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

[Federal Register: February 17, 2004 (Volume 69, Number 31)]

[Notices]

[Page 7517-7529]

From the Federal Register Online via GPO Access [wais.access.gpo.gov]

[DOCID:fr17fe04-134]

NUCLEAR REGULATORY COMMISSION

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

1. Background

Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from, January 22, 2004, through February 5, 2004. The last biweekly notice was published on February 3, 2004 (69 FR 5200).

Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene.

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60- day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30- day comment period should circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. Should the Commission take action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. Should the Commission make a final No Significant Hazards Consideration Determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's ``Rules of Practice for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/

reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the

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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 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.

If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must be filed by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff; (3) E-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, hearingdocket@nrc.gov; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at (301) 415-1101, verification number is (301) 415-1966. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and it is requested that copies be transmitted either by means of facsimile transmission to 301-415-3725 or by email to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

Nontimely requests and/or petitions and contentions will not be entertained absent a determination by the Commission or the presiding officer of the Atomic Safety and Licensing Board that the petition, request and/or the contentions should be granted based on a balancing of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).

For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there

are problems in accessing the documents located in ADAMS, contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by email to pdr@nrc.gov.

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: December 23, 2003.

Description of amendment request: The licensee proposed to revise Section 4.5.D.2 of the Technical Specifications. This change would allow the licensee to leak test the Main Steam Isolation Valves (MSIV) at a lower pressure to eliminate the risk of lifting the disc of the inboard MSIV from its seat, producing inaccurate test data. The inboard MSIV would then have to be plugged before the leak test can be repeated. The current leak rate requirement is 0.05(0.75)La at Pa, where La is the maximum allowable leak rate, and Pa is the calculated peak containment pressure. This amendment would change this requirement to a leak rate of [le]11.9 standard cubic feet per hour (scfh) at a pressure [ge]20 psig. The leak rate of 11.9 scfh is a more conservative value based on control room habitability analysis, and 20 psig is based on the fact that post-accident pressure peaks in 2 to 3 seconds after an accident and would quickly drop below 20 psig. There is no physical changes to plant design associated with this amendment.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration. The NRC staff has reviewed the licensee's analysis against the three standards of 10 CFR 50.92(c). The NRC staff's analysis is presented below:

The first standard requires that operation of the unit in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed amendment would change the pressure at which the leak rate of the MSIV is performed, while the leak rate test standard would be made more conservative than the current standard. No hardware design change is associated with the proposed amendment. Changing the MSIV leak test criterion would have no impact on the performance of the MSIVs. Thus, the proposed amendment would create no adverse effect on the functional performance of any plant structure, system, or component (SSC). All SSCs will continue to perform their design functions with no decrease in their capabilities to mitigate the consequences of previously analyzed postulated accidents. Accordingly, the proposed amendment would lead to no increase in the consequences of an accident previously evaluated, and no increase of the probability of an accident previously evaluated.

The second standard requires that operation of the unit in accordance with the proposed amendment will not create

[[Page 7519]]

the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment is not the result of a hardware design change, nor does it lead to the need for a hardware design change. There is no change in the methods the unit is operated. As a result, all SSCs will continue to perform as previously analyzed by the licensee, and previously evaluated and accepted by the NRC staff. Therefore, the proposed amendment will not create the possibility of a new or different kind of accident from any previously evaluated.

The third standard requires that operation of the unit in accordance with the proposed amendment will not involve a significant reduction in a margin of safety. Since the licensee did not propose to exceed or alter a design basis or safety limit, and did not propose to operate any component in a less conservative manner, the proposed amendment will not affect in any way the performance characteristics and intended functions of any SSC. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the NRC staff's analysis, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the proposed amendment request involves no significant hazards consideration.

Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, LLP, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Section Chief: Richard J. Laufer.

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendments request: December 19, 2003, as supplemented January 14, 2004.

Description of amendment request: The proposed change allows entry into a mode or other specified condition in the applicability of a technical specification (TS), while in a condition statement and the associated required actions of the TS, provided the licensee performs a risk assessment and manages risk consistent with the program in place for complying with the requirements of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.65(a)(4). Limiting Condition for Operation (LCO) 3.0.4 exceptions in individual TSs would be eliminated, and Surveillance Requirement (SR) 3.0.4 revised to reflect the LCO 3.0.4 allowance.

This change was proposed by the industry's Technical Specification Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a notice of opportunity for comment in the Federal Register on August 2, 2002 (67 FR 50475), on possible amendments concerning TSTF-359, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability of the following NSHC determination in its application dated December 19, 2003, as supplemented by letter dated January 14, 2004.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below: Criterion 1--The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. Being in a TS condition and the associated required actions is not an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on required actions as allowed by proposed LCO 3.0.4, are no different than the consequences of an accident while entering and relying on the required actions while starting in a condition of applicability of the TS. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From any Previously Evaluated.

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Entering into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant Reduction in a Margin of Safety.

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. The TS allow operation of the plant without the full complement of equipment through the conditions for not meeting the TS LCO. The risk associated with this allowance is managed by the imposition of required actions that must be performed within the prescribed completion times. The net effect of being in a TS condition on the margin of safety is not considered significant. The proposed change does not alter the required actions or completion times of the TS. The proposed change allows TS conditions to be entered, and the associated required actions and completion times to be used in new circumstances. This use is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The change also eliminates current allowances for utilizing required actions and completion times in similar circumstances, without assessing and managing risk. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

The NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Steven R. Carr, Associate General Counsel-- Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Section Chief: Allen G. Howe.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina

Date of amendment request: December 8, 2003.

Description of amendment request: The amendment involves a one-time revision to the steam generator (SG) inservice inspection frequency requirements in Technical Specification 4.4.5.3a. to allow a 40-month inspection interval after the first inservice inspection following SG replacement, rather than after two consecutive inspections resulting in C-1 classification.

Basis for proposed no significant hazards consideration determination:

[[Page 7520]]

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

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

   The proposed amendment revises the steam generator inspection frequency to allow a 40-month inspection frequency after the first inservice inspection following SG replacement, rather than after two consecutive inspections resulting in C-1 classification. The ``C-1'' category is defined in the Technical Specifications as having inspection results that indicate ``less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.''

