Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or proposed to be issued from, October 15, 2004, through October 28, 2004. The last biweekly notice was published on October 26, 2004 (69 FR 62467).
The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.
The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period should circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. Should the Commission take action prior to the expiration of either the comment period or the notice period, it will publish in the
Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555–0001, and should cite the publication date and page number of this
Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide
As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also set forth the specific contentions which the petitioner/requestor seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner/requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner/requestor intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner/requestor to relief. A petitioner/requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing.
If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must be filed by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555–0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff; (3) e-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission,
Nontimely requests and/or petitions and contentions will not be entertained absent a determination by the Commission or the presiding officer of the Atomic Safety and Licensing Board that the petition, request and/or the contentions should be granted based on a balancing of the factors specified in 10 CFR 2.309(a)(1)(I)–(viii).
For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site,
1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change to Technical Specification Surveillance Requirement (SR) 3.7.6.1 will allow a 5% stroke rather than a complete (100%) stroke of each turbine bypass valve (TBV), and will extend the surveillance frequency from 92 days to 120 days. The requirement to verify one complete cycle of each TBV once after each entry into MODE 4 will be retained.
The proposed testing requirements will provide a level of assurance, equivalent to that which now exists, that the TBVs will remain operable throughout the operating cycle, and that they will be able to perform their intended safety function if called upon to do so. Additionally, the reduction in the potential for plant transients that can result from the current testing requirements, will more than offset the small increase (less than one half of one percent) in TBV failure probability per cycle with the proposed testing regime. Thus the proposed changes will not significantly increase the probability of an accident previously evaluated.
Fermi 2 is analyzed for the increase in reactor pressure transient events with the assumption that the Main Turbine Bypass System (MTBS) is out-of-service. Feedwater Controller Failure Upscale represents the most limiting event in this analytical category, and provides the basis for the Minimum Critical Power Ratio (MCPR) operating limits that are applicable when the MTBS is out of service. Because the proposed testing requirements do not alter the assumptions for any of the increase in pressure transient events, the radiological consequences of an accident previously evaluated are not increased.
Therefore, this proposed amendment will not involve a significant increase in the probability or the consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change does not significantly affect the assumed performance of the TBVs, nor does it affect any other plant systems, structures, or components. In fact, these changes reduce the possibility of secondary plant transients and the potential for recirculation pump runbacks during the performance of this SR while at power. The proposed changes do not install any new plant equipment, nor is installed plant equipment being operated in a new or different manner. The proposed changes in test frequency and methodology will continue to ensure that the TBVs remain capable of performing their intended safety function. Therefore, this proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction in the margin of safety.
The proposed change will modify the scope and the frequency of the quarterly full stroke test of the TBVs. The operability requirements and functional characteristics of the TBVs remain unchanged. The proposed change to SR 3.7.6.1 from full stroke testing to 5% stroke testing, and from 92 days to 120 days has been evaluated to produce only a minimal increase in the failure probability of a TBV during each cycle (less than one half of one percent). This failure probability increase is outweighed by the reduction in the potential for plant transients resulting from full stroke testing during power operation. Both Alstom's sensitivity study, and actual industry experience at Ringhals Units 1 and 2 have shown that a partial stroke test will ensure that the valves remain mechanically operable throughout the operating cycle. The Alstom study further shows that a partial stroke test at 120 days, rather than at 92 days, will ensure that the valves remain mechanically operable throughout the operating cycle. Additionally, retaining the requirement to full stroke test each TBV once after each entry into MODE 4 will continue to verify that the valves are mechanically operable prior to their first use following each startup from MODE 4. The TBV response times are used in determining the effect on the MCPR. The surveillance test that ensures the MTBS meets the system's response time limits (SR 3.7.6.3) is not affected by these proposed changes and will continue to be performed at its current 18 month frequency. Therefore, this 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.
1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The basis of the Safety Limit Minimum Critical Power Ratio (SLMCPR) is to ensure no mechanistic fuel damage is calculated to occur if the limit is not violated. The new CPR value preserves the existing margin to transition boiling and probability of fuel damage is not increased. The derivation of the revised SLMCPR for Fermi 2 for incorporation into the Technical Specifications, and its use to determine plant and cycle-specific thermal limits, have been performed using NRC approved methods. These plant-specific calculations are performed each operating cycle and if necessary, will require future changes to these values based upon revised core designs. The revised SLMCPR values do not change the method of operating the plant and have no effect on the probability of an accident initiating event or transient.
