Federal Register: April 16, 2002 (Volume 67, Number 73)

DOCID: FR Doc 02-8866

SUBJECT CATEGORY:

Biweekly Notice Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

DOCUMENT SUMMARY:

I. Background

Pursuant to Public Law 97415, the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. Public Law 97415 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 March 22, 2002 through April 4, 2002. The last biweekly notice was published on April 2, 2002 (67 FR 15619). 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 30day 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 30day 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 Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 205550001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the NRC's Public Document Room (PDR), located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

By May 16, 2002, 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 NRC's PDR, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records
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will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/readingrm/doccollections/cfr/. 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 205550001, Attention: Rulemaking and Adjudications Staff, or may be delivered to the Commission's PDR, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland, by the above date. A copy of the petition should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 205550001, 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 PDR, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/readingrm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC PDR Reference staff at 1 8003974209, 3044154737 or by email to pdr@nrc.gov. Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50317 and 50 318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of amendments request: January 31, 2002.

Description of amendments request: The proposed amendments would change the method of verifying the boron concentration of each safety injection tank. Rather than taking a sample from each tank every 31 days, the proposed change would require leakage into the tanks to be monitored every 12 hours and a sample be taken every 6 months.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Would not involve a significant increase in the probability or consequences of an accident previously evaluated.

Boron concentration is controlled in the safety injection tanks (SITs) to prevent either excessive boron concentrations or insufficient boron concentrations. Postlossofcoolant accident (LOCA) emergency procedures directing the operator to establish simultaneous hot and cold leg injection are based on the worst case minimum boron precipitation time. Maintaining the maximum SIT boron concentration within the upper limit ensures that the SITs do not invalidate this calculation. The minimum boron requirements of 2300 ppm [parts per million] are based on beginningoflife reactivity values and are selected to ensure that the reactor will remain subcritical during the reflood stage of a large break LOCA. During a large break LOCA, all control element assemblies are assumed not to insert into the core, and the initial reactor shutdown is accomplished by void formation during blowdown.
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Sufficient boron concentration must be maintained in the SITs to prevent a return to criticality during reflood. Level and pressure instrumentation is provided to monitor the availability of the tanks during plant operation.

The Technical Specification Surveillance Requirement (SR 3.5.1.4) verifies that the boron concentration remains within the required range by sampling. Currently, the boron concentration in each SIT is required to be verified by taking a sample of the water in the SIT every 31 days. A containment entry is required to take a sample from each of the four SITs. In addition, the boron concentration of the water added to the SITs is also sampled at the discharge of the high pressure safety injection pump to ensure that the water being added to the SITs is within the required boron concentration limits prior to being added. All intentional sources of level increase have their boron concentrations administratively maintained to ensure that the SIT boron concentrations are within Technical Specification limits. However, the Reactor Coolant System boron concentration is lower during power operation than the boron concentration in the SITs. Two check valves in series prevent leakage from the Reactor Coolant System into the SITs.

This proposed amendment would require inleakage monitoring to be done every twelve hours in addition to taking samples from each SIT every six months. Samples would continue to be taken to verify the inleakage observations remain conservative. In addition, the requirement to sample the discharge of the operating high pressure safety injection pump prior to filling the SIT would remain.

As noted above, the SITs are used only to respond to an accident and are not an accident initiator. Therefore, the probability of an accident has not increased.

The engineering analysis and risk insights combine to demonstrate that the method of SIT boron concentration verification can be changed from sampling very 31 days to monitoring inleakage every twelve hours and sampling every six months. The inleakage monitoring is based on a calculation method that has sufficient conservatism to predict the boron concentration of the SITs as shown by sample. Therefore, the SITs would remain capable of responding to an accident as described above and the consequences of an accident previously evaluated are not increased.

Therefore the probability or consequences of an accident previously evaluated are not increased.

2. Would not create the possibility of a new or different [kind] of accident from any accident previously evaluated.

The proposed change does not alter the function of any equipment, nor has it to operate differently than it was designed to operate. All equipment required to mitigate the consequences of an accident would continue to operate as before. The proposed change alters the method of verification of the SIT boron concentration, but not the boron concentration requirements themselves.

Therefore, this change does not create the possibility of a new or different [kind] of accident from any accident previously evaluated.

