Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 35735-35743 [E5-3138]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 118 (Tuesday, June 21, 2005)]
[Notices]
[Pages 35735-35743]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E5-3138]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from May 26, 2005, to June 9, 2005. The last 
biweekly notice was published on June 7, 2005 (70 FR 33210).

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, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or 
petition for leave to intervene is filed within 60 days, the Commission 
or a presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted

[[Page 35736]]

with particular reference to the following general requirements: (1) 
The name, address, and telephone number of the requestor or petitioner; 
(2) the nature of the requestor's/petitioner's right under the Act to 
be made a party to the proceeding; (3) the nature and extent of the 
requestor's/petitioner's property, financial, or other interest in the 
proceeding; and (4) the possible effect of any decision or order which 
may be entered in the proceeding on the requestor's/petitioner's 
interest. The petition must also set forth the specific contentions 
which the petitioner/requestor seeks to have litigated at the 
proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer 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 ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
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 PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York

    Date of amendment request: May 25, 2005.
    Description of amendment request: The amendment would revise 
Technical Specification Section 3.4.9, ``Pressurizer,'' to revise the 
pressurizer water level limit during operation in Mode 3 (hot standby).
    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.
    Pressurizer water level is an assumed initial condition for 
certain accident analyses. Plant initial conditions are not accident 
initiators and do not have an effect on the probability of the 
accident occurring. The proposed change only revises the specified 
limit on water level in the pressurizer, so this change does not 
affect accident probability.
    Pressurizer water level is an assumed initial condition for 
accidents such as LOCA [loss-of-coolant accident], loss-of-load and 
loss-of-normal feedwater. The limiting accident analysis results 
occur at full power conditions when the available core thermal power 
is maximized. The proposed change does not affect the specified 
pressurizer level limit at any power level from zero to full power. 
That is, the pressurizer level limit is not being changed in Modes 1 
and 2. The proposed change does revise the specified pressurizer 
water level limit in Mode 3 (Hot Standby) but this does not affect 
accident analysis results because the limiting analyses will remain 
those that are postulated to occur in Mode 1 with the plant at full 
power.
    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 does not involve physical changes to 
existing plant equipment or the installation of any new equipment. 
The design of the pressurizer, the pressurizer level control system 
and the pressurizer safety valves is not being changed and the 
ability of these systems, structures, and components to perform 
their design or safety functions is not being affected. The proposed 
change revises the specified limit on pressurizer water level in 
Mode 3 (Hot Standby) to allow operators greater flexibility in 
performing a plant cooldown. The method used in performing the plant 
cooldown is not

[[Page 35737]]

