Spent Fuel Storage and Section 9.1.1, New Fuel Storage. - ANSI/ANS-57.2-1983, ``Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants,'' Section 6.4.2 - ANSI/ANS-57.3-1983, ``Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants,'' Section 6.2.4 - Qualification of Analytical Methodology Used In Spent Fuel Storage Rack Analyses, CE Benchmark, 590962-PHD-004 Revision 0 dated December 3, 1979. These regulations and guides require that for spent fuel racks the maximum calculated keff, including margin for uncertainty in calculational method and mechanical tolerances, be less than or equal to 0.95 with a 95% probability at a 95% confidence level. In order to assure the true reactivity will always be less than the calculated reactivity, the following conservative assumptions were made for spent fuel rack criticality analysis: - Pure, unborated water at 68 degrees Fahrenheit is used in all calculations, - An infinite array with no radial or axial leakage is modeled, and - Neutron absorption from spacer grids is neglected, i.e., replaced by water. For the new fuel vault, a dual criteria applies in which the maximum calculated keff, including uncertainties, is less than or equal to 0.95 when flooded and less than or equal to 0.98 under conditions of ``optimum moderation.'' Because the new fuel vault is normally dry, and low density moderation of ``optimum moderation'' produces strong coupling between assemblies, the following conservative assumptions are used: - The storage array was enclosed on all six sides by a tight fitting two foot concrete reflector, - Unborated water is introduced uniformly throughout the storage array and the space between fuel pins, - Water density is varied uniformly from flooded to dry, - Neutron absorption from spacer grids is neglected, i.e., replaced by water. In the new fuel vault criticality analysis, leakage is explicitly modeled, because the assumption of an infinite array with no radial or axial leakage is unrealistic under conditions of low density moderation. Leakage suppresses criticality at low moderator density. Without 3-D modeling of the array, erroneously high values of keff are calculated. Thus, the assumption on array leakage is relaxed, but reflection from the walls, floor and ceiling is included. In addition to the above discussion of the new fuel vault criticality analysis, the following conservative assumptions are applied to both analyses: - No credit is taken for the presence of burnable poison rods. These rods displace fuel rod positions and are an integral part of selected fuel assemblies. - An upper bound for the fuel density was included in the nominal configuration of the fuel densities. - The upper statistical bound of the fuel assembly enrichment, as based on the fuel fabrication specification, is included in the statistical evaluation of uncertainties. The criticality analysis of the Palo Verde new and spent fuel racks shows that the maximum radially averaged fuel enrichment of any axial enrichment zone within a fuel assembly which meets the appropriate NRC limit with uncertainties is higher than 4.30 w/o U- 235. Although a higher enrichment fuel cycle may result in fuel burnup consisting of a slightly different mixture of nuclides and inventory, the effect is insignificant because the isotopic mixture and inventory of an irradiated assembly is relatively insensitive to the fuel assembly's initial enrichment. Therefore, the doses from postulated accidents are not significantly affected and continue to be acceptable. Standard 2 -- Create the possibility of a new or different kind of accident from any accident previously analyzed. Operation of Palo Verde with the proposed enrichment limit change will not create any new or different kinds of accidents from those previously evaluated. The adequacy of a given core design shall be demonstrated for each core prior to reloading. The fuel enrichment is only one of the factors that must be considered in this determination. Fuel handling and storage of fuel with radially averaged enrichment of any axial enrichment zone within a fuel assembly of 4.30 w/o U-235 does not create the possibility of a new or different kind of accident from any accident previously evaluated. Standard 3 -- Involve a significant reduction in a margin of safety. This amendment request will not involve a significant reduction in a margin of safety. The evaluation performed for each reload core assures that the core design meets appropriate safety limits, including a consideration of a significant reduction in the margin of safety. See response provided in Standard 1 for information pertaining to the demonstration of the adequacy of each core design. Criticality analyses for fuel assemblies with a maximum radially averaged enrichment (of any axial enrichment zone) of 4.30% U-235 for the Palo Verde Fuel Pool configurations presented in this proposed Technical Specification amendment meet the criticality acceptance criterion for Keff listed in Technical Specification 5.6.1.1. Technical Specification 5.6.1.1 states: The spent fuel storage racks are designed and shall be maintained with: a. A Keff equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance of 2.6% delta k/k for uncertainties as described in Section 9.1 of the FSAR. . . . Based on the above evaluation, this proposed change does not constitute a significant hazards consideration. The NRC staff has reviewed the licensee's analysis and, based on that review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment requests involve no significant hazards consideration. Local Public Document Room location: Phoenix Public Library, 12 East McDowell Road, Phoenix, Arizona 85004 Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail Station 9068, Phoenix, Arizona 85072-3999 NRC Project Director: Theodore R. Quay Arizona Public Service Company, et al., Docket Nos. STN 50-528 and STN 50-529, Palo Verde Nuclear Generating Station, Units 1 and 2, Maricopa County, Arizona Date of amendment requests: October 27, 1993 Description of amendment requests: The proposed amendment would revise Technical Specification (TS) 6.9.1.8 to change the frequency for submitting the Radioactive Effluent Release Report from semiannual to annual, as allowed by the revised 10 CFR 50.36a requirements. 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: Standard 1 -- Involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes are administrative in nature and do not involve any change to the configuration or method of operation of any plant equipment that is used to mitigate the consequences of an accident. Also, the proposed changes do not alter the conditions or assumptions in any of the FSAR accident analyses. Since the FSAR accident analyses remain bounding, the radiological consequences previously evaluated are not adversely affected by the proposed changes. Therefore, it can be concluded that the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. Standard 2 -- Create the possibility of a new or different kind of accident from any accident previously analyzed. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes are administrative in nature and do not involve any change to the configuration or method of operation of any plant equipment that is used to mitigate the consequences of an accident. Accordingly, no new failure modes have been defined for any plant system or component important to safety nor has any new limiting failure been identified as a result of the proposed changes. Also, there will be no change in the types or increase in the amount of effluents released offsite. Therefore, it can be concluded that the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. Standard 3 -- Involve a significant reduction in a margin of safety. The proposed changes do not involve a significant reduction in a margin of safety. The proposed changes are administrative in nature and do not adversely impact the plant's ability to meet applicable regulatory requirements related to liquid and gaseous effluents, and solid waste releases. The proposed changes would also eliminate an unnecessary burden of governmental regulation without reducing protection for public health and safety. Therefore, it can be concluded that the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on that review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment requests involve no significant hazards consideration. Local Public Document Room location: Phoenix Public Library, 12 East McDowell Road, Phoenix, Arizona 85004 Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail Station 9068, Phoenix, Arizona 85072-3999 NRC Project Director: Theodore R. Quay Baltimore Gas and Electric Company, 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 8, 1993 Description of amendments request: As an active participant in the industry-NRC improved Standard Technical Specifications (STS) implementation effort, Baltimore Gas and Electric Company (BG&E) volunteered to develop criteria for determining what requirements are appropriate for inclusion in the Design Features section of the Technical Specifications (TSs), and to submit a lead plant license amendment applying those criteria. This proposed license amendment justifies the adoption of the ``Design Features'' section of the Combustion Engineering STS for the Calvert Cliffs Nuclear Power Plant and then applies the developed criteria to the STS. The proposed criteria for determining what requirements should be placed in the Design Features section of the TSs are: 1. The amount, kind, and source of special nuclear material required; 2. The place of the use of the special nuclear material; and 3. Those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have an immediate and significant effect on safety and are not covered in the safety limits, limiting conditions for operation or surveillance requirements of the Technical Specifications. The Design Features section is Section 5.0 for the Calvert Cliffs TS and Section 4.0 for the STS. Specifically, the following changes are proposed to the Calvert Cliffs TSs and the STS for the Design Features section. Section 5.1 Site Sections 5.1.1 and 5.1.2 contain maps of the site boundary and low population zones, respectively. These maps also show the major structures, effluent release points, meteorological tower location, and the minimum exclusion area radius. The STS requires inclusion of the site and exclusion area boundaries and the low population zone and allows descriptions instead of maps. These maps or descriptions do not meet any of the criteria. However, Criteria 2 requires that the place of use of the special nuclear material be described. BG&E proposes to delete the existing Sections 5.1.1 and 5.1.2 and to create Section 5.1, entitled ``Site Location,'' and to include a text description of the location of the site. The current information and maps in the sections will be relocated to the Updated Final Safety Analysis Report (UFSAR). This change is also proposed for Section 4.1 of the STS. Section 5.2 Containment This section does not meet Criteria 1, 2 or 3 in that modification of the containment would not create an immediate and significant effect on safety. Furthermore, containment integrity is covered in the Limiting Conditions for Operation and Surveillance Requirements sections of the TS. This Section is not included in the STS. BG&E proposes to eliminate it from the Calvert Cliffs Design Features. Section 5.3 Reactor Core Calvert Cliffs TS Section 5.3 and corresponding STS Specification 4.2.1 and meets Criteria 1, e.g., state the amount, kind, and source of special nuclear material. For consistency with the STS, BG&E proposes to adopt the STS titles and wording. Section 5.3.3 Control Element Assemblies Calvert Cliffs Section 5.3.3, ``Control Element Assemblies,'' and the corresponding STS Section 4.2.2 does not meet Criteria 1, 2 or 3. The safety significant aspects of control rods, e.g., the reactivity worth of control rods and their required insertion times, are included in other portions of the TSs. Therefore, control