3. Would not involve a significant reduction in a margin of safety.

The margin of safety defined by 10 CFR [Code of Federal Regulations] Part 100 has not been significantly reduced. The inleakage monitoring done to verify the concentration of boron in the SITs, is sufficiently conservative to ensure that the boron concentration would be underpredicted, leading to attempts to increase the boron concentration or a need to sample the affected SIT. Sampling of the SITs every six months will continue to be done to ensure that the inleakage monitoring remains conservative and representative. Water added to the SITs will also continue to be sampled to ensure that it meets the minimum boron concentrations. If the boron concentration is maintained in the SITs, the system operates as assumed in the Updated Final Safety Analysis Report Chapter 14 analyses and the analyses continue to meet the dose consequences acceptance criteria given in the Updated Final Safety Analysis Report.

Therefore, this proposed change does not involve a significant reduction in [a] margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendments request involves no significant hazards consideration.

Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Section Chief: Joel Munday, Acting.
Detroit Edison Company, Docket No. 50341, Fermi 2, Monroe County, Michigan

Date of amendment request: February 21, 2002.

Description of amendment request: The proposed amendment involves changes to the Fermi 2 Updated Final Safety Analysis Report (UFSAR) and Technical Requirements Manual which is incorporated by reference in the UFSAR to eliminate the chlorine detection function from the control room heating, ventilation, and air conditioning system. Changes to the UFSAR are subject to the requirements of 10 CFR 50.59; however, these changes are being submitted for Nuclear Regulatory Commission (NRC) review and approval since they involve the elimination of an automatic action in accordance with the Nuclear Energy Institute guidance document 9607, Revision 1.

1. The change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The chlorine detection system was originally added to the plant design when it was assumed that a chlorine rail car would be located on site for use in water treatment purposes; however, oneton chlorine cylinders were used instead. In 1992, the use of chlorine for on site water treatment was discontinued. There is no chlorine stored on site and no significant amounts are stored at any other facility within the 5mile radius of the plant. The only credible accident involving a chlorine release that could be carried into the control room is from a chlorine rail car accident on the three railroad tracks 3.4 to 3.8 miles away from the site. The probability of a rail car accident and spill of chlorine is not affected by the removal of the chlorine detectors located in the normal air intake for the CCHVAC [control room heating, ventilation and air conditioning] system; therefore, only the consequences of the event must be addressed as a result of the proposed change.

The chlorine detectors in the control room ventilation air intake are intended to provide protection to the control room occupants in the event of an accidental offsite chlorine release. Detroit Edison has performed a probabilistic risk assessment to determine the probability of reaching toxic chlorine concentration levels of 10 parts per million in the control room as a result of a chlorine railcar accident and spill within 5 miles of the plant. The probability analysis took no credit for any automatic or manual action to
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isolate the control room. The results of the analysis show that the total probability of 8.4E07 per year is below the 1.0E06 threshold specified in Regulatory Guide (RG) 1.78, Revision 1. Therefore, since the probability analysis results meet the RG criteria, the elimination of the chlorine detection function will not significantly increase the consequences of an offsite chlorine release.

2. The change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The probabilistic risk assessment evaluation demonstrates that the likelihood of creating hazardous conditions in the control room as a result of a chlorine accident is very small. RG 1.78, Revision 1, states that events of such low frequencies do not need to be considered in the plant design because the resultant low levels of radiological risk are considered acceptable. The probabilistic assessment assumed no automatic or manual action to isolate the control room or to filter outside air before it is discharged in the control room. The evaluation did not rely on any structure, system or component to perform a specific function; therefore, the elimination of the chlorine detection system does not create the potential for a new or different kind of accident from any accident previously evaluated.

3. The change does not involve a significant reduction in the margin of safety.

The elimination of the chlorine detection system will not affect the protection of the control room operators from the hazard of an offsite chlorine release. No significant amounts of chlorine are stored within 5 miles of the plant and the only chlorine accident risk is from a railroad car accident over 3 miles away. The probabilistic evaluation demonstrates the low risk associated with a chlorine accident that would incapacitate the operators such that their functions in mitigating a radiological event are impacted. Since the Regulatory Positions in RG 1.78, Revision 1 are satisfied, deletion of the chlorine detection system will not result in a significant reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 482261279.