being changed. This proposed change does not create new failure 
modes or malfunctions of plant equipment nor is there a new credible 
failure mechanism.
    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.
    Pressurizer level is an initial condition assumed in certain 
accident analyses involving an insurge in the pressurizer and an 
increasing reactor coolant system (RCS) pressure. These analyses 
demonstrate that the design pressure for the RCS is not exceeded for 
the limiting analyses based on the plant at full power. The proposed 
change does not affect the existing Technical Specification 
requirement for Mode 1 (Power Operation) or Mode 2 (Plant Startup) 
and therefore does not affect the assumptions or results of these 
accident analyses. The margin for RCS design pressure demonstrated 
by these analysis results is not being reduced. The proposed change 
only applies to the pressurizer level limit in Mode 3 (Hot Standby) 
when there is substantially lower thermal energy available to cause 
rapid expansion of reactor coolant and an insurge to the 
pressurizer. Protection of the RCS pressure boundary is still 
maintained by the pressurizer safety valves, which are not being 
modified by the proposed change in pressurizer water level.
    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: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: March 15, 2005.
    Description of amendment request: A change is proposed to revise 
the Waterford Steam Electric Station Unit 3 (Waterford 3) Technical 
Specification (TS) Section 4.4.4.4 to modify the steam generator tube 
inspection Acceptance Criteria for the ``Plugging or Repair Limit'' and 
the ``Tube Inspection,'' as contained in the Waterford 3 TS 
Surveillance Requirements (SR) 4.4.4.4.a.7 and 4.4.4.4.a.9, 
respectively. The purpose of these changes is to define the depth of 
the required tube inspections and to clarify the plugging criteria 
within the tubesheet region.
    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.
    Conducting the rotating Plus Point probe inspections to a 
minimum tubesheet length of 10.4 inches maintains the existing 
design limits and does not increase the probability or consequences 
of an accident involving tube burst or primary to secondary 
accident-induced leakage, as previously analyzed in the Waterford 3 
Final Safety Analysis Report. Also the NEI [Nuclear Energy 
Institute] 97-06 structural integrity and accident induced leakage 
of the steam generator tubes performance criteria will continue to 
be satisfied.
    Tube burst is precluded for a tube with defects within the 
tubesheet region because of the constraint provided by the 
tubesheet. As such, tube pullout resulting from the axial forces 
induced by primary to secondary differential pressures would be a 
prerequisite for tube burst to occur. Any degradation below C* is 
shown by empirical test results and analyses to be acceptable, 
thereby precluding an event with consequences similar to a 
postulated tube rupture event. WCAP-16208-P has shown that tube 
flaws below the C* length will not result in primary to secondary 
leakage greater than 0.1 gpm [gallons per minute] per steam 
generator. Inspection to the C* length will ensure that the 
postulated accident induced leakage for events that involve a 
faulted steam generator (e.g., a main steam line break (MSLB)) will 
remain within both the current and proposed extended power uprate 
(EPU) accident analyses of 720 gpd (0.5 gpm) and 540 gpd (0.375 
gpm), respectively.
    Therefore, the proposed change does not affect the probability 
or consequences of any Waterford 3 analyzed accidents.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Steam generator tube leakage and structural integrity will be 
maintained during all plant conditions upon implementation of the 
proposed inspection scope and plugging or repair limit changes to 
the Waterford 3 Technical Specifications. These changes do not 
introduce any new mechanisms that might result in a different kind 
of accident from those previously evaluated. Even with the limiting 
circumstances of a complete circumferential separation (360o through 
wall crack) of all of the tubes below the C* length, tube pullout is 
precluded and leakage is predicted to be maintained within both the 
current and proposed extended power uprate (EPU) accident analyses 
assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed inspection and plugging criteria will better assure 
that steam generator tube performance is maintained within its 
design basis and within the safety analysis assumptions. Operation 
with potential tube degradation below the C* inspection length 
within the tubesheet region of the steam generator tubing meets the 
intent of the inspection guidance of RG 1.83, Inservice Inspection 
of Pressurized Water Reactor Steam Generator Tubes, the requirements 
of General Design Criteria 14, 30 and 32 of 10 CFR 50, and the 
recommendations of NEI-97-06, Steam Generator Program Guidelines. 
The total leakage from an undetected flaw population below the C* 
inspection length under postulated accident conditions is accounted 
for to assure that the leakage criterion is met and bounded by both 
the current and the proposed EPU accident analyses assumptions. 
Adequate margin remains for other possible steam generator tube leak 
sources.
    The proposed changes also maintain the structural and accident-
induced leakage integrity of the steam generator tubes as required 
by NEI 97-06 and the plant design basis.
    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: N. S. Reynolds, Esquire, Winston & Strawn 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: David Terao.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, 
Beaver Valley Power Station, Unit No. 1 (BVPS-1), Beaver County, 
Pennsylvania

    Date of amendment request: April 11, 2005.
    Description of amendment request: The proposed amendment would 
revise the BVPS-1 Technical Specifications (TSs) to permit operation 
with replacement Model 54F steam generators (SGs) installed. These 
include changes to reactor core safety limits, reactor trip system and 
engineered safety features actuation system setpoints, and other safety 
analysis inputs related to the proposed new model 54F steam generators 
as well as changes to steam generator limiting conditions for operation 
and surveillance requirements. These proposed TS changes were 
originally submitted as part of the licensee's extended power uprate 
application,