   The 100% inspection of the open steam generator tubes performed during RFO [Refueling Outage]-11 represents a quantity of tubes inspected that is significantly greater than the amount required by the Technical Specifications over two successive inspective periods (i.e., 3% of the total number of tubes in all steam generators required in the first inspection following SG replacement and the same quantity of the tubes to be examined in the second inspection). The RFO-11 100% tube inspection did not indicate the tubes had experienced degradation from the cycle of operation.

   The assessment of the condition of the steam generator tubes indicated the structural condition of the tubing had not changed during the first cycle of operation following steam generator replacement and these results that indicated the tubes would still meet their structural criteria over the proposed inspection frequency. The steam generator tube inspection meets the current industry examination guidelines without performing inspections during the next refueling outage.

   The steam generator inspection frequency extension does not introduce a new failure mode or impact any other plant systems or components. The proposed change does not alter plant design. The HNP

   [Harris Nuclear Plant] steam generator tubes do not have an active damage mechanism which could lead to the potential of primary-to- secondary steam generator leakage.

   Therefore, the proposed inspection frequency change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

   The proposed change to extend the steam generator tube inspection frequency does not impact the design or operation of the steam generators or any other plant structure, system or component. Extending the inspection frequency of the steam generator tubes does not introduce any new failure modes. The proposed change does not alter plant design basis, or alter any potential accident previously evaluated.

   The proposed change revises the steam generator inspection frequency to allow a 40-month inspection interval after the first inservice inspection following SG replacement, rather than after two consecutive inspections resulting in C-1 classification. The first steam generator inspection following replacement inspected 100% of the open tubing in all three steam generators. This inspection exceeded the existing technical specification inspection over the two consecutive inspections. This inspection indicated there was no service-induced degradation in the steam generator tubes. The HNP first cycle inspection results were comparable with other recent Westinghouse model replacement steam generators.

   Therefore, the proposed inspection frequency change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The steam generator tubes are an integral part of the reactor coolant system pressure boundary. The tubes are expected to maintain primary system pressure and inventory. The tubes are a barrier to keep radioactive fission products in the reactor coolant system from transferring to the secondary system. The steam generator tubes transfer the heat from the primary system to the secondary system. The ability of the steam generator tubes to perform these functions depends on the integrity of the tubes.

   Steam generator tube integrity is a function of design, environment, and current physical condition. Extending the steam generator tube inspection frequency by one operating cycle will not alter the function or design of the steam generators. The steam generator tube inspections performed during the first outage following steam generator replacement demonstrated that the tubes do not have an active damage mechanism, and the scope of these inspections significantly exceeded the requirements of the Technical Specifications. These inspection results were comparable to similar inspection results for second generation Alloy 690 models of replacement steam generators installed at other plants, and subsequent inspections at those plants yielded results that support this extension request. The improved design of the replacement steam generators also provides reasonable assurance that significant tube degradation is not likely to occur over the proposed operating period.

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

   Based on the above, Progress Energy Carolinas, Inc. [Carolina Power & Light Company] concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of ``no significant hazards consideration'' is justified.

   The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 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: Steven R. Carr, Associate General Counsel-- Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

   NRC Section Chief: Allen Howe.

   Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

   Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

   Date of amendment request: November 5, 2003.

   Description of amendment requests: The proposed change allows entry into a mode or other specified condition in the applicability of a Technical Specification (TS), while in a condition statement and the associated required actions of the TS, provided the licensee performs a risk assessment and manages risk consistent with the program in place for complying with the requirements of 10 CFR 50.65(a)(4). Limiting Condition for Operation (LCO) 3.0.4 exceptions in individual TS would be eliminated, and Surveillance Requirement (SR) 3.0.4 revised to reflect the LCO 3.0.4 allowance.

   This change was proposed by the industry's Technical Specification Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a notice of opportunity for comment in the Federal Register on August 2, 2002 (67 FR 50475), on possible amendments concerning TSTF-359, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability of the following NSHC determination in its application dated November 5, 2003.

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

   Criterion 1--The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

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   The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. Being in a TS condition and the associated required actions is not an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on required actions as allowed by proposed LCO 3.0.4, are no different than the consequences of an accident while entering and relying on the required actions while starting in a condition of applicability of the TS. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

   Criterion 2--The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated.

   The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Entering into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

   Criterion 3--The Proposed Change Does Not Involve a Significant Reduction in a Margin of Safety.

   The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. The TS allow operation of the plant without the full complement of equipment through the conditions for not meeting the TS LCO. The risk associated with this allowance is managed by the imposition of required actions that must be performed within the prescribed completion times. The net effect of being in a TS condition on the margin of safety is not considered significant. The proposed change does not alter the required actions or completion times of the TS. The proposed change allows TS conditions to be entered, and the associated required actions and completion times to be used in new circumstances. This use is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The change also eliminates current allowances for utilizing required actions and completion times in similar circumstances, without assessing and managing risk. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

   The NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

   Attorney for licensee: Ms. Lisa F. Vaughn , Legal Department (PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, North Carolina 28201-1006.

   NRC Section Chief: John A. Nakoski.

   Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

   Date of amendment request: August 19, 2003.

   Description of amendment request: The proposed amendment would modify Note 5 to Pilgrim Nuclear Power Station Technical Specification (TS) Table 3.2.C-1, to change the Rod Block Monitor (RBM) power- dependent Low Power Set Point (LPSP) allowable value from [le] 29% to

   [le] 25.9%. The proposed change would make the RBM LPSP consistent with plant procedures and the Core Operating Limits Report (COLR) allowable value used in compliance with TS 5.6.5.

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

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

   Response: No.

   The proposed Rod Block Monitor (RBM) power dependent Low Power Set Point (LPSP) of [le] 25.9% corrects the incorrect value of [le] 29% in Note 5 of TS Table 3.2.C-1 and is more restrictive than the incorrect value. The proposed set point allowable value of [le] 25.9% provides rod block protection over a wider range from [le] 25.9% to 100%, instead of [le] 29% to 100%, thereby enforcing RBM protection against rod withdrawal error at a lower power level. Also, the proposed requirement is consistent with the core operating limits report and is in accordance with License Amendment 138.