Therefore, this proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change results only from a specific analysis for the Fermi 2 Cycle 10 and 11 cores. This change does not involve any new or different methods for operating the facility. No new initiating events or transients result from these changes. Therefore, this proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction in the margin of safety.
The new SLMCPR is calculated using NRC approved methods with plant and cycle-specific parameters for the Cycle 10 and 11 core designs. The SLMCPR value is established to ensure that greater than 99.9% of all fuel rods in the core will avoid transition boiling if the limit is not violated, thereby preserving the fuel cladding integrity. The operating MCPR limit is set appropriately above the safety limit value to ensure adequate margin when the cycle-specific transients are evaluated. Accordingly, the margin of safety is maintained with the revised values. Therefore, this proposed amendment does 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.
No. Approval and implementation of this LAR will have no effect on accident probabilities or consequences since the proposed changes are consistent with those previously reviewed and approved by the NRC in TS Amendment 182/164.
No. This LAR does not involve any physical changes to the plant. Therefore, no new accident causal mechanisms will be generated. The proposed changes are consistent with those previously reviewed and approved by the NRC in TS Amendment 182/164. Consequently, plant accident analyses will not be affected by these changes.
No. Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following accident conditions. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The performance of these barriers will not be affected by the proposed changes since they are consistent with those previously reviewed and approved by the NRC in TS Amendment 182/164. Therefore, the proposed changes in this license amendment will not result in a significant reduction in the facility's 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.
1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
This change clarifies, in various sections of the FSAR, that RCIC system operation is not required in order to mitigate the consequences of the CRDA. The proposed change involves no changes to plant systems or accident analyses. The accident analysis for the CRDA demonstrates that core design, the control rod pattern controls, and the scram signal from the reactor protection system (RPS) effectively prevent damage to the fuel rods as a result of the dropped rod. Furthermore, based on a prescribed source term provided from an assumed damage to less than 2% fuel in the core, the resulting radiological consequences are not affected by RCIC operation or failure to operate. As such, the change does not affect initiation of analyzed events or assumed mitigation of accidents or transients. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
This change clarifies, in various sections of the FSAR, that the RCIC system operation is not required in order to mitigate the consequences of the CRDA. The proposed change does not involve a physical alteration of the plant, add any new equipment, or require any existing equipment to be operated in a manner different from the present design. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction in a margin of safety.
This change clarifies, in various sections of the FSAR, that the RCIC system operation is not required in order to mitigate the consequences of the CRDA. The change has no effect on plant systems, operating practices or safety analyses assumptions. For these reasons, the proposed change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change requires a Steam Generator Program that includes performance criteria that will provide reasonable assurance that the steam generator (SG) tubing will retain integrity over the full range of design basis operating conditions (including startup, power operation, hot standby, cooldown, anticipated transients and postulated accidents). The SG performance criteria are based on tube structural integrity, accident induced leakage, and operational LEAKAGE. These criteria assure that the probability of an accident will not be increased.
The primary to secondary accident induced leakage rate for any design basis accidents, other than an SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. [The primary to secondary accident induced leakage rate is relatively inconsequential for the SG tube rupture analysis.] The operational LEAKAGE performance criterion meets current NRC regulations and NEI [Nuclear Energy Institute] 97–06 criteria for reactor coolant system (RCS) operational primary to secondary LEAKAGE through any one SG of 150 gallons per day. These criteria assure that accident doses will stay within regulatory and licensing basis limits.
Therefore, the proposed change does not affect the probability or consequences of any ANO–1 [Arkansas Nuclear One, Unit 1] analyzed accidents.
2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed performance based requirements are an improvement over the requirements imposed by the current technical specifications. Implementation of the proposed Steam Generator Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed change does not affect the design of the SGs, their method of operation, or primary or secondary coolant chemistry controls. The proposed change enhances SG inspection requirements.
Therefore, the proposed change does not create the possibility of a new or different type of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change is expected to result in an improvement in the tube integrity by implementing the Steam Generator Program to manage SG tube inspection, assessment, repair, and plugging. The requirements established by the Steam Generator Program are consistent with those in the applicable design codes and standards and are an improvement over the requirements in the current technical specifications.