NRC Section Chief: William D. Reckley, Acting.
Dominion Nuclear Connecticut, Inc., Docket No. 50336, Millstone Nuclear Power Station, Unit No. 2, New London County, Connecticut

Date of amendment request: February 5, 2002

Description of amendment request: The proposed amendment would revise the surveillance requirements associated with the Containment Isolation Valves (CIVs), Reactor Building Closed Cooling Water (RBCCW) System, and Service Water (SW) System. The proposed changes would remove redundant testing requirements that are already addressed by the Inservice Testing (IST) Program, which is required pursuant to Technical Specification 4.0.5, and would use Technical Specification 4.0.5 to control the specific acceptance criteria and frequency of test performance. Additional proposed changes would remove the post maintenance testing requirements associated with the CIVs, revise the wording of the RBCCW and SW Systems Limiting Conditions for Operation, and increase the allowed outage times for the RBCCW and SW Systems.

1. Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed Technical Specification changes associated with the limiting condition for operation requirements, surveillance requirements, and allowed outage times will not cause an accident to occur and will not result in any change in the operation of the associated accident mitigation equipment. The ability of the equipment associated with the proposed changes to mitigate the design basis accidents will not be affected. The proposed changes to the limiting condition for operation requirements will not affect the equipment operability requirements. The proposed surveillance requirements are adequate to ensure proper operation of the associated accident mitigation equipment. Proper operation of the containment isolation valves will still be verified, as appropriate, following maintenance activities. The proposed allowed outage times are reasonable and consistent with standard industry guidelines to ensure the accident mitigation equipment will be restored in a timely manner. The design basis accidents will remain the same postulated events described in the Millstone Unit No. 2 Final Safety Analysis Report, and the consequences of those events will not be affected. Therefore, the proposed changes will not increase the probability or consequences of an accident previously evaluated.

The additional proposed changes to the Technical Specifications (e.g., combining requirements, deleting an expired footnote, and renumbering a requirement) will not result in any technical changes to the current requirements. Therefore, these additional proposed changes 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.

The proposed changes to the Technical Specifications do not impact any system or component that could cause an accident. The proposed changes will not alter the plant configuration (no new or different type of equipment will be installed) or require any unusual operator actions. The proposed changes will not alter the way any structure, system, or component functions, and will not alter the manner in which the plant is operated. There will be no effect on plant operation or accident mitigation equipment. The response of the plant and the operators following an accident will not be different. In addition, the proposed changes do not introduce any new failure modes. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously analyzed.

3. Involve a significant reduction in a margin of safety.

The proposed Technical Specification changes associated with the limiting condition for operation requirements, surveillance requirements, and allowed outage times will not cause an accident to occur and will not result in any change in the operation of the associated accident mitigation equipment. The equipment associated with the proposed Technical Specification changes will continue to be able to mitigate the design basis accidents as assumed in the safety analysis. The proposed surveillance requirements are adequate to ensure proper operation of the affected accident mitigation equipment. The proposed allowed outage times are reasonable and consistent with standard industry guidelines to ensure the accident
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mitigation equipment will be restored in a timely manner. In addition, the proposed changes will not affect equipment design or operation, and there are no changes being made to the Technical Specification required safety limits or safety system settings. The proposed Technical Specification changes, in conjunction with existing administrative controls (e.g., IST Program), will provide adequate control measures to ensure the accident mitigation functions are maintained. Therefore, the proposed changes will not result in a reduction in a margin of safety.

The additional proposed administrative changes to the Technical Specifications (e.g., combining requirements, deleting an expired footnote, and renumbering a requirement) will not result in any technical changes to the current requirements. Therefore, these additional changes will not result in a reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 06385.

NRC Section Chief: James W. Clifford.
Dominion Nuclear Connecticut, Inc., Docket Nos. 50336 and 50423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London County, Connecticut

Date of amendment request: February 14, 2002.

Description of amendment request: The proposed Technical Specification (TS) changes will relocate selected Millstone Units 2 and 3 TSs related to the Reactor Coolant System (RCS) and Plant Systems to the Technical Requirements Manual (TRM). The proposed TSs for Unit 2 include 3/4.4.9.1, ``Pressure/Temperature Limits,'' 3/4.7.2, ``Steam Generator Pressure/Temperature Limitation,'' 3/4.7.5, ``Flood Level,'' 3/4.7.7, ``Sealed Source Contamination,'' 3/4.7.8, ``Snubbers,'' and related Tables, Figures, and Bases sections. The proposed TSs for Unit 3 include 3/4.4.9.1, ``Pressure/Temperature Limits,'' 3/4.7.2, ``Steam Generator Pressure/Temperature Limitation,'' 3/4.7.6, ``Flood Protection,'' 3/4.7.10, ``Snubbers,'' 3/4.7.11, ``Sealed Source Contamination,'' 3/4.7.14, ``Area Temperature Monitoring,'' and corresponding Tables, Figures, and Bases sections.

1. Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed technical specification changes will relocate to the TRM the following items: surveillance requirements for the withdrawal of reactor vessel material irradiation specimens of Millstone Unit Nos. 2 and 3 which are part of the Pressure/Temperature Limits technical specifications, Millstone Unit Nos. 2 and 3 technical specifications covering Steam Generator Pressure/Temperature Limitation, Flood Level, Sealed Source Contamination, and Snubbers. Also the Millstone Unit No. 3 technical specification covering Area Temperature Monitoring will be relocated to the TRM. Since the relocated requirements remain the same, the proposed changes will have no effect on plant operation, or the availability or operation of any accident mitigation equipment. Therefore, the relocation of the requirements associated with these technical specifications will not impact an accident initiator and cannot cause an accident. These changes 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.

The proposed technical specification changes will relocate the requirements of selected Millstone Unit Nos. 2 and 3 technical specifications as described above to the TRM. The proposed changes do not alter the plant configuration (no new or different type of equipment will be installed) or require any new or unusual operator actions. Since the requirements remain the same, the proposed changes do not alter the way any system, structure, or component functions and do not alter the manner in which the plant is operated. The proposed changes do not introduce any new failure modes. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Involve a significant reduction in a margin of safety.

The proposed technical specification changes will relocate to the TRM the following items: surveillance requirements for the withdrawal of reactor vessel material irradiation specimens of Millstone Unit Nos. 2 and 3 which are part of the Pressure/Temperature Limits technical specifications, Millstone Unit Nos. 2 and 3 technical specifications covering Steam Generator Pressure/Temperature Limitation, Flood Level, Sealed Source Contamination, and Snubbers. Also the Millstone Unit No. 3 technical specification covering Area Temperature Monitoring will be relocated to the TRM. Since the proposed changes are solely to relocate the existing requirements, the proposed changes will have no effect on plant operation, or the availability or operation of any accident mitigation equipment. The plant response to the Design Basis Accidents will not change. Therefore, there will be no reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Waterford, CT 061415127.

NRC Section Chief: James W. Clifford.
Nine Mile Point Nuclear Station, LLC, Docket No. 50220, Nine Mile Point Nuclear Station Unit No. 1, Oswego County, New York

Date of amendment request: March 15, 2002.

Description of amendment request: The proposed amendment would revise the Nine Mile Point Unit 1 Technical Specifications (TSs), Table 4.6.4, ``Shock Suppressors (Snubbers),'' consistent with the model snubber visual inspection and acceptance requirements conveyed in Generic Letter 9009, ``Alternative Requirements for Snubber Visual Inspection and Corrective Actions.''

1. The operation of Nine Mile Point Unit 1 in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.

Snubbers are utilized at Nine Mile Point Unit 1 (NMP1) to ensure the
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structural integrity of the reactor coolant system and other safety related (as well as certain nonsafety related) systems during and following a seismic event or other event initiating dynamic loads. The proposed change to the snubber visual inspection schedule is based on that delineated in NRC [Nuclear Regulatory Commission] Generic Letter (GL) 9009, ``Alternative Requirements for Snubber Visual Inspection and Corrective Actions.'' This change does not modify any accident initiators or change any equipment or procedures used to limit the consequences of any accidents previously evaluated.

Accordingly, the proposed amendment will not significantly increase the probability or consequences of an accident previously evaluated.

2. The operation of Nine Mile Point Unit 1 in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

No physical modifications are being made to any snubbers or to any systems supported by snubbers by this proposed amendment. No method of plant or system operation is varied by use of the alternate snubber visual inspection schedule delineated in GL 9009. Only the method utilized to determine future surveillance intervals for snubber visual inspections based on the previous inspection results is changed by the proposed amendment. This method was developed and published by the NRC in GL 9009 for generic application at nuclear power plants.

Accordingly, the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

The operation of Nine Mile Point Unit 1, in accordance with the proposed amendment, will not involve a significant reduction in a margin of safety.

In GL 9009, the NRC staff determined that use of the alternate snubber visual inspection schedule by nuclear power plants will maintain the same level of confidence as the previous schedule required by the plants' Technical Specifications. GL 9009 also recognized that snubber visual inspection is a complementary process to snubber functional testing and provides additional confidence in snubber operability. Snubber functional testing is not being modified by this proposed amendment.