   The proposed RBM LPSP value ensures safe operation of the plant during startup and run modes. This requirement is not an accident precursor. The proposed analytical value [le] 25.9% was derived from the Average Power Range Monitor, Rod Block Monitor and Technical Specification (ARTS) improvement program methodology that was approved by License Amendment 138 and complies with the analytical methods required by Technical Specification 5.6.5. The proposed change provides additional assurance that the core operating limits are followed for safe operation and assumptions for core operating limits are met.

   Therefore, the probability or consequences of an accident previously evaluated is not significantly increased.

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

   Response: No.

   The proposed change does not involve a change to the plant design or a new mode of equipment operation and enforces previously evaluated conditions. As a result, the proposed changes do not affect parameters or conditions that could contribute to the initiation of any new of different kind of accident. Therefore, this proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   Response: No.

   The proposed change increases the margin of safety by providing additional assurance that the RBM downscale trip is not bypassed for reactor power [ge] 25.9% of rated thermal power and is based on previously evaluated methodologies. 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.

   Attorney for licensee: J. M. Fulton, Esquire, Assistant General Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, Massachusetts 02360-5599.

   NRC Section Chief: Darrell J. Roberts (Acting).

   Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

   Date of amendment request: December 8, 2003.

   Description of amendment request: The proposed amendment would delete a portion of the Pilgrim Nuclear Power Station (Pilgrim) Technical Specification (TS) 4.6.A.2, ``Primary System Boundary-- Thermal and Pressurization Limitations,'' and the associated TS Table 4.6-3, ``Reactor Vessel Material Surveillance Program Withdrawal Schedule.'' The amendment would replace the existing Reactor Vessel Material Surveillance Program with the Boiling Water Reactor Vessel and Internal Project (BWRVIP) Integrated Surveillance Program (ISP) and Supplemental Surveillance Program

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   (SSP). The BWRVIP ISP/SSP would be incorporated into the Pilgrim Updated Final Safety Analysis Report.

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

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

   Response: No.

   The proposed changes to the licensing basis continue to assure that applicable regulatory requirements are met and the same assurance of reactor pressure vessel integrity continues to be provided. The proposed changes to the TS[s] and licensing basis follow the [U.S. Nuclear Regulatory Commission] NRC Safety Evaluation approving the implementation of the ISP. The proposed changes ensure that the reactor pressure vessel will continue to be operated within the design, operational, and testing limits.

   The proposed changes do not modify the reactor coolant pressure boundary, (i.e., there are no changes in operating pressure, materials, or seismic loading). The proposed changes do not adversely affect the integrity of the reactor coolant pressure boundary such that its function in the control of radiological consequences is affected.

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

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

   Response: No.

   The proposed change does not involve a modification to the design of plant structures, systems, or components. Thus, no new modes of operation are introduced by the proposed change. The proposed change will not create any failure mode not bounded by previously evaluated accidents. 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 implementation of ISP has been previously approved by the NRC and found to provide an acceptable alternative to plant- specific reactor vessel material surveillance programs. Operation of Pilgrim within the program ensures that the reactor vessel materials will continue to behave in a non-brittle manner, thereby preserving the original safety design bases. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

   Attorney for licensee: J. M. Fulton, Esquire, Assistant General Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, Massachusetts 02360-5599.

   NRC Section Chief: Darrell J. Roberts, Acting.

   Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

   Date of amendment request: January 16, 2004.

   Description of amendment request: The proposed amendment would approve an engineering evaluation performed in accordance with Pilgrim Nuclear Power Station Technical Specification (TS) 3.6.D.3 to justify continued power operation with safety relief valve (SRV) -3A and SRV-3D discharge pipe temperatures exceeding 212 degrees Fahrenheit (F) for greater than 24 hours as required by TS 3.6.D.4.

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

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

   Response: No.

   Indication of elevated Safety Relief Valve (SRV) discharge pipes temperature is attributed to leakage past the SRVs. Excessive leakage, corresponding to temperatures greater than 255 F, has the potential to affect SRV operability by affecting the SRV setpoint or response time. Continued operation with the discharge pipes of the SRVs indicating temperatures less than 255 F ensures that the leakage past the SRVs is maintained below the threshold for a leakage rate that would potentially have an effect on SRV setpoint or response time.

   Administrative controls are in place to ensure that margin to the 255 F value is maintained to assure reliable operation and to reduce the potential for damage to the pilot seat and disc. The SRVs continue to perform their intended design/safety function with no adverse effect because the leakage past the SRVs is maintained below the threshold for a leakage rate that could potentially have an adverse impact on the ability of the SRVs to perform their design functions. The impact of the leakage on other systems is small and all systems continue to be able to perform their intended design functions. Current accident analyses remain bounding and there is no significant increase in the consequences of any accident previously evaluated. In addition, as a result of the leakage, normal plant operating parameters are not affected and consequently there is no increased risk in a plant transient.

   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.

   Continued plant operation with elevated discharge pipe temperatures for SRV-3A & 3D within the bounds of the established administrative controls ensures that the leakage past the SRVs is maintained below the threshold for a leakage rate that would potentially have an effect on SRV setpoint or response time. This ensures that the SRVs will perform their intended design/safety function. The leakage does not adversely impact the ability of any system to perform its design function. The methods governing plant operation and testing remain consistent with current safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   Response: No.

   Continued operation with the discharge pipes of SRV-3A & 3D indicating temperatures in excess of 212 F does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. The leakage does not result in excess SRV setpoint drift or response time changes. The imposed administrative controls on plant operation provide assurance that there will be no adverse effect on the ability of the SRVs to perform their intended design/safety function. There are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

   Attorney for licensee: J. M. Fulton, Esquire, Assistant General Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, Massachusetts 02360-5599.

   NRC Section Chief: Darrell J. Roberts, Acting.

   Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

   Date of amendment request: November 25, 2003.

   [[Page 7523]]

   Description of amendment request: Exelon Generation Company, LLC, the licensee, is proposing a change to the Limerick Generating Station (LGS), Units 1 and 2, Technical Specifications (TSs) contained in Appendix A to Operating Licenses NPF-39 and NPF-85, respectively. The proposed changes involve relocating the Reactor Coolant System (RCS) chemistry Limiting Conditions for Operation (LCO) from the Technical Specifications (TSs) to the Technical Requirements Manual (TRM). Additionally, proposed changes to TS RCS specific activity requirements involve removing various items and modifying the surveillance frequency of the isotopic analysis for Dose Equivalent I-131 from at least once per 31 days to once per 7 days.