rods do not fall under the Criteria 3 as features not described in other sections of the TSs. BG&E proposes that this section be eliminated from the Calvert Cliffs Design Features and the STS. Section 5.4 Reactor Coolant System This section does not meet any of the criteria and is not included in the STS. It does not meet Criteria 3 in that the requirements on degradation, pressure, and temperature are contained in other portions of the TS and changes in the total water and steam volume would not have an immediate and significant impact on safety. BG&E proposes that it be eliminated from the Calvert Cliffs Technical Specifications. Section 5.5 Meteorological Tower Location This section does not meet any of the criteria and is not included in the STS. BG&E proposes to eliminate this section from the Calvert Cliffs Design Features section. Section 5.6 Fuel Storage Calvert Cliffs Section 5.6.1, ``Criticality - Spent Fuel,'' specifies the minimum center-to-center distance, a Keff limit, and the maximum enrichment for fuel in the spent fuel storage racks. These requirements are also contained in STS Section 4.3.1.1, Items a, b, and c. Section 5.6.2, ``Criticality - New Fuel,'' specifies the minimum center-to-center distance, a keff limit, and a maximum enrichment for fuel in the new fuel racks. These requirements are contained in STS Section 4.3.1.2, Items a, b, c, and d. These sections meet Criteria 3, e.g., geometries which, if altered, would have an immediate and significant impact on safety. These requirements do not appear in other sections of the TS. The STS language contains the same restrictions as the Calvert Cliffs Design Features while introducing no new requirements. BG&E proposes adopting the STS language and we will add information on uncertainties of the referenced sections of the UFSAR. Section 5.6.3 Drainage This section in the Calvert Cliffs Design Features section and the corresponding STS Section 4.3.2 does not meet any of the criteria. It does not meet Criteria 3 in that it does not describe geometry or materials of construction and because the requirements are contained in another portion of the TS. BG&E proposes that this section be eliminated from the Calvert Cliffs Design Features section and from the STS. Section 5.6.4 Capacity Section 5.6.4, ``Capacity,'' states the maximum spent fuel storage capacity. This meets Criteria 1 in that it limits the amount of special nuclear material that may be stored on site. The Calvert Cliffs TS language varies slightly from the STS language in that it makes clear that the storage capacity limit applies to the combined storage pool for Units 1 and 2. Therefore, we propose to retain the Calvert Cliffs TS language. Section 5.7 Component Cyclic or Transient Limits This section is not included in the STS and does not meet any of the criteria. We propose that it be eliminated from the Calvert Cliffs Design Features section. 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. Would not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change to the Design Features section adopts language from the Design Features section of the Standard Technical Specifications (STS) for Combustion Engineering Plants (NUREG-1432, September, 1992), based on the Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (July 16, 1993). Some requirements in the current Technical Specifications have been eliminated or relocated to the UFSAR based on the STS as evidence that the NRC no longer considers those requirements to meet the criteria for Design Features in 10 CFR 50.36(c)(4). In some cases, we have proposed elimination of some requirements in the STS and the Calvert Cliffs Technical Specifications based on a determination that they have no legal or regulatory basis. We propose eliminating the ``Site'' sections, present in the Calvert Cliffs and the STS Design Features sections, which contain maps or descriptions of the site boundary and low population zone. There are no legal or regulatory requirements for including this information in the Technical Specifications and this information will be relocated to the UFSAR. We propose adding a ``Site Location'' section which contains a description of the Calvert Cliffs location as required by the Atomic Energy Act of 1954 [,as amended]. The ``Containment,'' ``Reactor Coolant System,'' ``Meteorological Tower Location,'' and ``Component Cyclic or Transients Limits'' sections do not meet the legal or regulatory requirements for inclusion in the Design Features section, are not included in the STS, and have been deleted. The information in these sections is or will be contained in the UFSAR and will be controlled under 10 CFR 50.59. The ``Control Element Assemblies'' and ``Fuel Storage - Drainage'' sections are included in the STS and Calvert Cliffs Technical Specifications but eliminated in this change as these sections do not meet any of the legal or regulatory requirements for the Design Features section. This information is contained in the UFSAR and will be controlled under 10 CFR 50.59. We propose adopting the STS wording for the ``Reactor Core'' and ``Fuel Storage'' sections with no changes in the present limits or controls. Some information in the Calvert Cliffs Technical Specifications is not contained in the STS ``Reactor Core'' section. This information is contained in the UFSAR and will be controlled under 10 CFR 50.59. All information eliminated from the Design Features section of the Technical Specifications is or will be located in the UFSAR and will be controlled under 10 CFR 50.59. The design and operation of the plant have not changed. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Would not create the possibility of a new or different type of accident from any accident previously evaluated. The proposed change does not represent a change in the configuration or operation of the plant. All information eliminated from the Technical Specifications will continue to be controlled under 10 CFR 50.59. All legal and regulatory requirements for the Design Features section continue to be met. Therefore, the proposed change does not create the possibility of a new or different type of accident from any accident previously evaluated. 3. Would not involve a significant reduction in a margin of safety. The proposed change does not represent a change in the configuration or operation of the plant. All information eliminated from the Technical Specifications will continue to be controlled under 10 CFR 50.59. All legal and regulatory requirements for the Design Features section continue to be met. 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendments request involves no significant hazards consideration. Local Public Document Room location: Calvert County Library, Prince Frederick, Maryland 20678. Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037. NRC Project Director: Robert A. Capra Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts Date of amendment request: December 10, 1993 Description of amendment request: The proposed amendment would extend existing plant surveillance intervals to 24 months from 18 months. This is the third of three submittals and it changes specific setpoints to accommodate a 24 month fuel cycle and provides a justification for extending the surveillance interval for those components and systems that are not related to instrument setpoint changes. 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 operation of Pilgrim Station in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. This submittal results in changes to various equipment surveillance intervals and related instrument calibration frequencies and setpoints. The impact of lengthening the current 18 month interval to 24 months was evaluated and identified no significant system or component degradation as a consequence of lengthening the interval to 24 months; therefore, systems and components will continue to perform their design function. In some cases, the 24 month interval required setpoint changes to ensure instrument drift associated with the extended interval would not result in exceeding an instrument's acceptable setpoint tolerance. In other cases, justification of an extended interval was not developed because the surveillance could be performed on-line. In these cases, the ``once/cycle'' surveillance requirement is changed to the currently allowed 18 months. The proposed changes were developed using the guidance provided in Generic Letter 91-04 and Note 1 of Table 4.2.A through 4.2.G of Pilgrim's technical specifications. The proposed changes do not degrade the performance or increase the challenges to the associated safety systems assumed to function in the accident analysis. The impact of lengthening the current calibration/functional test interval from 1 to 3 months for certain components was also evaluated. The evaluation used the guidance in Generic Letter 91-04 and Note 1 of Table 4.2.A through 4.2.G of Pilgrim's technical specifications. No significant system or component degradation was identified as a consequence of lengthening the interval to 3 months. The proposed changes do not affect the availability of equipment or systems required to mitigate the consequences of an accident, and do not affect the availability of redundant systems or equipment. The plant will continue to operate within the limits specified in the Core Operating Limits Report (COLR) and will continue to take the same actions if setpoint limits are exceeded. Therefore, both the proposed setpoint and non-setpoint changes do not significantly increase the probability or consequences of an accident previously evaluated. 2. The operation of Pilgrim Station in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously analyzed. The proposed changes with one exception, do not add or remove active components and, therefore, do not introduce failure mechanisms of a different type than those previously evaluated. In one case, the EDG breaker time delay relays will be replaced with more accurate relays to ensure the specified time sequence for starting and accepting the emergency load remains within specification for the extended cycle. The replacement relays will be similar in size, weight, voltage and temperature operating range as those being replaced; therefore, the possibility of a new or different kind of accident is not created. In addition, the surveillance test requirements and the way surveillance tests are performed will remain unchanged. Since the intended operation and function of the analyzed systems do not change as a result of the setpoint and non-setpoint analyses, no new initiators are introduced capable of initiating an accident that would render these systems unable to provide their required protection. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. The operation of Pilgrim Station in accordance with the proposed amendment will not involve a significant reduction in the margin of safety. Although the proposed Technical Specification changes will result in an increase in the interval between surveillance tests, the existing margins of safety are maintained through our proposed setpoint revisions. The proposed setpoint changes either increase the plant safety margin or maintain the existing margin and do not significantly impact the availability, performance, or intended function of the affected systems. In the case of non-setpoint changes, evaluation of the affected systems indicates lengthening the interval to 24 months does not have significant impact on performance. Therefore, the assumptions in Pilgrim's accident analyses are not impacted, and the proposed Technical Specification changes do not significantly reduce the margin of safety. The NRC staff has reviewed the licensee's analysis, and based on this review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Plymouth Public Library, 11 North Street, Plymouth, Massachusetts 02360. Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 800 Boylston Street, 36th Floor, Boston, Massachusetts 02199. NRC Project Director: Walter R. Butler Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy 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: March 26, 1993 Description of amendment request: The proposed amendment is a part of Commonwealth Edison Company's (CECo's) Technical Specification Upgrade Program (TSUP) to improve the quality of the current Technical Specifications (TS) for Dresden and Quad Cities. The proposed amendment would for both Dresden and Quad Cities, upgrade the requirements of Section 3.9/4.9, ``Auxiliary Electrical Systems,'' to include operating and shutdown Limiting Condition(s) for Operation (LCO) and Surveillance Requirement(s) (SR) that are consistent with the Standard Technical Specifications (STS) and later operating plant provisions. Within the upgrade to Section 3.9/4.9, Emergency Diesel Generator (EDG) reliability provisions are added to implement the recommendations of Generic Letter(s) 84-15 and 91-09; Information Notice(s) 84-69 and 91- 62; and Regulatory Guide 1.9, draft Revision 3. Other Generic Letters considered in the proposed TS include 83-26, 83-30, and 87-09. 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: Involve a significant increase in the probability or consequences of an accident previously evaluated because: In general, the proposed changes represent the conversion of current requirement to a more generic format, or the addition of requirements which are based on the current safety analyses. Implementation of these changes will provide increased reliability of equipment assumed to operate in the current safety analyses, or provide continued assurance that specified parameters remain within their acceptance limits, and as such, will not significantly increase the probability or consequences of a previously evaluated accident. Some of the proposed changes represent minor curtailments of the current requirements which are based on generic guidance or previously approved provisions for other stations. These proposed changes are consistent with the current safety analyses and have been previously determined to represent sufficient requirements for the assurance and reliability of equipment assumed to operate in the safety analyses, or provide continued assurance that specified parameters remain within their acceptance limits. As such, these changes will not significantly increase the probability or consequences of a previously evaluated accident. A.C. Sources - Operating: The proposed modifications for Section 3/4.9.A, ``A.C. Sources - Operating'', administratively incorporate the requirements of STS, where applicable to Dresden and Quad Cities Stations. Most deviations from the STS requirements are based upon generic guidance and other approved requirements at other sites. Dresden and Quad Cities Station are retaining the current seven day allowed outage time from their current specifications for loss of an EDG. However, an additional verification of EDG operability has been proposed for approximately midway through the seven day AOT. The additional verification and additional details in the surveillances will significantly improve the overall safety of both Dresden and Quad Cities Station. Both Dresden and Quad Cities Station's EDG history have shown them to be very reliable. This can be demonstrated by the excellent pass/fail rate observed during the monthly surveillance tests. As stated previously, the addition of several new STS enhancements to fuel storage and transfer requirements and other miscellaneous EDG surveillances recommended by ASTM codes will improve EDG reliability. Therefore, because the EDG's for Dresden and Quad Cities Station do not act as accident initiators, the probability of an accident previously evaluated for the sites is not increased by the incorporation of the proposed requirements. Other changes based upon STS guidance are more restrictive and limit operation of the site with respect to all A.C. power sources. A.C. power sources do not act as initiators of accidents. Therefore, the probability of an accident previously evaluated for the sites is also not increased by the incorporation of the STS requirements. The consequences of any previously evaluated accidents are not increased as more restrictions and limitations are added to the current versions of both Dresden and Quad Cities specifications. The retention of the seven day allowed outage times for the EDG's and the offsite power sources does not increase the consequences of any previously analyzed accident for both sites as these are the current requirements. Therefore, the consequences of any previously evaluated accident is not increased as a result of the proposed changes. A.C. Sources - Shutdown: The proposed modification for Section 3/4.9.B, ``A.C. Sources - Shutdown'', administratively incorporate the requirements of STS where applicable to Dresden and Quad Cities Station. The STS requirements add additional provisions not in the current Technical Specifications for EDG fuel storage capability that will reduce the consequences of a previously analyzed accident. The probability of any previously evaluated accident is reduced during shutdown by the additional STS restrictions proposed for fuel handling type of activities. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of any previously evaluated accident for Dresden and Quad Cities Station. D.C. Sources - Operating: The proposed modifications for Section 3/4.9.C, ``D.C. Sources - Operating'', incorporate the requirements of STS where applicable for Dresden and Quad Cities Stations. Dresden and Quad Cities are proposing to retain the current provisions specified in the current version of Quad Cities Technical Specifications that allow the 125 and 250 VDC systems to be out-of- service for a period up to 72 hours. This 72-hour AOT may be extended, when applied to the 125 VDC systems, for up to a maximum period of 7 days with both units operating if the alternate 125 VDC battery is operable. These changes introduce a difference when compared to the STS requirements. However, the STS requirements as applied to Dresden and Quad Cities battery systems would prove to be overly burdensome requiring dual unit shutdowns to perform most maintenance or testing activities. The additional batteries in the design of the stations, with their surveillances, charger requirements, and breaker verifications, compensate for any STS deviations. The proposed requirements are comparable to the existing requirements and AOTs for Quad Cities Station; therefore, the consequences of any previously evaluated accident are not increased. The proposed changes add additional surveillance requirements to the D.C. systems at Dresden and Quad Cities to enhance their reliability and operational readiness. This also ensures the consequences of any previously evaluated accident are not increased. Because the D.C. system is not assumed as an accident initiator, the probability of any previously evaluated accident is not increased. D.C. Sources - Shutdown: The proposed modifications for Section 3/4.9.D, ``D.C. Sources - Shutdown'', administratively incorporate the requirements of STS where applicable to Dresden and Quad Cities Station. The proposed changes add additional surveillance requirements and more explicitly clarify the LCO's. The additional provision for fuel handling type of activities reduces the probability of previously evaluated accidents from occurring. The additional surveillance activities also improve D.C. reliability and thus, reduce the probability of D.C. system unavailability and hence, reduce the consequences of previously evaluated accidents. Because the D.C. system is not considered as an accident initiator, the probability of any previously evaluated accident is not increased. Distribution - Operating: The proposed modifications for Section 3/4.9.E, ``Distribution - Operating'', administratively incorporate the requirements of STS where applicable to Dresden and Quad Cities Station. The proposed changes add additional surveillance requirements and LCO's. The proposed requirements/actions for the D.C. distribution system are retained to be consistent to the proposed AOTs for the D.C. system. The STS requirements as applied to the Dresden/Quad Cities D.C. distribution system would prove to be overly burdensome, requiring plant shutdowns to perform maintenance or testing activities. The additional distribution system surveillance and LCO's compensate for the STS deviations. The proposed requirements are comparable to the existing requirements and AOTs for Quad Cities Station; therefore, the consequences of any previously evaluated accident are not increased. The additional surveillances and STS-type requirements ensures the reliability and operational readiness of the Distribution System and ensures the consequences of any previously evaluated accident are not increased. Because the Distribution System is not assumed as an accident initiator, the probability of any previously evaluated accident is not increased. Distribution - Shutdown: The proposed modification for Section 3/4.9.F, ``Distribution - Shutdown'', administratively incorporate the requirements of STS where applicable for Dresden and Quad Cities Station. The proposed changes add additional requirements that ensure the consequences and the probability of any previously evaluated accident are not increased. RPS Power Monitoring: The proposed modifications for Section 3/ 4.9.G, ``RPS Power Monitoring''. administratively incorporate the requirements of STS for Dresden and Quad Cities Station. The proposed changes add additional requirements for Dresden and clarify the existing requirements at Quad Cities. Therefore, the consequences and the probability of any previously evaluated accident are not increased. Create the possibility of a new or different kind of accident from any previously evaluated because: In general, the proposed changes represent the conversion of current requirements to a more generic format, or the addition of requirements which are based on the current safety analyses. Others represent minor curtailments of the current requirements which are based on generic guidance or previously approved provisions for other stations. These changes do not involve revisions to the design of the station. Some of the changes may involve revision in the operation of the station; however, these provide additional restrictions which are in accordance with the current safety analyses, or are to provide for additional testing or surveillances which will not introduce new failure mechanisms beyond those already considered in the current safety analyses. The retention of the current AOTs for EDGs, offsite power sources, and DC systems maintain the existing assumptions from the current accident analyses; therefore, these changes will not create the possibility of a new or different kind of accident from any previously evaluated. The proposed changes for Dresden and Quad Cities Station's Technical Specification Section 3/4.9 are based on STS guidelines or later operating BWR plants' NRC accepted changes. These proposed changes have been reviewed for acceptability at the Dresden and Quad Cities Nuclear Power Stations considering similarity of system or component design versus the STS of later operating BWRs. No new modes of operation are introduced by the proposed changes, considering the acceptable operational modes in present specifications, the STS, or later operating BWRs. Surveillance requirements are changed to reflect improvements in technique, frequency of performance or operating experience at later plants. Proposed changes to action statements in many places add requirements that are not in the present technical specifications or adopt requirements that have been used successfully at other operating BWRs with designs similar to Dresden and Quad Cities. The proposed changes maintain at least the present level of operability. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. Involve a significant reduction in the margin of safety because: In general, the proposed changes represent the conversion of current requirements to a more generic format, or the addition of requirements which are based on the current safety analyses. Others represent minor curtailments of the current requirements which are based on generic guidance or previously approved provisions for other stations. Some of the later individual items may introduce minor reductions in the margin of safety when compared to the current requirements. However, other individual changes are the adoption of new requirements which will provide significant enhancement of the reliability of the equipment assumed to operate in the safety analyses, or provide enhanced assurance that specified parameters remain with their acceptance limits. These enhancements compensate for the individual minor reductions, such that taken together, the proposed changes will not significantly reduce the margin of safety. The proposed changes to the Technical Specification Section 3/ 4.9 implement present requirements, or the intent of present requirements in accordance with the guidelines set forth in the STS. The proposed changes are intended to improve readability, usability, and the understanding of technical specification requirements while maintaining acceptable levels of safe operation. The proposed changes have been evaluated and found to be acceptable for use at Dresden and Quad Cities based on system design, safety analyses requirements and operational performance. The retention of the current AOTs for EDGs, offsite power sources, and DC systems maintain the existing assumptions from the current accident analyses. Since the proposed changes are based on NRC accepted provisions at other operating plants that are applicable at Dresden and Quad Cities and maintain necessary levels of system, component or parameter readability, the proposed changes do not involve a significant reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: for Dresden, the Morris Public Library, 604 Liberty Street, Morris, Illinois 60450, and for Quad Cities, the Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021 Attorney for licensee: Michael I. Miller, Esquire; Sidley and Austin, One First National Plaza, Chicago, Illinois 60690 NRC Project Director: James E. Dyer Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina Date of amendment request: November 11, 1993 Description of amendment request: The proposed amendments would consolidate the Quality Verification Department with the Nuclear Generation Department and realign the Nuclear Safety Review Board such that it reports to the Senior Vice-President of the Nuclear Generation Department. 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 amendments do not involve a significant increase in the probability or consequences of an accident previously evaluated.] The proposed revisions to consolidate the Quality Verification Department with the Nuclear Generation Department and realign the NSRB [Nuclear Safety Review Board] such that it reports to the Senior Nuclear Officer, change the reference from Semiannual to Annual, change the reference from group to division, delete titles of persons designated to approve modifications, clarify the responsibilities of the Safety Assurance Manager, and delete the requirement to perform an annual independent Fire Protection Audit will not involve a significant increase in the probability or consequences of an accident previously evaluated because the changes do not have any impact upon the design or operation of any plant systems or components. [2. The amendments do not create the possibility of a new or different kind of accident from any accident previously evaluated.] The proposed revisions will not create the possibility of a new or different kind of accident from any previously evaluated because the changes are administrative in nature and operation of Catawba, McGuire, and Oconee Nuclear Stations in accordance with these TS [technical specifications] will not create any failure modes not bounded by previously evaluated accidents. [3. The amendments do not involve a significant reduction in a margin of safety.] The proposed revisions will not involve a reduction in a margin of safety because they are administrative in nature. 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. Local Public Document Room location: Atkins Library, University of North Carolina, Charlotte (UNCC Station), North Carolina 28223 Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 South Church Street, Charlotte, North Carolina 28242 NRC Project Director: Loren R. Plisco, Acting Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana Date of amendment request: December 6, 1993 Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to add provisions to allow repair of steam generator tubes by the sleeving process. 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: The proposed change will allow the use of specific steam generator tubing sleeves to repair Waterford 3 steam generator tubes which exhibit degradation and can be sleeved (in the tube sheet crevice area and egg crate supports). The technical specification change is proposed to reference the following reports [following NRC approval], which qualify the use of steam generator tube sleeves as an alternative to tube plugging, Combustion Engineering Report CEN-605-P, ``Waterford 3 Steam Generator Tube Repair Using Leak Tight Sleeves'', Revision 00-P, dated December, 1992. Westinghouse Report WCAP-13698, ``Laser Welded Sleeves For 3/4 Inch Diameter Tube Feedring-Type and Westinghouse Preheater Steam Generators'', Revision 1, dated May, 1993. Babcock & Wilcox Report 51-1223750-00, ``BWNS Kinetic Sleeve Design For CE SGs with 0.048'' Wall Tubes'', Revision 00, dated June 29, 1993. These reports demonstrate that the repair of degraded steam generator tubes using tube sleeves will result in tube bundle integrity consistent with the original design basis. Sleeve design, materials, and joints were designed to the applicable ASME Boiler and Pressure Vessel Codes. Extensive analyses and testing were performed to demonstrate the adequacy of the tube sleeves. The analyses were performed using design and operating transient parameters which enveloped loads imposed during normal operating, upset and accident conditions. Mechanical testing was performed to demonstrate leak resistance and joint strength, including fatigue resistance. Corrosion testing was also performed to assess the corrosion resistance of the sleeve and joint. Based upon the results of the analytical and test programs described in detail in the above mentioned reports, these tube sleeves meet or exceed all the established design and operating criteria. Utilization of tube sleeves not only reduces the risk of primary to secondary leakage in the steam generator, but can also provide for more margin in the safety analysis. Sleeving a tube results in a primary flow reduction which has no significant effect on the steam generator performance with respect to heat transfer or system flow resistance and pressure drop. The cumulative impact of multiple sleeved tubes has been evaluated to ensure the effects remain within the design bases. The installation of tube sleeves can be accomplished within the tube plugging analysis. Based on the extensive analysis and test program performed and the ability to monitor and remove degraded sleeves from service, this change does not significantly increase the probability or consequences of an accident previously evaluated. A sleeved tube performs the same function in the same passive manner as the unsleeved tube. Tube sleeves are designed, qualified, and maintained under the stress and pressure limits of ASME Section III and Regulatory Guide 1.121. Eddy current testing is performed following installation of each sleeve in order to verify the proper installation of the tube sleeve and to obtain baseline eddy current data. This baseline data is used to monitor any subsequent degradation. Therefore, the use of tube sleeves does not create the possibility of a new or different kind of accident from any accident previously evaluated. Steam generator tube integrity is maintained under the same limits for sleeved tubes as for unsleeved tubes in accordance with ASME Section III and Regulatory Guide 1.1.21. The degradation limit at which a tube is considered inoperable remains unchanged and is detectable for sleeves as well as tubes. The technical specifications continue to require monitoring and restriction of primary to secondary system leakage through the steam generators, such that there remains reasonable assurance that a significant increase in leakage, due to failure of a sleeved (or unsleeved) tube, will be detected. The slight reduction in reactor coolant system flow, due to sleeving, is considered to have an insignificant impact on steam generator operation during normal operation and accident conditions and is clearly bounded by tube plugging evaluations. The technical specifications will continue to contain reporting requirements for tubes which have had their degradation spanned regardless of whether the tube is plugged or sleeved. Therefore, the proposed change will 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. Local Public Document Room location: University of New Orleans Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122 Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L Street N.W., Washington, D.C. 20005-3502 NRC Project Director: William D. Beckner Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana Date of amendment request: December 14, 1993 Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to revise the Azimuthal Power Tilt limit from less than or equal to 0.10 (10%) to less than or equal to 0.03 (3%) and to revise the action statement for control element assembly (CEA) misalignment to allow 24 hours to restore the tilt to less than 3%. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: The additional time for recovery from a CEA misalignment is acceptable for the following reasons: Consistent with the safety analyses, this TS, places a limit on tilt for steady state operation as an initial condition for the safety analyses. It is not a limit to be applied during transients. This is because accident analyses are initiated from steady state conditions and are not required to assume a core power distribution transient simultaneous with or immediately prior to the accident. This would in effect be two accidents occurring simultaneously. The probability of having an accident immediately after a CEA drop during the 24 hour period allowed for tilt to be restored to less than 3% is very low. Technical Specification 3/4.1.3 addresses the CEA misalignments and requires a 30% power reduction within one hour if the CEA cannot be restored to its proper position. Reducing power as required in TS 3.2.3 action b.2 will tend to increase the azimuthal tilt, making the transient worse. A lower power reduces the rate at which xenon near the dropped CEA can be burned out. Maintaining power will quicken the process and keep the tilt as low as possible. The additional time is only allowed for a confirmed CEA misalignment which operators have procedures to respond to. The change in tilt is expected and is not indicative of anomalous core power distribution behavior that might require more immediate action. This change conservatively reduces the Azimuthal Power Tilt technical specification limit to agree with the assumptions used in the safety analysis. The lower tilt represents a more even power distribution in the core. A CEA drop event may temporarily cause the azimuthal power tilt to exceed the 3% limit. However, for the reasons identified above and since the probability of having another event within the 24 hours allowed for recovery after the CEA drop is extremely low, this change does not involve a significant increase in the probability or consequences of any accident. The change in technical specification limit on tilt does not involve any change to any equipment or the manner in which the plant will be operated. This change will further restrict unevenness in the core power distribution. Therefore, this change does not create the possibility of a new or different kind of accident previously evaluated. The proposed change incorporates an Azimuthal Power Tilt technical specification limit to agree with the assumptions used in the safety analysis. Implementation of this change will preserve the margin of safety and be consistent with the safety analyses. Therefore, this change does not involve a significant reduction in 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. Local Public Document Room location: University of New Orleans Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122 Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L Street N.W., Washington, D.C. 20005-3502 NRC Project Director: William D. Beckner Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana Date of amendment request: December 14, 1993 Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) by removing the reactor vessel material specimen withdrawal schedule and by updating the reactor coolant system pressure-temperature (P-T) curves. The specimen withdrawal schedule will be relocated to the Updated Final Safety Analysis Report (UFSAR). 