Therefore, the margin of safety is not changed by the proposed change to the ANO–1 TSs.
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.
1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The NRC approved topical reports BAW–10227P–A, Evaluation of Advanced Cladding and Structural Material (M5) in PWR [Pressurized Water Reactor] Reactor Fuel, and BAW–10179P–A, Safety Criteria and Methodology for Acceptable Cycle Reload Analyses, provide the licensing basis for the Framatome ANP (FRA–ANP) advanced cladding and structural material, designated M5. The M5 material was shown in these documents to have equivalent or superior properties to the currently used Zircaloy-4 material. The cladding itself is not an accident initiator and does not affect accident probability. The M5 cladding has been shown to meet all 10 CFR 50.46 design criteria and, therefore, will not increase the consequences of an accident.
The proposed safety limit value ensures that fuel integrity will be maintained during normal operations and anticipated operational occurrences (AOOs), and that the design requirements will continue to be met. The core operating limits will be developed in accordance with the new methodology. The proposed safety limit value does not affect the performance of any equipment used to mitigate the consequences of an analyzed accident. There is no impact on the source term or pathways assumed in accidents previously evaluated. No analysis assumptions are violated and there are no adverse effects on the factors that contribute to offsite or onsite dose as the result of an accident.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
Use of M5 clad fuel will not result in changes in the operation or configuration of the facility. Topical report BAW–10227P–A demonstrated that the material properties of the M5 alloy are similar or better than those of Zircaloy-4. Therefore, M5 fuel rod cladding and fuel assembly structural components will perform similarly to those fabricated from Zircaloy-4, thus precluding the possibility of the fuel becoming an accident initiator and causing a new or different type of accident.
In addition, there will be no change in the level of controls or methodology used for processing radioactive effluents or handling solid radioactive waste. Since the material properties of M5 alloy are similar or better than those of Zircaloy-4, there will be no significant changes in the types of any effluents that may be released off-site. There will not be a significant increase in occupational or public radiation exposure.
The proposed safety limit value does not change the methods governing normal plant operation, nor are the methods utilized to respond to plant transients altered. The BHTP correlation is not an accident / event initiator. No new initiating events or transients result from the use of the BHTP correlation or the related safety limit changes.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed change will not involve a significant reduction in the margin of safety because it has been demonstrated that the material properties of the M5 alloy are not
The proposed safety limit value has been established in accordance with the methodology for the BHTP correlation, to ensure that the applicable margin of safety is maintained (
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the
The proposed change eliminates the TS reporting requirements to provide a monthly operating report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the Technical Specification reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change does not involve a physical alteration of the plant, add any new equipment, or require any existing equipment to be operated in a manner different from the present design. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
This is an administrative change to reporting requirements of plant operating information and occupational radiation exposure data, and has no effect on plant equipment, operating practices or safety analyses assumptions. For these reasons, the proposed change does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does not involve significance hazards consideration.
The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the
The proposed change eliminates the TS reporting requirements to provide a monthly operating report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the Technical Specification reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change does not involve a physical alteration of the plant, add any new equipment, or require any existing equipment to be operated in a manner different from the present design. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
This is an administrative change to reporting requirements of plant operating information and occupational radiation exposure data, and has no effect on plant equipment, operating practices or safety analyses assumptions. For these reasons, the proposed change does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does not involve significant hazards consideration.
1. Does the proposed change involve a significant increase in the probability or consequence of an accident previously evaluated?
SNC has chosen to reanalyze the criticality analyses for the VEGP Unit 1 and Unit 2 spent fuel racks. Westinghouse performed the revised analyses using methods that address the non-conservatisms previously identified in the current analyses. The methodologies used for the revised analysis have been previously approved for use by the NRC.
The analyses revised the enrichment, burnup, and Integral Fuel Burnable Absorber (IFBA) limits required to comply with the allowed storage configurations. The storage configurations and interface requirements in the current Technical Specifications were retained in the revised analyses. The boron dilution evaluation that supported the initial amendments to permit credit for the soluble boron at VEGP continues to remain valid. The analyses demonstrated that Keff remains below unity for the various storage configurations considered with zero soluble boron and that Keff remains less than or equal to 0.95 for the entire pool with credit for soluble boron under non-accident and accident conditions with a 95% probability at a 95% confidence level (95/95).