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

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

   Response: No. The proposed relocation of the reactor coolant system chemistry requirements from Technical Specifications (TS) to the Technical Requirements Manual (TRM) is administrative in nature and does not involve the modification of any plant equipment or affect basic plant operation. Conductivity, chloride and pH limits are not assumed to be an initiator of any analyzed event, nor are these limits assumed in the mitigation of consequences of accidents.

   The proposed elimination from TS of the reactor coolant system specific activity requirements involving E-bar, gross beta, and gross gamma does not involve the modification of any plant equipment or affect basic plant operation. Specific activity is not assumed to be an accident initiator, and the specific activity requirements remaining in TS provide reasonable assurance that the reactor coolant specific activity is maintained at a sufficiently low level to preclude offsite doses from exceeding a small fraction of the limits of 10 CFR part 100 in the event of an accident.

   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 to relocate the reactor coolant system chemistry requirements from TS to the TRM, and to eliminate the reactor coolant system specific activity requirements involving E-bar, gross beta, and gross gamma, do not involve any physical alteration of plant equipment and do not change the method by which any safety-related system performs its function. As such, no new or different types of equipment will be installed, and the basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions.

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

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

   Response: No. The proposed change to the reactor coolant system chemistry requirements involves the relocation of current TS requirements to the TRM based on regulatory guidance and previously approved changes for other stations. The proposed change is administrative in nature, does not negate any existing requirement, and does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by requirements that are retained, but relocated from the TS to the TRM.

   The proposed change also involves the elimination from TS of the reactor coolant system specific activity requirements involving E- bar, gross beta, and gross gamma. The specific activity requirements remaining in TS provide reasonable assurance that the reactor coolant specific activity is maintained at a sufficiently low level to preclude offsite doses from exceeding a small fraction of the limits of 10 CFR Part 100 in the event of an accident. As a result, the proposed change does not adversely affect existing plant safety margins.

   Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

   The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 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: Mr. Edward Cullen, Vice President & General Counsel, Exelon Generation Company, LLC, 2301 Market Street, Philadelphia, PA 19101.

   NRC Section Chief: Darrell Roberts, Acting.

   Florida Power Corporation, et al., Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant, Citrus County, Florida

   Date of amendments request: December 19, 2003, as supplemented January 14, 2004.

   Description of amendment request: The proposed change allows entry into a mode or other specified condition in the applicability of a technical specification (TS), while in a condition statement and the associated required actions of the TS, provided the licensee performs a risk assessment and manages risk consistent with the program in place for complying with the requirements of Title 10 of the Code of Federal Regulations (10 CFR), part 50, Section 50.65(a)(4). Limiting Condition for Operation (LCO) 3.0.4 exceptions in individual TSs would be eliminated, and Surveillance Requirement (SR) 3.0.4 revised to reflect the LCO 3.0.4 allowance.

   This change was proposed by the industry's Technical Specification Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a notice of opportunity for comment in the Federal Register on August 2, 2002 (67 FR 50475), on possible amendments concerning TSTF-359, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability of the following NSHC determination in its application dated December 19, 2003, as supplemented by letter dated January 14, 2004.

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

   Criterion 1--The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

   The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. Being in a TS condition and the associated required actions is not an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on required actions as allowed by proposed LCO 3.0.4, are no different than the consequences of an accident while entering and relying on the required actions while starting in a condition of applicability of the TS. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or

   [[Page 7524]]

   consequences of an accident previously evaluated.

   Criterion 2--The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated.

   The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Entering into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

   Criterion 3--The Proposed Change Does Not Involve a Significant Reduction in a Margin of Safety.

   The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. The TS allow operation of the plant without the full complement of equipment through the conditions for not meeting the TS LCO. The risk associated with this allowance is managed by the imposition of required actions that must be performed within the prescribed completion times. The net effect of being in a TS condition on the margin of safety is not considered significant. The proposed change does not alter the required actions or completion times of the TS. The proposed change allows TS conditions to be entered, and the associated required actions and completion times to be used in new circumstances. This use is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The change also eliminates current allowances for utilizing required actions and completion times in similar circumstances, without assessing and managing risk. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

   The NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

   Attorney for licensee: Steven R. Carr, Associate General Counsel-- Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

   NRC Section Chief: Allen G. Howe.

   Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York

   Date of amendment request: January 9, 2004.

   Description of amendment request: The licensee proposed to revise the Technical Specifications (TSs) and the Updated Final Safety Analysis Report (UFSAR) by replacing the current plant-specific reactor pressure vessel (RPV) material surveillance program with the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP). Specifically, the proposed amendment would (1) delete the current reactor vessel material specimen surveillance schedule in Section 3/4.2.2, ``Minimum Reactor Vessel Temperature for Pressurization;'' (2) delete the special reporting requirement regarding material surveillance specimen examination in Section 6.6.6.a; and (3) approve changes in the UFSAR to reflect the licensee's participation in the ISP and use of a methodology for determining neutron fluences.

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

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

   Response: No.

   The proposed changes implement an ISP that has been evaluated by the NRC as meeting the requirements of paragraph III.C of Appendix H to 10 CFR [Part] 50; remove a TS surveillance requirement that prescribes a plant-specific withdrawal schedule for RPV surveillance specimens; and delete an unnecessary reporting requirement relating to RPV surveillance specimen examination. The proposed changes provide the same assurance of RPV integrity as has always been provided. Implementation of an ISP is not a precursor or initiator of any accident previously evaluated. No physical changes to the plant will result from the proposed changes. The proposed changes will not cause the RPV or interfacing systems to be operated outside of any design or testing limits, and will not alter any assumptions or initial conditions previously used in evaluating the radiological consequences of accidents.

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

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

   Response: No.

   The proposed changes revise the NMP1 licensing bases to reflect participation in the BWRVIP ISP. The ISP was approved by the NRC staff as an acceptable material surveillance program that complies with 10 CFR [Part] 50, Appendix H. No physical changes to the plant will result from the proposed changes. The proposed changes do not affect the design or operation of any system, structure, or component. As an alternate monitoring program, the ISP cannot create a new failure mode involving the possibility of a new or different kind of accident.

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

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

   Response: No.