[[Page 35738]]

dated October 4, 2004, however, delays in the review of that 
application have required the licensee to separately request these 
proposed TS changes in order to support SG replacement during and 
startup from the BVPS-1 2006 refueling outage.
    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 will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The safety and radiological dose consequence analyses confirmed 
that safety analysis and dose consequence analysis acceptance 
criteria will be satisfied for the Model 54F BVPS Unit No. 1 
replacement steam generators, including changes to reactor core 
safety limits, reactor trip system (RTS) and engineered safety 
features actuation system (ESFAS) setpoints, and other safety 
analysis inputs related to the proposed changes. The analyses are 
conservative and bounding with respect to operation with RSGs 
[replacement steam generators] at the current licensed maximum power 
level.
    For the purpose of this evaluation, the proposed changes to 
Technical Specifications 3.4.1.3, Reactor Coolant system Shutdown, 
and 3.4.5, Steam Generators, which will directly address the new 
Unit No. 1 replacement steam generators (RSG) can be grouped in the 
following areas:
    (a) The first area of change is to remove the references to 
repair of tubes by sleeving since they are not applicable to the RSG 
tubes.
    The accidents of interest are [steam generator] tube rupture and 
steam line break. A reduction in tube integrity could increase the 
possibility of a tube rupture accident and could increase the 
consequences of a steam line break. The tubing in the RSGs is 
designed and evaluated consistent with the margins of safety 
specified in the ASME Code [American Society of Mechanical 
Engineers, Boiler and Pressure Vessel Code], Section III. The 
program for periodic inservice inspection provides sufficient time 
to take proper and timely corrective action if tube degradation is 
present. The basis for the 40% through wall plugging limit is 
applicable to the RSGs just as it was to the original steam 
generators (OSG). An analysis has been performed consistent with the 
guidance in Draft Regulatory Guide 1.121 to justify the 
applicability of the 40% through wall plugging limit. As a result, 
there is no reduction in tube integrity for the RSGs.
    Elimination of the repair option and the associated references 
to repair of the OSG tubes is an administrative adjustment since the 
sleeve design is not applicable to the RSGs. The elimination of the 
repair option does not alter the requirements for inservice 
inspection or reduce the plugging limit for the RSG tubes.
    (b) The second area of change is to remove the references to 
voltage-based repair criteria on tube-to-tube support plate 
intersections since they are not applicable to the RSG tubes.
    Elimination of the repair option and the associated repair of 
the OSG tubes is an administrative adjustment since the voltage 
based repair criteria is not applicable to the RSGs. The elimination 
of the repair option does not alter the requirements for inservice 
inspection or reduce the plugging limit for the RSG tubes.
    (c) The third area of change is to update the wording and 
content of the TS to provide clarification and to incorporate 
wording enhancements consistent with the updates made to the subject 
TS for several other plants that have replaced steam generators. 
Since the RSGs will be subjected to a preservice inspection prior to 
installation, there is no need to perform inservice inspection 
following installation.
    The changes to update the wording and content of the TS to 
provide clarification and to incorporate wording enhancements are 
administrative changes that provide clarifications. These changes do 
not alter the requirements for inservice inspection or the plugging 
limit for the tubes.
    (d) The fourth area of change is to revise the steam generator 
water levels.
    The proposed steam generator water level setpoint changes do not 
impact the initiation of accidents; therefore, they do not involve 
an increase in the probability of an accident previously evaluated. 
The proposed changes do impact the safety analyses for accidents 
that credit the applicable trips and associated system actions; 