Therefore, the proposed change will not adversely affect any structure, system, component, or function that is safetyrelated or important to safety. Accordingly, the proposed amendment 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.

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1400 L Street, NW., Washington, DC 200053502.

NRC Section Chief: Joel Munday, Acting.
Nuclear Management Company, LLC, Docket No. 50305, Kewaunee Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: March 19, 2002.

Description of amendment request: The proposed amendment would revise the Kewaunee Nuclear Power Plant accident source term used for design basis radiological analyses.

1. Involve a significant increase in the probability or consequences of an accident previously evaluated.

Differences between the original source term and the proposed AST [accident source term] cannot affect the previously analyzed core damage frequency (CDF) and large early release frequency (LERF). Since there are no modifications proposed with this request for AST, Limiting Safety System Settings and Safety Limits specified in the Technical Specifications remain unchanged. Reanalysis of design basis accidents as described herein demonstrates that regulatory dose acceptance criteria continue to be satisfied. Thus, nothing in this proposal will cause an increase in the probability or consequence of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

There are no physical changes to the plant associated with this request, and the plant conditions for which [Nuclear Management Company] (NMC) evaluated designbasis accidents remain valid. Consequently, this proposal introduces no new failure modes. Thus, this proposal does not create the possibility of a new or different kind of accident.

3. Involve a significant reduction in the margin of safety.

The revised designbasis accident offsite and controlroom dose calculations proposed herein remain within regulatory acceptance criteria set forth in 10 CFR 100 and 10 CFR 50 Appendix A, General Design Criterion 19. They also use the TEDE [total effective dose equivalent] dose acceptance criteria as directed by the Commission. An acceptable margin of safety is inherent in the limits described thereby. Thus, changes proposed by this request 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.

Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, P.O. Box 1497, Madison, WI 537011497.

NRC Section Chief: William D. Reckley, Acting.
Nuclear Management Company, LLC, Docket Nos. 50266 and 50301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin

Date of amendment request: February 28, 2002.

Description of amendment request: The proposed amendment would revise technical specifications (TS) 1.1, ``Definitions,'' ``CREFS Actuation Instrumentation,'' TS 3.4.16, ``RCS Specific Activity,'' TS 3.3.5, ``CREFS Actuation Instrumentation,'' TS 3.4.16, ``RCS Specific Activity,'' TS 3.7.9, ``CREFS,'' and TS 3.7.13, ``Secondary Specific Activity,'' and delete TS 3.9.3, ``Containment Penetrations.''

The accident source term used in the selection of the designbasis offsite and control room dose analysis would be replaced by the implementation of an alternative source term.

The specific TS changes would be as follows: (1) TS 1.1, ``Definitions:'' Revise the definition of La (containment leakage) by changing 0.4 percent to 0.2 percent. (2) TS 3.3.5, ``CREFS Actuation Instrumentation:'' Revise table 3.3.51 to indicate that either RE101 or RE235 must be operable to ensure that the control room radiation instrumentation necessary to initiate the CREFS emergency makeup mode is operable. Add the Control Room Area Monitor and Control Room Air Intake trip setpoints to Note ``d'' of table 3.3.51.
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(3) TS 3.4.16, ``RCS Specific Activity:'' Revise LCO Action Condition A to indicate 1.0 Ci/gm as the maximum reactor coolant dose equivalent iodine 131 (DE I131) value. Revise Figure 3.4.161 to indicate 60 Ci/gm DE I131 as the maximum RCS limit for operations at or above 80 percent of rated thermal power. Revise SR 3.4.16.2 to verify 1.0 Ci/gm as the maximum reactor coolant DE I131 value. (4) TS 3.7.9, ``CREFS:'' Delete SR 3.7.9.5. (5) TS 3.7.13, ``Secondary Specific Activity:'' Revise LCO 3.5.13 and SR 3.7.13 to indicate that the secondary specific activity shall be less than or equal to 0.1 Ci/gm. (6) TS 3.9.3, ``Containment

Penetrations:'' Delete Section 3.9.3.

1. Involve a significant increase in the probability or consequences of any accident previously evaluated.

The Alternative Source Term (AST) and those plant systems affected by implementing the proposed changes to the TS are not accident initiators and cannot increase the probability of an accident. The AST does not adversely affect the design or operation of the facility in a manner that would create an increase [in] the probability of an accident. Rather, the AST is used to evaluate the dose consequences of a postulated accident. The revised dose calculations, except those for LOCA, use the values in the proposed TS. The limiting design bases accidents at PBNP have been evaluated for implementation of the AST.