   The proposed changes have no impact on the margin of safety of any TS. There is no impact on safety limits or limiting safety system settings. The changes do not affect any plant safety parameters or setpoints. No physical or operational changes to the plant will result from the proposed changes.

   The RPV material surveillance program requirements contained in 10 CFR [Part] 50, Appendix H provide assurance that adequate margins of safety exist during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the reactor coolant pressure boundary may be subjected over its service lifetime. The BWRVIP ISP has been approved by the NRC staff as an acceptable material surveillance program that complies with 10 CFR [Part] 50, Appendix H. The ISP will provide the material surveillance data that will assure that the safety margins required by the NRC regulations are maintained.

   Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

   The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 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 20005-3502.

   NRC Section Chief: Richard J. Laufer.

   Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station Unit No. 2 (NMP2), Oswego County, New York

   Date of amendment request: January 9, 2004.

   Description of amendment request: The licensee proposed to revise the licensing basis documented in the Updated Safety Analysis Report (USAR) by replacing the current plant-specific reactor pressure vessel (RPV) material surveillance program with the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP). Specifically, the proposed amendment would approve revising the USAR to reflect the licensee's participation in the ISP and

   [[Page 7525]]

   use of a methodology for determining neutron fluences.

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

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

   Response: No.

   The proposed change implements an ISP that has been evaluated by the NRC [Nuclear Regulatory Commission] as meeting the requirements of paragraph III.C of Appendix H to 10 CFR [Part] 50. The proposed change provides the same assurance of RPV integrity as has always been provided. Implementation of an ISP is not a precursor or initiator of any accident previously evaluated. No physical changes to the plant will result from the proposed change. The proposed change will not cause the RPV or interfacing systems to be operated outside of any design or testing limits, and will not alter any assumptions or initial conditions previously used in evaluating the radiological consequences of accidents.

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

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

   Response: No.

   The proposed change revises the NMP2 licensing bases to reflect participation in the BWRVIP ISP. The ISP was approved by the NRC staff as an acceptable material surveillance program that complies with 10 CFR [Part] 50, Appendix H. No physical changes to the plant will result from the proposed change. The proposed change does not affect the design or operation of any system, structure, or component. As an alternate monitoring program, the ISP cannot create a new failure mode involving the possibility of a new or different kind of accident.

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

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

   Response: No.

   The proposed change has no impact on the margin of safety of any TS [Technical Specification]. There is no impact on safety limits or limiting safety system settings. The change does not affect any plant safety parameters or setpoints. No physical or operational changes to the plant will result from the proposed change.

   The RPV material surveillance program requirements contained in 10 CFR [Part] 50, Appendix H provide assurance that adequate margins of safety exist during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the reactor coolant pressure boundary may be subjected over its service lifetime. The BWRVIP ISP has been approved by the NRC staff as an acceptable material surveillance program that complies with 10 CFR [Part] 50, Appendix H. The ISP will provide the material surveillance data that will assure that the safety margins required by the NRC regulations are 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.

   Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

   NRC Section Chief: Richard J. Laufer.

   Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear Power Plant, Kewaunee County, Wisconsin

   Date of amendment request: January 16, 2004.

   Description of amendment request: The proposed amendment is to revise Technical Specifications (TS) for the Kewaunee Nuclear Power Plant (KNPP). The proposed change would revise (1) the containment closure TS to allow the equipment hatch to be open during refueling operations and/or during movement of irradiated fuel assemblies within containment, (2) the containment tests TS to require verification of the ability to close the equipment hatch periodically during refueling operations, and (3) the control room post-accident recirculation system TS to include requirements for operability during fuel handling operations in which the fuel that is being moved has been irradiated less than 30 days ago.

   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. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

   Response: No.

   The proposed change would allow the containment equipment hatch to remain open during irradiated fuel movement in containment. This penetration is not an initiator of any accident. The probability of a fuel handling accident (FHA) in the containment is unaffected by the position of the equipment hatch. Adoption of this change requires analyses, approved by the [Nuclear Regulatory Commission] NRC staff, demonstrating that the dose consequences of a FHA with the equipment hatch open are acceptable. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

   Response: No.

   The proposed change does not involve the addition or modification of any plant equipment. Also, the proposed change would not alter the design, configuration, or method of operation of the plant beyond the standard functional capabilities of the equipment. The proposed change involves a change to the Technical Specifications (TS) that would allow the equipment hatch to remain open during irradiated fuel movement within the containment. Having the equipment hatch open does not create the possibility of a new accident. 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.

   Analysis demonstrates that the resultant doses associated with a fuel handling accident are well within the appropriate acceptance limits. This change removes a defense-in-depth barrier that the analysis did not credit but provides additional restrictions on fission product release. Thus, this proposed change has the potential for an increased dose at the site boundary due to a FHA; however, the analysis demonstrates that the resultant doses are well within the appropriate acceptance limits. Without the containment structure, analysis demonstrates that the dose consequences are still approximately 20% of the allowable value for the control room dose and less than 2% of the allowable value for offsite dose. Thus, the margin of safety has not been significantly reduced. Administrative provisions that facilitate closing the equipment hatch following an evacuation of the containment further reduces the offsite doses in the event of a FHA and provides additional margin to the calculated offsite doses. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

   Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, P.O. Box 1497, Madison, WI 53701-1497.

   NRC Section Chief: L. Raghavan.

   [[Page 7526]]

   Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska

   Date of amendment request: December 1, 2003.

   Description of amendment request: The proposed amendment will modify Fort Calhoun Station Technical Specification (TS) 2.7, ``Electrical Systems,'' TS Table 3-5, ``Minimum Frequencies for Equipment Tests,'' and TS 5.0, ``Administrative Controls.'' This proposed amendment modifies the requirements for diesel generator (DG) fuel oil for consistency with the Improved Standard Technical Specifications (ISTS) and adds requirements for DG lubricating oil and DG starting air. The proposed changes will assure that the required quality and quantity of DG fuel oil is maintained and also will assure that sufficient DG lubricating oil and DG starting air is maintained.