however, they do not alter these accidents or the associated 
accident acceptance criteria. The safety analyses for these 
accidents have been performed at 2900 MWt [megawatts thermal] (which 
is conservative and bounding for the current licensed power level of 
2689 MWt) and show acceptable results. Therefore, the proposed 
changes do not involve a significant increase in the consequences of 
an accident previously evaluated.
    The proposed change to steam generator water level used to 
verify steam generator operability in Modes 4 and 5, i.e., TS 
3.4.1.3, does not impact the initiation of accidents; therefore, it 
does not involve an increase in the probability of an accident 
previously evaluated. The proposed change does not alter the safety 
analyses for accidents or the associated accident acceptance 
criteria. Therefore, the proposed change does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    The proposed changes, due to the replacement steam generators, 
do not alter the requirements for tube inspection, tube integrity, 
or tube plugging limit, therefore they do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Use of the VIPRE computer code and the WRB-2M correlation at 
BVPS for departure from nucleate boiling (DNB) analysis for those 
Updated Final Safety Analysis Report (UFSAR) transients and 
accidents for which DNB might be a concern will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated for the following reasons. The code 
and correlation are evaluation tools that are independent of the 
probability of an accident. Use of the code and correlation 
establish DNB limits such that core damage will not occur during 
postulated design basis accidents. Thus, use of the code and 
correlation will not involve a significant increase in the 
consequences of an accident previously evaluated.
    Use of the 1979 ANS [American Nuclear Society] Decay Heat + 
2[sigma] 4 model for MSLB [main steam line break] outside 
containment M&E [mass and energy] releases will not have a 
significant increase in the probability or consequences of an 
accident previously evaluated because the model is not an accident 
initiator.
    The remaining changes, which include the changes to the 
Overtemperature [Delta]T and Overpower [Delta]T equations, the 
change to the charging pump discharge pressure, and the additions of 
WCAP-14565-P-A and WCAP-15025-P-A to the list of NRC approved 
methodologies in TS 6.9.5, will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated because none of the changes are accident initiators.
    The RSG radiological analysis reflects an expansion of the 
selective application of the AST methodology and incorporation of 
the ARCHON96 methodology for on-site atmospheric dispersion factors. 
The radiological analysis concludes that normal operation of the 
BVPS Unit No. 1 with the RSGs with an atmospheric containment will 
not impact the unit's compliance with the normal operation operator 
exposure limits set forth in 10 CFR 20 [Title 10 of the Code of 
Federal Regulations, Part 20], or the public exposure limits set 
forth in 10 CFR 20, 10 CFR 50, Appendix I and 40 CFR 190, or with 
the post-accident exposure limits set forth by 10 CFR 100 or 10 CFR 
50.67, as supplemented by Regulatory Guide 1.183, for the plant 
operator and the public.
    The effects on accident radiation dose considered the 
replacement of the Unit No. 1 steam generators, a core power level 
to 2900 MWt, incorporation of the ARCHON96 methodology and the 
expansion of the selective implementation of the AST methodology. 
None of these changes are initiators of any design basis accident or 
event, and therefore, will not increase the probability of any 
accident previously evaluated. The probability of any evaluated 
accident or event is independent of these changes.
    These proposed changes required alteration of some assumptions 
previously made in the radiological consequence evaluations. The 
assumption alterations were necessary to reflect the replacement 
steam generators for Unit No. 1 and the incorporation of the 
ARCHON96 and AST methodologies. These changes were evaluated for 
their effect on accident dose consequences. The updated dose 
consequence analyses demonstrate compliance with the limits set 
forth for AST