These analyses have demonstrated that, with the proposed changes, the dose consequences meet the regulatory acceptance criteria of 10 CFR 50.67 and RG 1.183. A comparison of the current offsite dose calculations to the revised offsite dose calculations indicate that the proposed changes will not result in a significant increase in the predicted dose consequences for any of the analyzed accidents. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of any of the selected previously analyzed accidents.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed amendment will not create the possibility for a new or different type of accident from any accident previously evaluated. Changes to the allowable activity in the primary and secondary systems do not result in changes to the design or operation of these systems. The evaluation of the effects of the proposed changes indicates that all design standard and applicable safety criteria limits are met.

The systems affected by the changes are used to mitigate the consequences of an accident that has already occurred. The proposed TS changes and modifications do not significantly affect the mitigative function of these systems. Equipment important to safety will continue to operate as designed. Component integrity is not challenged. The changes do not result in any event previously deemed incredible being made credible. The changes do not result in more adverse conditions or result in any increase in the challenges to safety systems.

Therefore, the proposed changes do not create the possibility of a new or different type of accident from any accident previously evaluated.

3. Involve a significant reduction in a margin of safety.

The implementation of the proposed changes does not significantly reduce the margin of safety. These changes have been evaluated in the revisions to the analysis of the consequences of the design basis accidents for PBNP. The radiological analysis results in concert with the proposed TS changes, meet the regulatory acceptance criteria of 10 CFR 50.67 and RG 1.183. These acceptance criteria have been developed for the purpose of use in design basis accident analyses such that meeting these limits demonstrates adequate protection of public health and safety. The proposed changes will not degrade the plant protective boundaries, will not cause a release of fission products to the public and will not degrade the performance of any SSCs important to safety.

Therefore, the proposed changes to the TS would 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.

Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Section Chief: William D. Reckley, Acting.
Nuclear Management Company, LLC, Docket Nos. 50282 and 50306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, Minnesota

Date of amendment request: February 2, 2001, supplemented August 31, 2001.

Description of amendment request: The proposed amendments would revise the technical specifications (TSs) to clarify the plant conditions under which various specifications are applicable. The licensee stated in its amendment request that a literal reading of the current technical specifications wording may result in situations where a routine plant shutdown would seem to be prohibited by TSs and, thereby, require entry into TS 3.0.C. This amendment request also makes several administrative changes to the TSs, including revising references to the Chief Nuclear Corporate Officer, capitalizing defined terms, and updating references to previously relocated TS paragraphs and correcting the List of Figures. The licensee's supplement to the amendment request, dated August 31, 2001, proposed a correction of a typographical error in TS Table 3.52B, Action 33.

1. Does operation of the facility with the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed changes are administrative in nature and clarify existing specifications without reducing or altering the requirements imposed by existing specifications. The proposed changes do not significantly affect any system that is a contributor to initiating events for previously evaluated accidents. Neither do the changes significantly affect any system that is used to mitigate any previously evaluated accidents. Therefore, the proposed changes do not involve any significant increase in the probability or consequence of an accident previously evaluated.

2. Does operation of the facility with the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes are administrative in nature and clarify existing specifications without reducing or altering the requirements imposed by existing specifications. The proposed
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changes do not alter the design, function, or operation of any plant component and do not install any new or different equipment, therefore a possibility of a new or different kind of accident from those previously analyzed has not been created.

3. Does operation of the facility with the proposed amendment involve a significant reduction in a margin of safety?

The proposed changes are administrative in nature and clarify existing specifications without reducing or altering the requirements imposed by existing specifications. Thus, the proposed change[s] do not involve a significant reduction in the margin of safety associated with the safety limits inherent in either the princip[al] barriers to a radiation release (fuel cladding, RCS [reactor coolant system] boundary, and reactor containment), or the maintenance of critical safety functions (subcriticality, core cooling, ultimate heat sink, RCS inventory, RCS boundary integrity, and containment integrity).

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 requests involve no significant hazards consideration.

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Section Chief: William D. Reckley, Acting.
Southern California Edison Company, et al., Docket Nos. 50361 and 50 362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego County, California

Date of amendment requests: March 11, 2002.