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

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change will revise Technical Specification (TS) 2.7, ``Electrical Systems,'' TS Table 3-5, ``Minimum Frequencies for Equipment Tests,'' and TS 5.0, ``Administrative Controls.'' This proposed amendment modifies the requirements for Diesel Generator (DG) Fuel Oil for consistency with the Improved Standard Technical Specifications (ISTS) and adds requirements for DG Lubricating Oil, and DG Starting Air. The Surveillance interval of Diesel Fuel Supply Surveillance [Table 3-5, Item 9 (changed to 9a)] is being changed from daily to monthly. The 31 day Surveillance interval is adequate to ensure that a sufficient supply of fuel oil is available, since low level alarms are provided and unit operators would be aware of any large uses of fuel oil during this period. Therefore, this change does not significantly increase the probability of a previously analyzed accident. Further, an increase of the Surveillance interval will not affect the capability of the component or system to perform its function. Therefore, this change does not significantly increase the consequences of a previously analyzed accident. All other changes are more restrictive changes. The changes will ensure that proper Limiting Conditions for Operation are entered for equipment or functional inoperability. There are no physical alterations being made to the DGs or related systems.

   With regards to TSTF-254, Rev. 2, the proposed change does not require any physical change to any plant systems, structures, or components nor does it require any change in systems or plant operations. The proposed change does not require any change in safety analysis methods or results. The water content of the DG fuel oil system is not considered an accident initiator. The change to reduce the fuel oil sampling frequency for water content from 31 days to 92 days does not present a significant impact to DG operability or significantly degrade DG performance and, therefore, does not present a significant detrimental impact on structures, systems, or components that support accident recovery.

   With regards to TSTF-374, Rev. 0, the proposed changes relocate the specific ASTM Standard references from the Administrative Controls Section of TS to a licensee-controlled document. Since any change to the licensee-controlled document will be evaluated pursuant to the requirements of 10 CFR 50.59, ``Changes, tests and experiments,'' no increase in the probability or consequences of an accident previously evaluated is involved. In addition, the ``clear and bright'' test used to establish the acceptability of new fuel oil for use prior to addition to storage tanks has been expanded to allow a water and sediment content test to be performed to establish the acceptability of new fuel oil. The proposed changes revise Bases for TS 3.2 to reference the current specific ASTM Standards. The Bases for TS 3.2 are revised to indicate that the API gravity is tested in accordance with ASTM D287.

   Relocating the specific ASTM Standard references from the TS to a licensee-controlled document, allowing a water and sediment content test to be performed to establish the acceptability of new fuel oil, and revising the TS Bases will not affect nor degrade the ability of the DGs to perform their specified safety function. Fuel oil quality will continue to meet ASTM requirements.

   The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures.

   Therefore, the proposed changes do 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 changes will not result in any physical alterations to the DGs, any plant configuration, systems, equipment, or operational characteristics. There will be no changes in operating modes, or safety limits, or instrument limits. With the proposed changes in place, Technical Specifications will retain requirements for the DGs.

   With regards to TSTF-254, Rev. 2, the accident analyses do not consider the water content of the EDG fuel oil systems. Failure of a DG to start and load upon accident initiation is considered in the accident analyses, but is not affected by the proposed change to the fuel oil sampling Surveillance intervals. The existing analyses remain unchanged and the proposed TS change does not affect any accident initiators that would create a new accident.

   With regards to TSTF-374, Rev. 0, the proposed changes relocate the specific ASTM Standard references from the Administrative Controls Section of the TS to a licensee-controlled document. In addition, the ``clear and bright'' test used to establish the acceptability of new fuel oil for use prior to addition to storage tanks has been expanded to allow a water and sediment content test to be performed to establish the acceptability of new fuel oil. The proposed changes [ ] also revise the Bases of TS 3.2 to reference the current specific ASTM Standards. The Bases for TS 3.2 is revised to indicate that the API gravity is tested in accordance with ASTM D287.

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

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

   The proposed changes clarify the regulatory requirements for the DGs. The Completion Times and Frequencies established are within those invoked by the present Technical Specifications or equal to those previously reviewed and approved for use by the NRC. The proposed changes will not alter any physical or operational characteristics of the DGs and associated systems and equipment.

   With regards to TSTF-254, Rev. 2, the proposed change does not require any change in accident analysis methods or results. The safety margin as established in the current license basis remains unchanged. Reducing the Surveillance interval for DG fuel oil sampling does not, in itself, result in a measurable impact on the operability of the DGs. The water content of the DG fuel oil systems will continue to be assessed and corrective action taken should any condition adverse to DG operability be detected.

   With regards to TSTF-374, Rev. 0, [t]he proposed changes relocate the specific ASTM Standard references from the Administrative Controls Section of [the] TS to a licensee-controlled document. Instituting the proposed changes will continue to ensure the use of current applicable ASTM Standards to evaluate the quality of both new and stored fuel oil designated for use in the emergency DGs. The detail associated with the specific ASTM Standard references is not required to be in the TS to provide adequate protection of the public health and safety, since the TS still retain the requirement for compliance with the applicable ASTM Standard. Changes

   [[Page 7527]]

   to the licensee-controlled document are performed in accordance with the provisions of 10 CFR 50.59. Should it be determined that future changes involve a potential reduction in a margin of safety, NRC review and approval would be necessary prior to implementation of the changes. This approach provides an effective level of regulatory control and provides for a more appropriate change control process.

   The ``clear and bright'' test used to establish the acceptability of new fuel oil for use prior to addition to storage tanks has been expanded to allow a water and sediment content test to be performed to establish the acceptability of new fuel oil. The proposed changes revise the Bases for TS 3.2 to reference the current specific ASTM Standards. The Bases for TS 3.2 is revised to indicate that the API gravity is tested in accordance with ASTM D287. The level of safety of facility operation is unaffected by the proposed changes since there is no change in the intent of the TS requirements of assuring fuel oil is of the appropriate quality for emergency DG use. The proposed changes provide the flexibility needed to maintain state-of-the-art technology in fuel oil sampling and analysis methodology.

   Therefore, the proposed changes do not involve 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: James R. Curtiss, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

   NRC Section Chief: Stephen Dembek.

   PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

   Date of amendment request: December 12, 2003.

   Description of amendment request: Proposed changes to the Technical Specifications (TSs) of the Control Room Emergency Filtration System (CREPS) would no longer require it to be OPERABLE in COLD SHUTDOWN. However, CREPS would have to be operable during operations with potential for draining the reactor vessel. The TSs for the Control Room Ventilation Radiation Monitor would be revised so that OPERABILITY would no longer be required during refueling. However, OPERABILITY would be required for operations with potential for draining the reactor vessel.

   Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the change involve a significant increase in the probability or consequences of an accident previously analyzed?

   Response: No.

   The proposed changes to Table 3.3.7.1-1, Radiation Monitoring Instrumentation, and Table 4.3.7.1-1, Radiation Monitoring Instrumentation Surveillance Requirements, adds ``recently'' to modify irradiated fuel in the ``*'' footnote to provide consistency with TSTF-51, Rev. 2. Proposed changes to eliminate Operational Condition 5 from Tables 3.3.7.1-1 and 4.3.7.1-1, Control Room Ventilation Radiation Monitor, Operational Condition 4 from Control Room Emergency Filtration (CREF) System and adding operations with the potential for draining the reactor vessel (OPDRV) to Tables 3.3.7.1-1 and 4.3.7.1-1 footnote ``*'' and the CREF System are consistent with NUREG-1433 Vol. 1, Rev. 2, Standard Technical Specifications, General Electric Plants.

   The proposed changes associated with the fuel handling accident (FHA) do not involve a change to structures, components, or systems that would affect the probability of an accident previously evaluated in the Hope Creek Updated Final Safety Analysis Report (UFSAR). The FHA for the Hope Creek Generating Station (HCGS) is defined as a drop of a fuel assembly over irradiated assemblies in the reactor core 24 hours after reactor shutdown. Alternative Source Term (AST) is used to evaluate the dose consequences of a postulated accident. The FHA has been analyzed without credit for Secondary Containment, Filtration Recirculation and Ventilation System (FRVS), and CREF system. The resultant radiological consequences are within the acceptance criteria set forth in 10 CFR 50.67 and Regulatory Guide (RG) 1.183. This amendment does not alter the methodology or equipment used in fuel handling operations. The equipment hatch, personnel air locks, other containment penetrations, or any component thereof is not an accident initiator. Actual fuel handling operations are not affected by the proposed changes.

   Consequently the probability of a previously analyzed FHA is not affected by the proposed amendment. No other accident initiator is affected by the proposed changes.

   Therefore, this proposed amendment does not involve a significant increase in the probability of occurrence or radiological consequences of an accident previously analyzed.

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

   Response: No.

   The proposed changes to Table 3.3.7.1-1, Radiation Monitoring Instrumentation, and Table 4.3.7.1-1, Radiation Monitoring Instrumentation Surveillance Requirements, adds recently'' to modify irradiated fuel in the ``*'' footnote provides consistency with TSTF-51, Rev. 2. Proposed changes to eliminate Operational Condition 5 from Tables 3.3.7.1-1 and 4.3.7.1-1, Control Room Ventilation Radiation Monitor, Operational Condition 4 from CREF System and adding OPDRV to Table 3.3.7.1-1 and 4.3.7.1-1 footnote ``*'' and the CREF System are consistent with NUREG-1433 Vol. 1, Rev. 2, Standard Technical Specifications, General Electric Plants.

   The proposed amendment will not create the possibility for a new or different type of accident from any accident previously evaluated because 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. 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. The systems affected by the changes are used to mitigate the consequences of an accident that has already occurred. The proposed TS changes do not significantly affect the mitigative function of these systems.

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

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

   Response: No.

   The proposed changes to Table 3.3.7.1-1, Radiation Monitoring Instrumentation, and Table 4.3.7.1-1, Radiation Monitoring Instrumentation Surveillance Requirements, adds ``recently'' to modify irradiated fuel in the ``*'' footnote provides consistency with TSTF-51, Rev. 2. Proposed changes to eliminate Operational Condition 5 from Tables 3.3.7.1-1 and 4.3.7.1-1 Control Room Ventilation Radiation Monitor, Operational Condition 4 from CREF System and adding OPDRV to Table 3.3.7.1-1 and 4.3.7.1-1 footnote ``*'' and the CREF System are consistent with NUREG-1433 Vol. 1, Rev. 2, Standard Technical Specifications, General Electric Plants.

   The proposed changes revise the TS to establish operational conditions where specific activities represent situations during which significant radioactive releases can be postulated. These operational conditions are consistent with the design basis analysis and are established such that the radiological consequences remain at or below the regulatory guidelines. Safety margins and analytical conservatisms are retained to ensure that the analysis adequately bounds all postulated event scenarios. The proposed TS continue[s] to ensure that the total effective dose equivalent (TEDE) for the control room (CR), the exclusion area boundary (EAB), and low population zone (LPZ) boundaries are below the corresponding acceptance criteria specified in I0 CFR 50.67 and RG 1.183.

   Therefore, these changes do not involve a significant reduction in margin of safety.

   The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three

   [[Page 7528]]

   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: Jeffrie J. Keenan, Esquire, Nuclear Business Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

   NRC Section Chief: Darrell Roberts, Acting.

   PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

   Date of amendment request: December 24, 2003.

   Description of amendment request: The amendment request changes the Technical Specifications (TSs) to allow the use of GE14 fuel in reload cycle 13. Specifically, the proposed changes modify the TSs to reflect the use of General Electric (GE) core reload analysis methodologies. The proposed changes would revise the limiting conditions for operation for the recirculation loops to modify and add action statements to provide further thermal limit control during single-loop operation to be consistent with GE methodology specified in the core operating limits report. The proposed changes also modify the TS definitions and TS requirements for average planar linear heat generation rate consistent with NUREG-1433, ``Standard Technical Specifications (STS) General Electric Plants, BWR/4,'' Revision 2. Additionally, TS Section 6.9.1.9 would be revised to correct an error in a previous amendment that inadvertently removed a reference. The NRC-approved reference would be restored to TS 6.9.1.9 in the format prescribed in NUREG-1433, Revision 2.

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

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

   Response: No.

   The revised information and references relative to the fuel vendor's calculation methodologies throughout the Technical Specifications are considered to be administrative in nature because they reflect the NRC approved methodologies to be used by PSEG Nuclear LLC and the fuel vendor to develop operating and safety limits for the fuel and core designs. The changes to the Recirculation System Action statements ensure the appropriate adjustments are made to core operating limits for single loop operation, and the Core Operating Limits Report (COLR) will still be developed in accordance with NRC approved methods. These proposed changes do not alter the method of operating the plant and have no effect on the probability of an accident initiating event or transient.