[[Page 35739]]

applications in 10 CFR 50.67, as supplemented by Regulatory Guide 
1.183 or 10 CFR part 100.
    Therefore, in conclusion, none of the proposed changes involve a 
significant increase in the probability of an accident previously 
evaluated, and the dose consequences remain within the allowable 
limits set forth for AST applications in 10 CFR 50.67, as 
supplemented by Regulatory Guide 1.183 or 10 CFR part 100.
    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 will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The areas of change described previously for the Unit No. 1 RSGs 
do not adversely affect the design or function of any other safety-
related component. With respect to postulated accident conditions, 
the OSGs and the RSGs are the same. There is no mechanism to create 
a new or different kind of accident for the RSGs by eliminating 
repair criteria or by clarifying the applicability of inservice 
inspection requirements because a baseline of tube conditions is 
established and plugging limits are maintained to ensure that 
defective tubes are identified and removed from service.
    The proposed changes to steam generator water level setpoints, 
and the steam generator water level used to verify steam generator 
operability in Modes 4 and 5 do not impact the initiation of 
accidents. They do not alter the accidents that credit the 
associated trips or accident acceptance criteria. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed changes do not alter the requirements for tube 
inspection, tube integrity, or tube plugging limit; therefore, they 
do not create the possibility of a new or different kind of accident 
from any previously evaluated.
    Use of the VIPRE computer code and WRB-2M correlation at BVPS 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated because the code and 
correlation are evaluation tools. They are not accident initiators. 
Thus, their use cannot create a new or different kind of accident.
    Use of the 1979 ANS Decay Heat + 2[sigma] model for MSLB outside 
containment M&E releases will not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the model does not alter how any equipment is operated.
    The remaining changes, which include the changes to the 
Overtemperature [Delta]T and Overpower [Delta]T equations, the 
change to the charging pump discharge pressure, and the additions of 
WCAP-14565-P-A and WCAP-15025-P-A to the list of NRC approved 
methodologies in TS 6.9.5, will not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
because these changes do not alter how any equipment is operated.
    The radiological changes will not create the possibility of a 
new or different kind of accident from any previously evaluated 
because they do not affect how components or systems are operated, 
nor do they create new components or systems failure modes.
    Therefore, in conclusion, none of the proposed changes create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The proposed changes will not involve a 
significant reduction in a margin of safety.
    The steam generator tube integrity provides the margin of 
safety. The tubing in the RSGs is designed and evaluated consistent 
with the margins of safety specified in the ASME Code, Section III. 
The program for periodic inservice inspection provides sufficient 
time to take proper and timely corrective action if tube degradation 
is present. The basis for the 40% through wall plugging limit is 
applicable to the RSGs just as it was to the OSGs. A Regulatory 
Guide 1.121 analysis was performed to confirm the applicability of 
the 40% through wall plugging limit. As a result, there is no 
reduction in tube integrity for the RSGs.
    The proposed changes to steam generator water level setpoints do 
not alter the reactor trip system/engineered safety features 
actuation system setpoint analysis methodology, or the associated 
accident analysis methodology or acceptance criteria. The safety 
analyses for these accidents have been performed at a power level of 
2900 MWt ( which is conservative and bounding for the current 
licensed power level of 2689 MWt) and show acceptable results. 
Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The proposed change to the steam generator water level used to 
verify steam generator operability in Modes 4 and 5 does not alter 
the steam generator water level uncertainty and setpoint analysis 
methodology or the associated natural circulation analysis 
methodology or acceptance criteria. Therefore, the proposed change 
does not involve a significant reduction in a margin of safety.
    The proposed changes to update the wording and content of the TS 
to provide clarification and to incorporate wording enhancements are 
administrative changes that provide clarifications.
    The proposed changes do not alter the requirements for tube 
integrity, tube inspection or tube plugging limit; therefore, they 
do not involve a significant reduction in a margin of safety.
    Use of the VIPRE computer code and the WRB-2M correlation at 
BVPS will not involve a significant reduction in a margin of safety 
because the code and correlation are used to establish a margin of 
safety previously approved by the NRC such that core damage will not 
occur.
    Use of the 1979 ANS Decay Heat + 2[sigma] model for MSLB outside 
containment M&E releases will not involve a significant reduction in 
a margin of safety because the results of the subject accident have 
been shown to produce acceptable results.
    The remaining changes, which include changes to the 
Overtemperature [sigma]T and Overpower [sigma]T equations, the 
change to the charging pump discharge pressure, and the additions of 
WCAP-14565-P-A and WCAP-15025-P-A to the list of NRC approved 
methodologies in TS 6.9.5, will not involve a significant reduction 
in a margin of safety because they are being made to maintain the 
existing margin of safety.
    The radiological changes will not involve a significant 
reduction in a margin of safety because BVPS compliance with the 
limits set forth in 10 CFR 20, 10 CFR 50, Appendix I, 40 CFR 190, 10 
CFR 100 and 10 CFR 50.67, as supplemented by Regulatory Guide 1.183, 
will be maintained following approval of the requested changes.
    A FENOC assessment of the cumulative effect of the proposed 
changes provides [a] reasonable expectation that collectively they 
will not result in a significant reduction in the overall margin of 
safety. The results of the analyses demonstrate that the applicable 
design and safety criteria and regulatory requirements will continue 
to be met following approval of the proposed changes.
    Therefore, in conclusion, none of the propose changes 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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Richard J. Laufer.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendment request: June 1, 2005.
    Description of amendment request: The amendments proposed by 
Southern Nuclear Operating Company (SNC) would revise the Technical 
Specifications (TS) to replace the previous TS requirement to implement 
a Containment Tendon Surveillance Program based on Regulatory Guide 
1.35, Rev. 2, ``Inservice Inspection of Ungrouted Tendons in 
Prestressed Concrete Containment Structures,'' with a Containment 
Inspection Program that complies with the current requirements of Title 
10 of the Code of Federal Regulations (10 CFR) Section 50.55a, ``Codes 
and Standards,'' in order to reflect the latest requirements for tendon 
surveillance.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 35740]]

licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    1. The proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change replaces the previous TS requirement to 
implement a Containment Tendon Surveillance Program based on 
Regulatory Guide 1.35, Rev. 2, with a Containment Inspection Program 
that complies with the current requirements of 10 CFR 50.55a. This 
regulation requires licensees to implement a Containment Inspection 
Program in compliance with the 1992 Edition with the 1992 Addenda of 
Subsection IWE, ``Requirements for Class MC and Metallic Liners of 
Class CC Components of Light-Water Cooled Plants,'' and with 
Subsection IWL, ``Requirements for Class CC Concrete Components of 
Light-Water Cooled Plants,'' of Section XI, Division 1, of the 
American Society of Mechanical Engineers Boiler and Pressure Vessel 
Code (ASME Code) with additional modifications and limitations as 
stated in 10 CFR 50.55a(b)(2)(ix). SNC has implemented a Containment 
Inspection Program that complies with the regulatory requirements. 
This proposed TS amendment is requested to update the TS to the 
latest 10 CFR 50.55a regulatory requirements.
    In addition, reporting requirements that are redundant to 
existing regulations are deleted, minor editorial changes are made, 
and the applicability of [Surveillance Requirement] SR 3.0.2 to the 
tendon surveillance program is deleted since surveillance 
frequencies and associated extensions are specified in ASME Section 
XI, Subsection IWL.
    By complying with the regulatory requirements described in 10 
CFR 50.55a, the probability of a loss of containment structural 
integrity is maintained as low as reasonably achievable. Maintaining 
containment structural integrity as described in the revised 
Containment Inspection Program does not impact the operation of the 
reactor coolant system (RCS), containment spray (CS) system, or 
emergency core cooling system (ECCS). The Containment Inspection 
Program ensures that the containment will function as designed to 
provide an acceptable barrier to release of radioactive materials to 
the environment. The proposed change does 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 change does not impact any accident initiators or analyzed 
events, nor does it impact the types or amounts of radioactive 
effluent that may be released offsite. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Maintaining containment structural integrity does not impact the 
operation of the RCS, CS system, or ECCS. The proposed change does 
not involve a modification to the physical configuration of the 
plant or a change in the methods governing normal plant operation. 
The proposed change does not introduce a new accident initiator, 
accident precursor, or malfunction mechanism. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed license amendment does not involve a significant 
reduction in a margin of safety.
    By complying with the regulatory requirements described in 10 
CFR 50.55a, the probability of a loss of containment structural 
integrity is maintained as low as reasonably achievable. The 
Containment Inspection Program ensures that the containment will 
function as designed to provide an acceptable barrier to release of 
radioactive materials to the environment. The proposed change does 
not adversely affect plant operation or existing safety analyses. 
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: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Section Chief: Evangelos C. Marinos.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application request: May 26, 2005.
    Description of amendment request: The amendment would change 
Technical Specification (TS) 3.7.2, ``Main Steam Isolation Valves 
(MSIVs),'' by adding the MSIV actuator trains to (1) the limiting 
condition for operation (LCO) and (2) the conditions, required actions, 
and completion times for the LCO. The existing conditions and required 
actions in TS 3.7.2 are renumbered to account for the new conditions 
and required actions.
    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 changes to incorporate requirements for the MSIV 
actuator trains do not involve any design or physical changes to the 
facility, including the MSIVs and actuator trains themselves. The 
design and functional performance requirements, operational 
characteristics, and reliability of the MSIVs and actuator trains 
are thus unchanged. There is therefore no impact on the design 
safety function of the MSIVs to close (as an accident mitigator), 
nor is there any change with respect to inadvertent closure of an 
MSIV (as a potential transient initiator). Since no failure mode or 
initiating condition that could cause an accident (including any 
plant transient) evaluated per the FSAR [Callaway Final Safety 
Analysis Report]-described safety analyses is created or affected, 
the [proposed] change[s] cannot involve a significant increase in 
the probability of an accident previously evaluated.
    With regard to the consequences of an accident and the equipment 
required for mitigation of the accident, the proposed changes 
involve no design or physical changes to the MSIVs or any other 
equipment required for accident mitigation. With respect to [the] 
MSIV actuator train allowed outage times [(i.e., completion times)], 
the consequences of an accident are independent of equipment allowed 
outage times as long [as] adequate equipment availability is 
maintained. The proposed MSIV actuator train allowed outage times 
take into account the redundancy of the MSIV actuator trains and are 
limited in extent consistent with other allowed outage times 
specified in the Technical Specifications. Adequate equipment (MSIV) 
availability would therefore continue to be required by the 
Technical Specifications. On this basis, the consequences of 
applicable, analyzed accidents (such as a main steam line break) are 
not significantly impacted by the proposed changes. Based on all of 
the above, the proposed changes do 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.
    None of the proposed changes, i.e., the addition of Conditions, 
Required Actions and Completion Times [and addition to the LCO] to 
[the] Technical Specifications for the MSIV actuator trains, involve 
a change in the design, configuration, or operational 
characteristics of the plant. No physical alteration of the plant is 
involved, as no new or different type of equipment is to be 
installed. The proposed changes do not alter any assumptions made in 
the safety analyses, nor do they involve any changes to plant 
procedures for ensuring that the plant is operated within analyzed 
limits. As such, no new failure modes or mechanisms that could cause 
a new or different kind of accident from any previously evaluated 
are being introduced.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.