Description of amendment requests: The proposed amendment would revise the Technical Specifications (TSs) for San Onofre Nuclear Generating Station, Units 2 and 3. Specifically, TS Section 1.1, Definitions, would be revised to change the definition of response time testing as it is applied to the Engineered Safety Features, and the Reactor Protective System. The proposed change is based on approved Technical Specification Task Force (TSTF) Traveler TSTF368, Revision 0, ``Incorporate Combustion Engineering Owners Group (CEOG) Topical Report to Eliminate Pressure Sensor Response Time Testing.''

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed amendment to the Technical Specification (TS) Definitions for Engineered Safety Feature (ESF) Response Time and Reactor Protective System (RPS) Response Time allows substitution of an allocated sensor response time in lieu of measuring sensor response time. Response time testing is not an initiator of any accident previously evaluated. Further, overall system response time will continue to meet Technical Specification requirements. The allocated sensor response times allowed in lieu of measurement have been determined to adequately represent the response time of the components such that the safety systems utilizing those components will continue to perform their accident mitigation function as assumed in the safety analysis. Therefore, the proposed change does not involve a significant increase in the probability or consequences of any 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 amendment to TS Section 1.1, ``Definitions,'' allows the substitution of an allocated sensor response time in lieu of sensor response time testing for selected components. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. 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 amendment to TS Section 1.1, ``Definitions,'' allows the substitution of an allocated sensor response time in lieu of measured sensor response time for certain pressure sensors. The allocated pressure sensor response times allowed in lieu of measurement have been determined to adequately represent the response time of the components such that the safety systems utilizing those components will continue to perform their accident mitigation function as assumed in the safety analysis. 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 requests involve no significant hazards consideration.

Attorney for licensee: Douglas K. Porter, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770.

NRC Section Chief: Stephen Dembek.
Tennessee Valley Authority, Docket Nos. 50327 and 50328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendments: March 4, 2002 (TS 0004).

Brief description of amendments: The proposed amendment would change the Sequoyah (SQN) Unit 1 and 2 Technical Specification (TS) to relocate the current requirements for ice condenser ice bed temperature and inlet door position monitoring systems to the SQN Technical Requirements Manual (TRM). These relocated specifications are consistent with the latest version of the improved Standard TS (NUREG 1431). The affected functions have been evaluated in accordance with Title 10 of the Code of Federal Regulations, Section 50.36 (10 CFR 50.36) for applicability to the criteria for requirements that must be retained in the TS. In each case, the four criteria of 10 CFR 50.36 did not apply to these functions. This revision will provide better consistency between the SQN TS and NUREG1431.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), Tennessee Valley Authority (TVA), the licensee, has provided its analysis of the issue of no significant hazards consideration, which is presented below:

A. The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed revision relocates the ice bed temperature monitoring system and the inlet door position monitoring system to the TRM. Relocation to the TRM continues to provide an acceptable level of applicability to plant operation and requires revisions to be processed in accordance with the provisions in 10 CFR 50.59. Evaluations of revisions in accordance with 10 CFR 50.59 will continue to ensure that these specifications adequately control the
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functions of ice bed temperature and inlet door positions to maintain safe operation of the plant. These systems are not postulated to be the initiator of a design basis accident. Since there are no changes to these functions and their operation will remain the same, the probability of an accident is not increased by relocating these requirements to the TRM. Additionally, the accident mitigation capability and offsite dose consequences associated with accidents will not change because these functions will not be altered by the proposed relocation. Therefore, the consequences of an accident are not increased by this relocation to the TRM and the control of revisions to these specifications in accordance with 10 FR 50.59.

B. The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed revision will not alter the functions for the ice bed temperature or inlet door positions such that accident potential would be changed. The location of these specifications in the TRM and the performance of revisions in accordance with 10 CFR 50.59 will continue to maintain acceptable operability requirements. Therefore, the possibility of an accident of a new or different kind is not created by the proposed relocation and deletion.

C. The proposed amendment does not involve a significant reduction in a margin of safety.

The proposed specification relocation will not affect plant setpoints or functions that maintain the margin of safety. This is based on the relocation to the TRM. The TRM continues to maintain the same level of operability requirements and surveillance testing to adequately ensure functionality of the ice bed temperature monitoring system and the inlet door position monitoring system. The TRM is controlled in accordance with requirements of 10 CFR 50.59. Therefore, the proposed relocation and deletion is acceptable and will not reduce the margin of safety.

The NRC has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.