   There are no significant increases in the radiological consequences of an accident previously evaluated. The basis of the COLR and the PSEG Nuclear LLC and fuel vendor calculation methodologies is to ensure that no mechanistic fuel damage is calculated to occur if the limits on plant operation are not violated. The COLR will continue to preserve the existing margin to fuel damage and the probability of fuel damage is not increased.

   Therefore, the proposed change does not involve an increase in the probability or radiological 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.

   These changes do not involve any new method for operating the facility, any changes to setpoints, or any new facility modifications for the reload core operation. No new initiating events or transients result from these changes.

   The revised information and references relative to the fuel vendor's calculation methodologies throughout the Technical Specifications are considered to be administrative in nature because they reflect the NRC approved methodologies to be used by PSEG Nuclear LLC and the fuel vendor to develop operating and safety limits for the fuel and core designs. The changes to the Recirculation System Action statements ensure the appropriate adjustments are made to core operating limits for single loop operation, and the COLR will still be developed in accordance with NRC-approved methods.

   Therefore, the proposed Technical Specification changes do 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 revised information and references relative to the fuel vendor's calculation methodologies throughout the Technical Specifications are considered to be administrative in nature because they reflect the NRC approved methodologies to be used by PSEG Nuclear LLC and the fuel vendor to develop operating and safety limits for the fuel and core designs. The changes to the Recirculation System Action statements ensure the appropriate adjustments are made to core operating limits for single loop operation, and the COLR will still be developed in accordance with NRC approved methods. The proposed changes will continue to ensure that the plant is operated within specified acceptable fuel design limits. Therefore, the proposed Technical Specifications changes do not involve a significant reduction in a margin of safety.

   The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 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: Jeffrie J. Keenan, Esquire, Nuclear Business Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

   NRC Section Chief: Darrell Roberts, Acting.

   Notice of Issuance of Amendments to Facility Operating Licenses

   During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

   Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for A Hearing in connection with these actions was published in the Federal Register as indicated.

   Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated.

   For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic

   [[Page 7529]]

   Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there

   are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301- 415-4737 or by email to pdr@nrc.gov.

   Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi

   Date of application for amendment: May 12, 2003, as supplemented by letter dated October 29, 2003.

   Brief description of amendment: The amendment changes administrative Technical Specification (TS) 5.5.12 regarding containment integrated leakage rate testing (ILRT) and TS 3.6.5.1.1 regarding drywell bypass leak rate testing (DWBT). The change would allow for a one-time extension of the interval from 10 to 15 years for performance of the next ILRT and DWBT.

   Date of issuance: January 28, 2004.

   Effective date: As of the date of issuance and shall be implemented within 60 days of issuance.

   Amendment No: 164.

   Facility Operating License No. NPF-29: The amendment revises the Technical Specifications.

   Date of initial notice in Federal Register: June 10, 2003 (68 FR 34666).

   The October 29, 2003, supplemental letter provided clarifying information that did not change the scope of the original Federal Register notice or the original no significant hazards consideration determination.

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 28, 2004.

   No significant hazards consideration comments received: No.

   PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

   Date of application for amendment: March 13, 2003.

   Brief description of amendment: The amendment deletes Technical Specification (TS) 6.8.4.c, ``Post Accident Sampling,'' and thereby eliminates the requirements to have and maintain the post accident sampling system at the Hope Creek Generating Station.

   Date of issuance: January 29, 2004.

   Effective date: As of the date of issuance and shall be implemented within 180 days.

   Amendment No.: 149.

   Facility Operating License No. NPF-57: This amendment revised the TSs.

   Date of initial notice in Federal Register: May 27, 2003 (68 FR 28856).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 29, 2004.

   No significant hazards consideration comments received: No.

   PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

   Date of application for amendment: June 17, 2003.

   Brief description of amendment: The amendment corrects typographical errors in the Technical Specification (TS) Index and deletes TS 4.6.2.1.b.2.b, verification that thermal power is less than or equal to 1% of rated thermal power at least once per hour when the suppression chamber temperature exceeds 95 F. The proposed TS change is consistent with the standard TSs for General Electric Plants, Boiling-Water Reactor/4 (NUREG-1433, Revision 2).

   Date of issuance: January 30, 2004.

   Effective date: As of the date of issuance, to be implemented within 60 days.

   Amendment No.: 150.

   Facility Operating License No. NPF-57: This amendment revised the TSs.

   Date of initial notice in Federal Register: July 8, 2003 (68 FR 40717).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 30, 2004.

   No significant hazards consideration comments received: No.

   PSEG Nuclear, LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

   Date of application for amendments: June 6, 2003.

   Brief description of amendments: The amendments modify the Salem Nuclear Generating Station, Unit Nos. 1 and 2, Technical Specifications (TSs) by: (1) Adding a footnote to TS 3/4.11.2.5 to clarify the applicability of the Limiting Condition for Operation while the system is removed from service for maintenance; (2) revising Surveillance Requirement 4.11.2.5 to delete the reference to hydrogen concentration; and (3) revising the corresponding TS Bases.

   Date of issuance: January 29, 2004.

   Effective date: As of the date of issuance, and shall be implemented within 60 days.

   Amendment Nos.: 261 and 243.

   Facility Operating License Nos. DPR-70 and DPR-75: The amendments revised the TSs.

   Date of initial notice in Federal Register: August 5, 2003 (68 FR 46246).

   The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 29, 2004.

   No significant hazards consideration comments received: No.

   TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

   Date of amendment request: July 18, 2003.

   Brief description of amendments: The amendments modified Technical Specification (TS) requirements for mode change limitations to adopt Industry/TS Task Force (TSTF) change TSTF-359, ``Increase Flexibility in Mode Restraints.''

   Date of issuance: January 23, 2004.

   Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance.

   Amendment Nos.: 109 and 109.

   Facility Operating License Nos. NPF-87 and NPF-89: The amendments revised the Technical Specifications.

   Date of initial notice in Federal Register: October 14, 2003 (68 FR 59222).

   The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 23, 2004.

   No significant hazards consideration comments received: No.

   Dated at Rockville, Maryland, this 9th day of February 2004.

   For the Nuclear Regulatory Commission. Ledyard B. Marsh, Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.

   [FR Doc. 04-3180 Filed 2-13-04; 8:45 am]

   BILLING CODE 7590-01-P