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: Although the Reactor Vessel material specimens withdrawal schedule will be removed from the Technical Specifications, the Technical Specifications bases will continue to provide background information on the use of the data obtained from material specimens. Also, updates to the schedule will continue to be submitted to the NRC for approval prior to implementation. Operating the plant in accordance with the new, updated P-T Curves will assure preserving the structural integrity of the reactor vessel over the life of the plant. The pressure and temperature limits were developed in accordance with 10 CFR [Part] 50 Appendix G requirements. Removing the requirements associated with the previous exemption to Appendix H (TS 4.4.8.1.2 items a & b) is purely an administrative change. Therefore, the proposed changes will not significantly increase the probability or consequences of any accident previously evaluated. Removal of the Reactor Vessel material specimen schedule from the Technical Specifications has no impact on accidents at the plant. Updates to the schedule will still be required to be submitted to the NRC prior to implementation per Section II.B.3 of Appendix H to 10 CFR Part 50. Also, updates to the P-T Curves will not create a new or different type [of] accident. The reactor vessel beltline P-T limits were revised applying the general guidance of the ASME Code, Appendix G procedures with the necessary margins of safety for heatup, cooldown and inservice hydro test conditions. The change to TS 4.4.8.1.2 items a & b is a purely administrative. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated. Removal of the schedule for Reactor Vessel material specimen withdrawal from the Technical Specifications does not impact the margin of safety. The schedule will continue to receive NRC review and approval prior to implementation of updates to the schedule. Updates to the P-T Curves are provided to preserve the margin to [sic] safety to assure that when stressed under operating, maintenance and testing the boundary behaves in a non-brittle manner and the probability of rapidly propagating fracture is minimized. The change to TS 4.4.8.1.2 items a & b is a purely administrative. Therefore, the proposed changes will not result in a significant reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: University of New Orleans Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122 Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L Street N.W., Washington, D.C. 20005-3502 NRC Project Director: William D. Beckner GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania Date of amendment request: November 30, 1993 Description of amendment request: The purpose of the request is to change the plant Technical Specifications (TS) by removing the protective and maximum allowable setpoint limits for axial power imbalance and the trip setpoint for nuclear overpower based on reactor coolant system (RCS) flow (flux-to-flow) from the TS and relocating them to the existing TMI-1 Core Operating Limits Report (COLR). The proposed change is in accordance with Generic Letter 88-16 guidance with regard to placing cycle-dependent parameters into the COLR and the NRC-approved Babcock and Wilcox Fuel Company (BWFC) Topical Report BAW- 10179P-A, ``Safety Criteria and Methodology for Acceptable Cycle Reload Analyses.'' The TMI-1 Cycle 10 COLR, submitted to the NRC on November 7, 1993, includes these protective and maximum allowable setpoint limits and nuclear overpower trip setpoints to support this Technical Specifications change. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated. The proposed amendment relocates protective and maximum allowable setpoint limits from Technical Specifications, and design nuclear power peaking factors and the maximum allowable local linear heat rate limit from Technical Specification Bases, to the TMI-1 Core Operating Limits Report in accordance with NRC-approved Topical Report BAW-10179P-A. The proposed amendment provides continued control of the values of these limits and assures these values remain consistent with all applicable limits of the safety analyses addressed in the TMI-1 FSAR [Final Safety Analysis Report]. The Technical Specifications retain the requirement to maintain the plant within the appropriate bounds of these limits. Therefore, the proposed amendment has no effect on the probability of occurrence or consequences of an accident previously evaluated. 2. Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment relocates protective and maximum allowable setpoint limits, design nuclear power peaking factors and maximum allowable local linear heat rate limit to the TMI-1 Core Operating Limits Report. The Technical Specifications retain the requirement to maintain the plant within the appropriate bounds of these limits. Therefore, the proposed amendment has no effect on the possibility of creating a new or different kind of accident from any accident previously evaluated. 3. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety. The proposed amendment provides continued control of the values of these limits and assures these values remain consistent with all applicable limits of the safety analyses addressed in the TMI-1 FSAR. Therefore, it is concluded that operation of the facility in accordance with the proposed amendment 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. Local Public Document Room location: Government Publications Section, State Library of Pennsylvania, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, Pennsylvania 17105. Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037. NRC Project Director: John F. Stolz Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York Date of amendment request: December 14, 1993 Description of amendment request: The amendment would revise Technical Specification (TS) 3/4.8.2, ``DC Sources,'' to delete two notes that indicate that two 125-volt full capacity battery chargers are required when the Uninterruptible Power Supply is powered by its backup DC power supply. These notes apply to the Divisions I and II DC sources during operating and shutdown conditions. The licensee has determined that only one battery charger is required to meet current design requirements and the criteria delineated in Regulatory Guide 1.32, ``Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants.'' The amendment would also revise TS 3/4.8.2 to increase the minimum allowable electrolyte temperature for the 125-volt batteries from 60 deg.F to 65 deg.F. This proposed change would establish consistency between the TSs, the Updated Safety Analysis Report, and applicable battery capacity calculations. The amendment would also make administrative changes to TS 3/4.8.4, ``Electrical Equipment Protective Devices,'' and the TS Bases. 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: The operation of Nine Mile Point Unit 2, in accordance with the proposed amendment, will not involve a significant increase in the probability or consequences of an accident previously evaluated. The existing ``notes'' at the bottom of pages 3/4 8-14 and 3/4 8-19 indicate that two (2) 125-volt full capacity chargers are required when the Uninterruptible Power Supply is powered by its backup DC power supply. These ``notes'' were based on an overly conservative calculation which determined that both chargers were required to supply adequate power to connected loads including an Uninterruptible Power Supply. More recent calculations indicate that one (1) charger is adequate to supply power to connected loads and an Uninterruptible Power Supply. Based on these more recent calculations, Niagara Mohawk proposes to delete these ``notes.'' Because it has been determined adequate power will be provided to connected loads with one (1) charger, deletion of these ``notes'' will not affect the reliability of connected loads nor their ability to perform their intended function. Therefore, this change will not involve a significant increase in the probability or consequences of an accident previously evaluated. Existing Technical Specification 4.8.2.1 requires that the 125- volt batteries be demonstrated to be operable by verifying that the average electrolyte temperature of one out of five connected cells is above 60 deg.F. The Nine Mile Point Unit 2 Updated Safety Analysis Report and current battery capacity calculations assume a battery electrolyte temperature of at least 65 deg.F. The change from 60 deg.F to 65 deg.F is conservative in that battery capacity is increased. Increasing the capacity of the batteries will not adversely affect the reliability of connected loads nor their ability to perform their intended function. Therefore, this change will not involve a significant increase in the probability or consequences of an accident previously evaluated. The changes made to Technical Specification 3/4.8.4 and to the Bases of Specifications 3/4.6.3 and 3/4.8.4 are administrative changes and do not affect plant systems or operation. Accordingly these changes will not involve a significant increase in the probability or consequences of an accident previously evaluated. The operation of Nine Mile Point Unit 2, in accordance with the proposed amendment, will not create the possibility of a new or different kind of accident from any accident previously evaluated. The function of the battery chargers is to provide adequate power to connected loads. Since it has been determined that one (1) battery charger is sufficient to provide adequate power, deletion of these ``notes'' which require that two (2) chargers be available does not affect the capability of the chargers to perform their function. The proposal to change the required electrolyte temperature from 60 deg.F to 65 deg.F increases the capacity of the batteries and therefore improves the capability of the batteries to perform their intended function. The remaining changes are administrative changes and do not affect plant systems or operation. The proposed changes do not introduce any new accident precursors and do not involve any physical alterations to plant configurations which could initiate a new or different kind of accident. The changes do not adversely affect the design or performance characteristics of the batteries, battery chargers or connected loads. The proposed change to increase the required electrolyte temperature will increase battery capacity. Therefore, the proposed amendment will not create the possibility of a new or different kind of accident from any previously evaluated. The operation of Nine Mile Point Unit 2, in accordance with the proposed amendment, will not involve a significant reduction in a margin of safety. Niagara Mohawk proposes to delete the existing ``notes'' that indicate that two (2) 125-volt full capacity chargers are required when the Uninterruptible Power Supply is powered by its backup DC power supply. Niagara Mohawk engineering has determined that one (1) battery charger is adequate to meet the maximum DC load demands including the Uninterruptible Power Supply. The proposed change to increase the required electrolyte temperature will increase battery capacity. The remaining changes are administrative. These changes will not adversely affect the design or performance characteristics of the batteries, battery chargers, or connected loads nor will they affect the capability of the batteries, battery chargers, or connected loads to perform their intended function. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Reference and Documents Department, Penfield Library, State University of New York, Oswego, New York 13126. Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502. NRC Project Director: Robert A. Capra Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power Plant, Unit 3, Humboldt County, California Date of application for amendment: July 7, 1993 (Reference LAR 93- 01) Brief description of amendment: The proposed amendment would revise the Technical Specifications (TS) for the Humboldt Bay Power Plant Unit No. 3. This proposed revision would change TS VII.H.3, ``Semiannual Radioactive Effluent Release Report,'' to extend the reporting period from semiannually to annually and to change the report submission date from 60 days after January 1 and July 1 of each year to before April 1 of each year. 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: a. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? The proposed TS changes are administrative in nature. The proposed changes to TS VII.H.3 are consistent with 10 CFR 50.36a report requirements. The proposed changes do not affect accident evaluations. The proposed changes are administrative in nature, should result in improved administrative practices, and do not affect plant operations. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. b. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed changes are administrative in nature, do not result in physical alterations or changes to the operation of the plant, and cause no change in the method by which any safety-related system performs its function. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. c. Does the change involve a significant reduction in a margin of safety? These administrative changes do not alter the basic regulatory requirements and do not affect any safety analyses. The proposed change to TS VII.H.3. does not alter any administrative controls over radioactive effluent, nor does the proposed change involve any physical alterations to the plant with respect to radioactive effluents. Therefore, the proposed change would not affect the meaning, application, and function of the TS requirements. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment requests involve no significant hazards consideration. Local Public Document Room location: Humboldt County Library, 636 F Street, Eureka, California 95501 Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas & Electric Company, P.O. Box 7442, San Francisco, California 94120 NRC Branch Chief: John H. Austin Portland General Electric Company, et al., Docket No. 50-344, Trojan Nuclear Plant, Columbia County, Oregon Date of amendment request: January 27, 1993 Description of amendment request: The proposed amendment, by Portland General Electric Company, PGE or the licensee, would change the Trojan Nuclear Plant (Trojan) Appendix A Technical Specifications to reflect the permanently defueled status of the facility. The permanent cessation of power generation at Trojan and the May 5, 1993 amendment to the license which granted the licensee a Possession Only License for the facility has rendered many of the existing provisions of the current Appendix A Technical Specifications inappropriate. PGE has developed Permanently Defueled Technical Specifications (PDTS) for Trojan using NUREG-1431, ``Standard Technical Specifications, Westinghouse Plants,'' as a basis for the PDTS scope and format. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided an analysis of the issue of no significant hazards consideration based upon the following: 1. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed amendment shows that the worst case design basis accident for this plant, in its permanently shutdown defueled state, is a postulated spent fuel handling accident at the Trojan facility. The licensee has also identified a second postulated design basis accident scenario, a loss of forced cooling to the spent fuel pool. Other Trojan Final Safety Analysis Report (FSAR) accident scenarios addressed in Chapter 15 are no longer applicable to Trojan in the permanently defueled mode. The proposed amendment does not lessen any of the requirements associated with handling spent fuel and therefore, the probability of a fuel handling accident occurring is unchanged. The licensee has analyzed the a loss of forced spent fuel pool cooling accident and has shown that this scenario would not result in a radiological release. The proposed amendment does not change the consequences of the accident since it does not affect the magnitude, detection, or mitigation of either accident scenario. Additionally, the ability of the spent fuel pool to withstand other applicable FSAR events, natural phenomena, and fires is either unchanged from the existing licensing basis or is improved during the permanently defueled condition. 2. Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. Maintaining the permanently defueled facility in accordance with the PDTS does not create the possibility of a new or different kind of accident from any previously considered. Most of the existing plant systems and functions will not be operational in the permanently defueled condition since power operations are prohibited and all of the fuel at Trojan is stored in the spent fuel pool. However, all structures, systems and components that are necessary for safe fuel handling and storage activities will be maintained operable during the permanently defueled condition. The proposed PDTS provide operation and surveillance requirements and administrative controls which are sufficient to ensure that the required structures, systems and components will be maintained operable in the permanently defueled condition. 3. Operation of the facility in accordance with the proposed amendment does not involve a significant reduction in a margin of safety. The proposed PDTS are sufficient to ensure no reduction in a margin of safety, in part, because of the reduced range of design basis accidents against which the facility must be protected now that the facility is prohibited from power operations and is permanently defueled. Only a fuel handling accident or a loss of forced cooling to the spent fuel pool are relevant during the permanently defueled condition. The margins of safety for both of these accidents will remain the same or improve by maintaining the facility in accordance with the proposed PDTS. None of the other Chapter 15 FSAR accidents are applicable since power operations are prohibited and the facility is permanently defueled. Additionally, the margins of safety for other applicable FSAR events, natural phenomena, and fires are either unchanged from the existing licensing basis or are improved during the permanently defueled condition. 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. Local Public Document Room location: Branford Price Millar Library, Portland State University, 934 S.W. Harrison Street, P.O. Box 1151, Portland, Oregon 97207. Attorney for licensees: Leonard A. Girard, Esq., Portland General Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204. NRR Project Director: Seymour H. Weiss Power Authority of The State of New York, Docket No. 50-286, Indian Point Nuclear Generating Unit No. 3, Westchester County, New York Date of amendment request: December 20, 1993 Description of amendment request: The licensee has requested an amendment to the Technical Specifications (TS) to revise Section 3.3.D (Weld Channel and Penetration Pressurization System) to allow certain portions of the Weld Channel Pressurization System (WCPS) to be disconnected if they are determined to be inoperable and not practicably accessible for repair. The WCPS continuously pressurizes channels over welds in the steel liner of the containment building. To be disconnected, an inoperable portion of the WCPS must be covered by concrete such that repairs would involve removal of part of the containment structure or the inoperable portion of the system is located behind plant equipment inside the containment building such that repairs would involve relocation of the equipment. The licensee has requested this TS amendment since one portion of the WCPS has become inoperable and, since it is buried in concrete below the containment floor, cannot be practically repaired. In addition, administrative changes would be made to TS Section 3.3.B (Containment Cooling and Heat Removal) and TS Section 3.3.E (Component Cooling System) to correct typographical errors. Specifically, TS Section 3.3.B.3.b references 3.3.A.3 when is should reference 3.3.B.1 and TS Section 3.3.E.3.b references 3.3.A.3 when it should reference 3.3.E.1. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: Consistent with the criteria of 10 CFR 50.92, the enclosed application is judged to involve no significant hazards based on the following information: (1) Does the proposed license amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: The proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. The consequences of an accident previously evaluated would not be affected by the disconnection of portions of the WCPS [Weld Channel Pressurization System] because the accident analyses do not assume the operation of any portion of the WC & PPS [Weld Channel & Penetration Pressurization System]. Additionally, operation of the WC & PPS is not taken credit for in any offsite accident dose calculations. The probability of an accident previously evaluated would not be increased because the disconnection of any portion of the WCPS could not initiate an accident. The administrative changes correct errors in the technical specifications and technical specification bases. These administrative changes have no affect on the probability or consequences of an accident previously evaluated. (2) Does the proposed license amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. The allowance for the disconnection of sections of the WCPS will allow the Authority [Power Authority of the State of New York] to avoid repairs that can potentially degrade containment integrity or the condition of vital equipment. The proposed license amendment does not create the possibility of a new accident because the disconnection of any portion of the WCPS could not initiate an accident. The administrative changes correct errors in the technical specifications and technical specification bases. These administrative changes can not [cannot] 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? Response: The proposed amendment would not involve a significant reduction in a margin of safety. The allowance for the disconnection of sections of the WCPS will allow the Authority to avoid repairs that can potentially degrade containment integrity or the condition of vital equipment. The WC & PPS will still provide continuous pressurization and monitoring of leak-tightness for the zones incorporated into the containment penetrations and for at lease 80% of the channels over the welds in the steel inner of the containment building. The WC & PPS will continue to provide assurance that the containment leak rate in the event of an accident is lower than that assumed in the accident analyses because the accident analyses do not assume that any section of the WC & PPS is operating. The administrative changes correct errors in the technical specifications and technical specification bases. These administrative changes have no affect on any margin of safety. 