Core design procedures ensure that new fuel can be stored in one or more of the allowed storage configurations. Administrative controls during fuel fabrication ensure that the fuel is fabricated accordingly to ensure proper loading of the fuel in the fuel assemblies. Administrative controls used to load fuel assemblies into the spent fuel pool ensure that fuel assemblies are stored in compliance with the allowed storage configurations. Fuel handling is performed under many administrative controls and physical limitations. These controls provide reasonable assurance that a criticality accident, fuel fabrication error, or fuel handling accident will not occur.
The change to the page number of Figure 5.5.6–1 on Page vi of the Table of Contents is administrative in nature.
Therefore, based on the conclusions of the above analysis, the proposed change does not involve a significant increase in the probability or consequence of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or different kind of accident from any previously evaluated?
The types of accidents previously evaluated include fuel fabrication errors, criticality accidents, and fuel handling accidents. The analyses revised the enrichment, burnup, and Integral Fuel Burnable Absorber (IFBA) limits required to comply with the allowed storage configurations. No new or other kind of accident can be postulated as a result of the revised analyses.
The change to the page number of Figure 5.5.6–1 on Page vi of the Table of Contents is administrative in nature.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant decrease in the margin of safety?
The analyses revised the enrichment, burnup, and Integral Fuel Burnable Absorber (IFBA) limits required to comply with the allowed storage configurations. The boron dilution evaluation that supported the initial amendments to permit credit for soluble boron at VEGP was shown to remain valid. The analyses demonstrated that Keff remains below unity for the various storage configurations considered with zero soluble boron and that Keff remains less than or equal to 0.95 for the entire pool with credit for soluble boron under non-accident and accident conditions with a 95% probability at a 95% confidence level (95/95).
The change to the page number of Figure 5.5.6–1 on Page vi of the Table of Contents is administrative in nature.
Therefore, the proposed change does not involve a significant decrease 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.
1. Does the proposed Technical Specification change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
Neither Reactor Motor Operated Valve (RMOV) Boards D and E, the equipment they power, nor the automatic power transfer
The proposed deletion of the requirement to maintain an automatic transfer capability for the power supply to the LPCI inboard injection and recirculation pump discharge valves does not change the number of Emergency Core Cooling System (ECCS) subsystems credited in the BFN licensing basis. Therefore, the proposed TS changes will not significantly increase the consequences of an accident previously evaluated.
2. Does the proposed Technical Specification change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed deletion of the requirement to maintain an automatic transfer capability for the power supply to the LPCI inboard injection and recirculation pump discharge valves does not introduce new equipment, which could create a new or different kind of accident. No new external threats, release pathways, or equipment failure modes are created. Therefore, the proposed deletion of the requirement to maintain an automatic transfer capability for the power supply to the LPCI inboard injection and recirculation pump discharge valves will not create a possibility for an accident of a new or different type than those previously evaluated.
3. Does the proposed Technical Specification change involve a significant reduction in a margin of safety?
Response: No.
The proposed deletion of the requirement to maintain an automatic transfer capability for the power supply to the LPCI inboard injection and recirculation pump discharge valves does not change the number of ECCS subsystems credited in the BFN licensing basis. The requirements of 10 CFR 50.46 and Appendix K continue to be met. Therefore, the proposed change does 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.
1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The safety-related function of the CAD system is to mitigate the effects of a loss-of-coolant-accident (LOCA) by limiting the volumetric concentration of oxygen in the primary containment atmosphere. The CAD System is not an event initiator, therefore, the probability of the occurrence of an accident is not affected by this proposed Technical Specification change. Emergency procedures preferentially use the normal containment inerting system to provide post accident vent and purge capability, with the CAD system only serving in a backup role to this system. Hence, in the event of the inoperability of both CAD subsystems, the proposed TS require the normal containment inerting system to be verified available as an alternate oxygen control means. Therefore, the proposed TS change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not introduce new equipment, which could create a new or different kind of accident. This proposed change does not result in any changes to the CAD equipment design or capabilities or to the operation of the plant. No new external threats, release pathways, or equipment failure modes are created. Therefore, the implementation of the proposed change will not create a possibility for an accident of a new or different type than those previously evaluated.