NRC Section Chief: Richard P. Correia.
Tennessee Valley Authority, Docket Nos. 50327 and 50328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendments: March 4, 2002 (TS 0103)

Brief description of amendments: The proposed amendment would change the Sequoyah (SQN) Unit 1 and 2 Technical Specifications (TSs) to delete one definition and modify several subsections contained in TS Section 6.0, Administrative Controls. These proposed changes have been prepared based on existing NRC guidance. The changes are being proposed in the following areas:

Definition 1.17``Member(s) of the Public.'' (NUREG1431, Revision 2)

TS 6.2.2.g, Overtime. (TS Travelers Form (TSTF)258, Revision 4)

TS 6.3, Facility Staff Qualifications. (TSTF258, Revision 4)

TS 6.8.4.a.ii, Primary Coolant Sources Outside Containment. (TSTF299)

TS 6.8.4.f, Radioactive Effluent Controls Program. (TSTF 258, Revision 4 and TSTF308, Revision 1)

TS 6.8.4.i, Deletion of the ``Configuration Risk Management Program.'' (10 CFR 50.65)

The second paragraph in TS 6.9.1.5 associated with specific activity limits. (NUREG1431, Revision 2)

TS 6.9.1.14, Monthly Reactor Operating Report contents revision. (TSTF258, Revision 4)

TS 6.12, High Radiation Areas revision. (TSTF258, Revision 4)

TS 6.15, Deletion of Major Changes To Radioactive Waste Treatment Systems (Liquid, Gaseous, and Solid). (NUREG1431, Revision 2)

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No. The proposed changes that involve the rewording or reformatting of the existing TSs do not involve technical changes. Therefore, this change is administrative and does not affect the initiators of analyzed events or assumed mitigation of accidents or transient events.

Three of the changes remove programs from TSs based on present regulatory controls. Specifically 10 CFR 50.59, 10 CFR 50.65, 10 CFR 50.71(e), 10 CFR 50.73, and Performance Indicator data. Based on the requirements residing in existing regulations it is acceptable to remove them from TS. Additionally, any changes to these programs will be evaluated based on regulatory requirements, no significant increase in the probability or consequences of an accident previously evaluated will be allowed.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No. The proposed changes do not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No. The proposed changes will not reduce the margin of safety because they have no effect on any safety analysis assumptions. Additionally, the proposed programs to be removed from TSs are contained in existing plant programs required by existing regulations. Since any future changes to these programs will be evaluated, no significant reduction in a margin of safety will be allowed.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, TVA concludes that the proposed amendment(s) present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no significant hazards consideration'' is justified.

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.

NRC Section Chief: Richard P. Correia.
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Wolf Creek Nuclear Operating Corporation, Docket No. 50482, Wolf Creek Generating Station, Coffey County, Kansas

Date of amendment request: February 21, 2002.

Description of amendment request: The proposed amendment would revise Required Actions for Limiting Conditions for Operation (LCOs) 3.3.1, ``Reactor Trip System (RTS) Instrumentation;'' 3.4.5, ``RCS [Reactor Coolant System] LoopsMODE 3;'' 3.4.6, ``RCS LoopsMODE 4;'' 3.4.7, ``RCS LoopsMODE 5, Loops Filled;'' 3.4.8, ``RCS LoopsMODE 5, Loops Not Filled;'' 3.8.2, ``AC SourcesShutdown;'' 3.8.5, ``DC SourcesShutdown;'' 3.8.8, ``InvertersShutdown;'' 3.8.10,
``Distribution SystemsShutdown;'' 3.9.3, ``Nuclear Instrumentation;'' 3.9.5, ``Residual Heat Removal (RHR) and Coolant CirculationHigh Water Level;'' and 3.9.6, ``Residual Heat Removal (RHR) and Coolant CirculationLow Water Level'' in the Wolf Creek Generating Station Technical Specifications (TSs). The Required Actions proposed would suspend operations involving positive reactivity additions or RCS boron concentration reductions. In addition, the proposed amendment would revise Notes, for several of the above LCOs, that preclude reductions in RCS boron concentration. This amendment would revise these Required Actions and LCO Notes to allow small, controlled, safe insertions of positive reactivity, but limit the introduction of positive reactivity such that compliance with the required shutdown margin or refueling boron concentration limits will still be satisfied.

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Overall protection system performance will remain within the bounds of the previously performed accident analyses since there are no hardware changes. The RTS instrumentation and reactivity control systems will be unaffected. Protection systems will continue to function in a manner consistent with the plant design basis. All design, material, and constructio

SUMMARY:

Operating licenses, amendments; no significant hazards considerations; biweekly notices,

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