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. Local Public Document Room location: White Plains Public Library, 100 Martine Avenue, White Plains, New York 10601. Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle, New York, New York 10019. NRC Project Director: Robert A. Capra Public Service Electric & Gas Company, Docket Nos. 50-272 and 50- 311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey Date of amendment request: December 8, 1993. The December 8, 1993 request supersedes an earlier request dated November 17, 1992, which was previously noticed (58 FR 52994). This notice supersedes the previous notice. Description of amendment request: The proposed amendments would revise the Technical Specifications for Salem Units 1 and 2 to incorporate the guidance provided by the staff in Generic Letter 90-06 (GL 90-06) as follows: 1. Specification 3/4.4.3 and 3/4.4.5 for Salem 1 and Salem 2, respectively, ``RELIEF VALVES'', will incorporate the guidance of GL 90-06 with the following exceptions: a. The surveillance requirement to test the emergency power supply for the power operated relief valves (PORVs) and block valves has not been incorporated. The PORVs and block valves are powered from the emergency busses. b. The entry conditions for one or both PORVs inoperable will not be based on excessive seat leakage alone. Entry conditions will be based on the capability of the PORV to be manually cycled consistent with the Action Statements contained in NUREG-1431, Standard Technical Specifications for Westinghouse Plants. c. With both PORVs inoperable in Modes 1, 2, or 3 and not capable of being manually cycled, or both block valves are inoperable, an allowed outage time of 6 hours to restore on block valve or PORV to operable status has been requested. 2. Specification 3.4.9.3 and 3.4.10.3 for Salem 1 and Salem 2, respectively, ``OVERPRESSURE PROTECTION SYSTEMS'' will incorporate the guidance of GL 90-06. 3. In addition to the guidance provided by GL 90-06, the following changes have also been proposed. a. For Salem 1, the reference to the specific American Society of Mechanical Engineers (ASME) valve category would be deleted from Specification 3.4.9.3.1. b. Specification 3.5.3. ``ECCS SUBSYSTEMS - TAVE 131 and its transient dose equivalent I131 reactor coolant specific activity will be reduced by a factor of 4 in order to increase the allowable leakage in the event of a steam 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, which is presented below: 1) Operation of Farley Unit 1 in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. Testing of model boiler specimens for free standing tubes at room temperature conditions show burst pressures as high as approximately 5000 psi [per square inch] for indications of outer diameter stress corrosion cracking with voltage measurements as high as 26.5 volts. Burst testing performed on pulled tubes with up to 7.5 volt indications show burst pressures in excess of 5900 psi at room temperature. As stated earlier, tube burst criteria are inherently satisfied during normal operating conditions by the presence of the tube support plate. Furthermore, correcting for the effects of temperature on material properties and minimum strength levels (as the burst testing was done at room temperature), tube burst capability significantly exceeds the R.G. [Regulatory Guide] 1.121 criterion requiring the maintenance of a margin of 1.43 times the steam line break pressure differential on tube burst if through- wall cracks are present without regard to the presence of the tube support plate. Based on the existing data base this criterion is satisfied with bobbin coil indications with signal amplitudes over twice the 2.0 volt interim repair criteria, regardless of the indicated depth measurement. This structural limit is based on a lower 95 [percent] confidence level limit of the data. The 2.0 volt criteria provides an extremely conservative margin of safety to the structural limit considering expected growth rates of outside diameter stress corrosion cracking at Farley. Alternate crack morphologies can correspond to a voltage so that a unique crack length is not defined by a burst pressure to voltage correlation. However, relative to expected leakage during normal operating conditions, no field leakage has been reported from tubes with indications with a voltage level of under 7.7 volts for a 3/4 inch tube with a 10 volt correlation to 7/8 inch tubing (as compared to the 2.0 volt proposed interim tube repair limit). Thus, the proposed amendment does not involve a significant increase in the probability or consequences of an accident. Relative to the expected leakage during accident condition loadings, the accidents that are affected by primary-to-secondary leakage and steam release to the environment are Loss of External Electrical Load and/or Turbine Trip, Loss of All AC Power to Station Auxiliaries, Major Secondary System Pipe Failure, Steam Generator Tube Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a Control Rod Drive Mechanism Housing. Of these, the Major Secondary System Pipe Failure is the most limiting for Farley in considering the potential for off-site doses. The offsite dose analyses for the other events which model primary-to secondary leakage and steam release from the secondary side to the environment assume that the secondary side remains intact. The steam generator tubes are not subjected to a sustained increase in differential pressure, as is the case following a steam line break event. This increase in differential pressure is responsible for the postulated increase in leakage and associated offsite doses following a steam line break event. In addition, the steam line break event results in a bypass of containment for steam generator leakage. Upon implementation of the interim repair criteria, it must be verified that the expected distribution of cracking indications at the tube support plate intersections are such that primary-to-secondary leakage would result in site boundary dose within the current licensing basis. Data indicate that a threshold voltage of 2.8 volts would result in through-wall cracks long enough to leak at steam line break conditions. Application of the proposed repair criteria requires that the current distribution of a number of indications versus voltage be obtained during the refueling outages. The current voltage is then combined with the rate of change in voltage measurement and a voltage measurement uncertainty to establish an end of cycle voltage distribution and, thus, leak rate during steam line break pressure differential. The leak rate during a steam line break is further increased by a factor related to the probability of detection of the flaws. If it is found that the potential steam line break leakage for degraded intersections planned to be left in service coupled with the reduced specific activity levels allowed result in radiological consequences outside the current licensing basis, then additional tubes will be plugged or repaired to reduce steam line break leakage potential to within the acceptance limit. Thus, the consequences of the most limiting design basis accident are constrained to present licensing basis limits. 2) The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. Implementation of the proposed interim tube support plate elevation steam generator tube repair criteria does not introduce any significant changes to the plant design basis. Use of the criteria does not provide a mechanism which could result in an accident outside of the region of the tube support plate elevations. Neither a single or multiple tube rupture event would be expected in a steam generator in which the repair criteria has been applied (during all plant conditions). The bobbin probe signal amplitude repair criteria is established such that operational leakage of excessive leakage during a postulated steam line break condition is not anticipated. Southern Nuclear has previously implemented a maximum leakage rate limit of 140 gpd [gallons per day] per steam generator on Unit 1. The R.G. 1.121 criterion for establishing operational leakage rate limits that require plant shutdown are based upon leak-before-break considerations to detect a free span crack before potential tube rupture. The 140 gpd limit provides for leakage detection and plant shutdown in the event of the occurrence of an unexpected single crack resulting in leakage that is associated with the longest permissible crack length. R.G. 1.121 acceptance criteria for establishing operating leakage limits are based on leak-before-break considerations such that plant shutdown is initiated if the leakage associated with the longest permissible crack is exceeded. The longest permissible crack is the length that provides a factor of safety of 1.43 against bursting at steam line break pressure differential. A voltage amplitude approximately 9 volts for typical outside diameter stress corrosion cracking corresponds to meeting this tube burst requirement at the 95 [percent] prediction interval on the burst correlation. Alternate crack morphologies can correspond to a voltage so that a unique crack length is not defined by the burst pressure versus voltage correlation. Consequently, typical burst pressure versus through- wall crack length correlations are used below to define the ``longest permissible crack'' for evaluating operating leakage limits. The single through-wall crack lengths that result in tube burst at 1.43 times steam line break pressure differential and steam line break conditions are about 0.53 inch and 0.84 inch, respectively. Normal leakage for these crack lengths would range from about 0.4 gallons per minute to 4.5 gallons per minute respectively while lower 95 [percent] confidence level leak rates would range from about 0.06 gallons per minute to 0.6 gallons per minute, respectively. An operating leak rate of 140 gpd per steam generator has been implemented on Unit 1. This leakage limit provides for detection of 0.4 inch long cracks at nominal leak rates and 0.6 inch long cracks at the lower 95 [percent] confidence level and for three times normal operating pressure differential at less than nominal leak rates. Based on the above, the implementation of interim plugging criteria will not create the possibility of a new or different kind of accident from any previously evaluated. 