3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
As stated in GL [Generic Letter] 84–09, a Mark I type boiling water reactor (BWR) plant does not rely upon purge/repressurization systems such as CAD as its primary means of hydrogen control when the unit is operated in accordance with certain technical criteria. The BFN units are operated in accordance with these criteria. The BFN Unit 1 containment is inerted with nitrogen during normal operation, nitrogen from the containment inerting system with a backup from the CAD system is used for pneumatically operated components inside containment, and there are no potential sources of oxygen generation inside containment other than the radiolytic decomposition of water. The system preferred by the Emergency Operating Instructions (EOIs) for oxygen control post-accident is the normal primary containment inerting system. Because the probability of an accident involving hydrogen and oxygen production is small, CAD is not the primary system used to mitigate the creation of combustible containment atmosphere mixtures, and because the requested LCO where both CAD subsystems is inoperable is not long, no significant reduction in the margin of safety is associated with this proposed amendment.
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.
1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
No. TVA's proposed TS revisions do not involve a significant increase in the probability of any accidents previously evaluated. TVA's proposed TS revisions provide improvements to the RCS and ECCS requirements to include appropriate reference to SQN's PTLR [PressureTemperature Limits Report] requirements. The proposed revision is a TS improvement that remains consistent with the improved standard TS requirements for Pressurized Water Reactors (PWRs) (NUREG–1431, Revision 3). TVA's proposed revision to delete SQN TS 3/4.4.2.1, “Reactor Coolant Safety Valves—Shutdown,” does not involve a significant increase in the probability of any accident previously evaluated. Pressurizer code safety valve requirements are not applicable for plant shutdown conditions (
The proposed revisions are not the result of changes to plant equipment, test methods or operating practices. The proposed changes do not contribute to the generation or assumptions for postulated accidents. The proposed changes do not affect the design basis accidents or their assumptions. The revisions to SQN TSs continue to support SQN's required safety functions.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
No. The proposed revisions are not the result of changes to plant equipment or plant design. The proposed revisions adopt standard TS requirements that are consistent with SQN's safety analysis and design and provide improvements over the existing requirements. The safety functions of the RCS and ECCS remain unchanged and do not affect any assumptions in SQN's accident analyses.
TVA's proposed change to delete the mode 4 and mode 5 TS requirements for pressurizer safety valves is consistent with the Policy Criterion of 10 CFR 50.36. The pressurizer code safety valves are not assumed to function for any safety analysis in modes 4 and 5 and consequently, the proposed changes do not create the possibility of a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a margin of safety?
No. The proposed TS change does not involve a significant reduction in a margin of safety. TVA's proposed revisions will not result in changes to system design features or plant features that could be precursors to accidents or potential degradation of accident mitigation systems. The proposed changes to the RCS and ECCS requirements remain consistent with the current TS requirements for equipment operability. Therefore, the proposed changes do not involve a significant reduction in the margin of safety.
TVA's proposed change that removes the requirement for a pressurizer safety valve in modes 4 and 5 does not affect any margin of safety because the lift setting of the pressurizer code safety valves (2485 pounds per square inch gauge [psig] ±3 percent) is well above the limit needed to protect the RCS during low temperature operation and would not provide any safety function for overpressure protection in the lower modes. The TS requirements associated with low temperature operation are governed by SQN TS 3/4.4.12, LTOP system. The LTOP system provides the necessary overpressure protection for SQN's RCS in modes 4 and 5.
Accordingly, TVA's proposed deletion of operability requirements for SQN's pressurizer code safety valves for modes 4 and 5 will not affect the margin of safety.
The United States Nuclear Regulatory Commission (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.
The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration.
For details, see the individual notice in the
During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for A Hearing in connection with these actions was published in the
Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated.
For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 25, 2004.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 15, 2004.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 15, 2004.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 27, 2004.
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 19, 2004.
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 14, 2004.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 15, 2004.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 21, 2004.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 25, 2004.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 15, 2004.
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 20, 2004.
The Commission's related evaluation of the amendment is contained in a safety evaluation dated October 21, 2004.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 22, 2004.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 28, 2004.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 20, 2004.
For the Nuclear Regulatory Commission.