3) The proposed license amendment does not involve a significant reduction in margin of safety. The use of the interim tube support plate elevation repair criteria is demonstrated to maintain steam generator tube integrity commensurate with the requirements of R.G. 1.121. R.G. 1.121 describes a method acceptable to the NRC staff for meeting [General Design Criteria] 2, 14, 15, 31, and 32 by reducing the probability of the consequences of steam generator tube rupture. This is accomplished by determining the limiting conditions of degradation of steam generator tubing, as established by inservice inspection, for which tubes with unacceptable cracking should be removed from service. Upon implementation of the criteria, even under the worst case conditions, the occurrence of outside diameter stress corrosion cracking at the tube support plate elevations is not expected to lead to a steam generator tube rupture event during normal or faulted plant conditions. The most limiting effect would be a possible increase in leakage during a steam line break event. Excessive leakage during a steam line break event, however, is precluded by verifying that, once the criteria are applied, the expected end of cycle distribution of crack indications at the tube support plate elevations would result in minimal, and acceptable primary to secondary leakage during the event and hence help to demonstrate radiological conditions are less than an appropriate fraction of the 10 CFR [Part] 100 guideline. The margin to burst for the tubes using the interim repair criteria is comparable to that currently provided by existing technical specifications. In addressing the combined effects of LOCA [loss-of-coolant accident] + SSE [safe shutdown earthquake] on the steam generator component (as required by GDC 2), it has been determined that tube collapse may occur in the steam generators at some plants. This is the case as the tube support plates may become deformed as a result of lateral loads at the wedge supports at the periphery of the plate due to either the LOCA rarefaction wave and/or SSE loadings. Then the resulting pressure differential on the deformed tubes may cause some of the tubes to collapse. There are two issues associated with steam generator tube collapse. First, the collapse of steam generator tubing reduces the RCS [reactor coolant system] flow area through the tubes. The reduction in flow area increases the resistance to flow of steam from the core during a LOCA which, in turn, may potentially increase Peak Clad Temperature (PCT). Second, there is a potential the partial through-wall cracks in tubes could progress to through-wall cracks during tube deformation or collapse or that short through- wall indications would leak at significantly higher leak rates than included in the leak rate assessments. Consequently, a detailed leak-before-break analysis was performed and it was concluded that the leak-before-break methodology (as permitted by GDC 4) is applicable to the Farley Unit 1 [RCS] primary loops and, thus, the probability of breaks in the primary loop piping is sufficiently low that they need not be considered in the structural design basis of the plant. Excluding breaks in RCS primary loops, the LOCA loads from the large branch line breaks were analyzed at Farley Unit 1 and were found to be of insufficient magnitude to result in steam generator tube collapse or significant deformation. Regardless of whether or not leak-before-break is applied to the primary loop piping at Farley, any flow area reduction is expected to be minimal (much less than 1 [percent]) and PCT margin is available to account for this potential effect. Based on analyses results, no tubes near wedge locations are expected to collapse or deform to the degree that secondary to primary in-leakage would be increased over current expected levels. For all other steam generator tubes, the possibility of secondary-to-primary leakage in the event of a LOCA + SSE event is not significant. In actuality, the amount of secondary-to-primary leakage in the event of a LOCA + SSE is expected to be less than that currently allowed, i.e., 500 gpd per steam generator. Furthermore, secondary-to-primary in- leakage for the same pressure differential since the cracks would tend to tighten under a secondary-to-primary pressure differential. Also the presence of the tube support plate is expected to reduce the amount of in-leakage. Addressing the R.G. 1.83 considerations, implementation of the tube repair criteria is supplemented by 100 [percent] inspection requirements at the tube support plate elevations having outside diameter stress corrosion cracking indications, reduced operating leak rate limits, eddy current inspection guidelines to provide consistency in voltage normalization, and rotating pancake coil inspection requirements for the larger indications left in service to characterize the principal degradation mechanism as outside diameter stress corrosion cracking. As noted previously, implementation of the tube support plate elevation repair criteria will decrease the number of tubes which must be taken out of service with tube plugs or repaired. The installation of steam generator tube plugs or tube sleeves would reduce the RCS flow margin, thus implementation of the interim repair criteria will maintain the margin of flow that would otherwise be reduced through increased tube plugging or sleeving. Based on the above, it is concluded that the proposed change does not result in a significant reduction in margin with respect to plant safety as defined in the Final Safety Analysis Report or any bases of the plant Technical Specifications. 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. Local Public Document Room location: Houston-Love Memorial Library, 212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302 Attorney for licensee: James H. Miller, III, Esq., Balch and Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, Alabama 35201 NRC Project Director: S. Singh Bajwa Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia Date of amendment request: December 10, 1993 Description of amendment request: The proposed changes would modify the surveillance frequency of the auxiliary feedwater system pumps and valves from monthly to quarterly. Various administrative changes are being proposed such as 1) punctuation and grammar, 2) correction of system or component names, and 3) capitalization of defined words. 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: Specifically, operation of Surry Power Station in accordance with the proposed Technical Specifications changes will not: 1. Involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated. Changing the surveillance test frequency of the Auxiliary Feedwater System pumps and valves does not significantly affect the probability of occurrence or consequences of any previously evaluated accidents. The probability of an accident occurrence is not increased in itself by the proposed changes in surveillance testing of the Auxiliary Feedwater System pumps and valves. [A]uxiliary feedwater pump testing is performed through a full-flow test line, thereby not affecting normal plant operations. Changes to the testing therefore do not affect the probability of an accident occurrence. Redundant trains of the Auxiliary Feedwater System remain available during surveillance testing, therefore, the consequences of an accident are unchanged by the proposed changes in the Auxiliary Feedwater System surveillance test frequencies. Quarterly testing of the pumps and valves will continue to assure that the Auxiliary Feedwater System is capable of performing its intended functions for either unit if called upon. Consistent with Generic Letter 93-05, ``Line-Item Technical Specifications Improvements to Reduce Surveillance Requirements for Testing During Power Operation,'' the new testing frequency should reduce Auxiliary Feedwater System unavailability resulting from failures and equipment degradation during testing, thereby resulting in improved system reliability. Invoking ASME Section XI as the acceptance criteria for testing the Auxiliary Feedwater Pumps is an enhancement to the acceptance criteria presently specified. Furthermore, the operability requirements for the Auxiliary Feedwater System remain unchanged. Therefore, the probability or consequences of any previously analyzed accident are not increased by the proposed changes in surveillance requirements for the Auxiliary Feedwater System. 2. Create the possibility of a new or different kind of accident from any accident previously evaluated. Changes in test frequency and acceptance criteria for the Auxiliary Feedwater System pumps and valves do not involve any physical modification of the plant or result in a change in a method of operation. Quarterly testing of the pumps and valves during both operation and shutdown will continue to assure that the Auxiliary Feedwater System will be capable of performing its intended function for either unit. Invoking ASME Section XI acceptance criteria for testing the Auxiliary Feedwater System pumps is an enhancement to the acceptance criteria presently specified. The operability requirements for the Auxiliary Feedwater System remain unchanged. Furthermore, new or different failure modes are not introduced by these changes in surveillance requirements. Therefore, a new or different type of accident is not created by these proposed changes in surveillance requirements for the Auxiliary Feedwater System. 3. Involve a significant reduction in a margin of safety. Changing the surveillance requirements of the Auxiliary Feedwater System pumps and valves does not affect any safety limits or limiting safety system settings. System operating parameters are unaffected. This reduced pump and valve testing frequency should reduce Auxiliary Feedwater System unavailability due to actual testing, as well as failures and equipment degradation during testing. Thus reduced testing results in an improved system reliability. Quarterly testing of the pumps and valves during operation and shutdown will continue to assure that the Auxiliary Feedwater System will be capable of performing its intended functions for either unit. Therefore, the reduction in surveillance testing requirements for the Auxiliary Feedwater System pumps and valves does not reduce any margin of safety. 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. Local Public Document Room location: Swem Library, College of William and Mary, Williamsburg, Virginia 23185. Attorney for licensee: Michael W. Maupin, Esq., Hunton and Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 23219. NRC Project Director: Herbert N. Berkow Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee Nuclear Power Plant, Kewaunee County, Wisconsin Date of amendment request: September 17, 1993 Description of amendment request: The proposed amendment would revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications (TS) by incorporating technical and administrative changes to TS 4.5, Emergency Core Cooling System and Containment Air Cooling System Tests; TS 4.7, Main Steam Isolation Valves; and Table TS 4.1-3, Minimum Frequencies for Equipment Tests. Changes are proposed for the safety injection (SI) system automatic initiation test; the internal containment spray system (ICS) flow blockage test; the SI, ICS and residual heat removal pumps' periodic tests; the main steam isolation valves' test; and the periodic control rod functional test. 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: (a) TS 4.5.a.1.A The proposed change was reviewed in accordance with the provisions of 10 CFR 50.92 to show no significant hazards exist. The proposed change will not: 1) involve a significant increase in the probability or consequences of an accident previously evaluated. The probability of an accident previously evaluated is not increased by the TS change. The changes do not affect any structure, system, or component that initiates an accident analyzed in the Updated Safety Analysis Report (USAR). The probability of an accident occurring is independent of the availability of emergency core cooling components used to mitigate an accident. The consequences of an accident previously evaluated will not be increased by this TS change. Revising the TS wording to clarify that the pumps may be operated during the periodic surveillance tests does not decrease their availability and therefore does not decrease their ability to mitigate the consequences of accidents previously evaluated in the USAR. Clarifying the TS wording will not increase the probability or consequences of an accident previously evaluated. 2) create the possibility of a new or different kind of accident from any accident previously evaluated. A new or different kind of accident from those previously evaluated in the USAR will not be created by this TS change. The automatic SI actuation is designed to respond to various events analyzed in the USAR which take credit for SI in the event mitigation. The test required by TS 4.5.a.1.A verifies that the valves, pump circuit breakers, and automatic circuitry receive the SI signal in the proper sequence. The procedural prerequisites for the performance of this test ensure that an adequate flow path and overpressure protection are available for the pumps during the test. This proposed amendment does not alter the plant configuration, operating setpoints, or overall plant performance. It simply provides clarification that the pumps may start and operate in conjunction with the automatic circuitry test; howev [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%3AX94-10119. Public record. Not legal advice.