[[Page 35741]]

    3. [Do] the proposed change[s] involve a significant reduction 
in a margin of safety.
    Response: No.
    The proposed addition of Conditions, Required Actions and 
Completion Times [and proposed addition to the LCO] to the Technical 
Specifications for the MSIV actuator trains does not alter the 
manner in which safety limits or limiting safety system settings are 
determined. [There are no proposed changes to safety limits or 
limiting safety system settings.] No changes to instrument/system 
actuation setpoints are involved. The safety analysis acceptance 
criteria are not impacted by [these proposed] change[s], and the 
proposed change[s] will not permit plant operation in a 
configuration outside the design basis.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Robert A. Gramm.

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 (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 Systems (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
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 PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail 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: October 21, 2004.
    Description of amendment request: The amendment deletes the 
Technical Specification (TS) requirements to submit monthly operating 
reports and annual occupational radiation exposure reports. The change 
is consistent with Revision 1 of the Nuclear Regulatory Commission 
approved Technical Specifications Task Force (TSTF) Change Traveler, 
TSTF-369, ``Elimination of Requirements for Monthly Operating Reports 
and Occupational Radiation Exposure Reports.'' This TS improvement was 
published in the Federal Register (69 FR 35067) on June 23, 2004, as 
part of the Consolidated Line Item Improvement Process.
    Date of issuance: June 8, 2005.
    Effective date: June 8, 2005.
    Amendment No.: 254.
    Facility Operating License No. DPR-16: Amendment revises the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. April 8, 2005 (70 FR 18056). The notice 
provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. Comments received from the State of New Jersey are 
discussed in Section 7.0 of the related safety evaluation. The notice 
also provided an opportunity to request a hearing by June 7, 2005, but 
indicated that if the Commission makes a final NSHC determination, any 
such hearing would take place after issuance of the amendment. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated June 8, 2005.
    Attorney for licensee: Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Richard J. Laufer.

Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423, 
Millstone Power Station, Unit No. 3, New London County, Connecticut

    Date of application for amendment: September 7, 2004.
    Brief description of amendment: The amendment revised the required 
frequency of quench and recirculation spray nozzle surveillances from 
once every 10 years to ``following maintenance which could result in 
nozzle blockage.'' The change also revised wording to correct grammar.
    Date of issuance: May 31, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days from the date of issuance.
    Amendment No.: 222.
    Facility Operating License No. NPF-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 7, 2004 (69 FR 
70715).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 31, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 3, 2003, as supplemented 
by letter dated January 18 and May 10, 2005.
    Brief description of amendments: The amendments would add a note to 
Limiting Condition of Operation 3.7.11, ''Auxiliary Building Filtered 
Ventilation Exhaust System (ABFVES),'' that would allow the Auxiliary 
Building pressure boundary to be opened intermittently under 
administrative control.
    Date of issuance: June 2, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 229 and 211.

[[Page 35742]]

    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 16, 2004 (69 FR 
12365). The supplements dated January 18 and May 10, 2005, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 2, 2005.
    No significant hazards consideration comments received: No.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: December 20, 2004.
    Brief description of amendment: The amendment deletes the 
requirements related to monthly operating reports and occupational 
radiation exposure reports.
    Date of issuance: May 25, 2005.
    Effective date: As of the date of issuance, and shall be 
implemented 90 days from the date of issuance.
    Amendment No.: 145.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 1, 2005 (70 FR 
9990).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 25, 2005.
    No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of application for amendment: December 7, 2004.
    Brief description of amendment: This amendment revised the 
Technical Specifications (TSs) by removing the surveillance requirement 
(SR) for testing the setting of the standby liquid control system 
pressure relief valves. Also, the SR for the recirculation pump 
discharge valves was revised to remove stroke time specifications.
    Date of Issuance: June 1, 2005.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 224.
    Facility Operating License No. DPR-28: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2889).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated June 1, 2005.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendments: April 6, 2004, as supplemented 
by four letters dated April 15, 2005.
    Brief description of amendments: The amendments convert the current 
Technical Specifications (CTS) to the improved Technical Specifications 
(ITS) and relocate license conditions to the ITS or other license 
controlled documents. The ITS are based on NUREG-1431, ``Standard 
Technical Specifications, Westinghouse Plants,'' dated April 30, 2001, 
and guidance provided in the Commission's Final Policy Statement, ``The 
U.S. Nuclear Regulatory Commission Final Policy Statement on Technical 
Specifications (TSs) Improvements for Nuclear Power Reactors,'' 
published on July 22, 1993 (58 FR 39132), and 10 CFR Part 50.36, 
``TSs.'' The overall objective of the proposed amendments was to 
rewrite, reformat, and streamline the CTS to improve plant safety and 
the understanding of the bases underlying the TSs.
    Date of issuance: June 1, 2005.
    Effective date: As of the date of issuance and shall be implemented 
by October 30, 2005.
    Amendment Nos.: 287, 269.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: September 29, 2004 (69 
FR 58205). The supplemental letters contained clarifying information 
and did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 1, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 22, 2004.
    Brief description of amendment: The amendment deleted Sections 5.3, 
``Reactor Vessel,'' 5.4, ``Containment,'' and 5.6, ``Seismic Design,'' 
relocating all information, which pertains to design details, to the 
Updated Final Safety Analysis Report.
    Date of issuance: June 6, 2005.
    Effective date: As of the date of issuance to be implemented within 
90 days.
    Amendment No.: 189.
    Facility Operating License No. DPR-63: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 7, 2004 (69 FR 
70719).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 6, 2005.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of application for amendments: November 5, 2003, as 
supplemented by letter dated April 22, 2004.
    Brief description of amendments: The amendments revised the Point 
Beach Nuclear Plant (PBNP), Units 1 and 2, Updated Final Safety 
Analysis Report [UFSAR] to reflect the Commission staff's approval of 
the WCAP-14439-P, Revision 2 analysis entitled, ``Technical 
Justification for Eliminating Large Primary Loop Pipe Rupture as the 
Structural Design Basis for the Point Beach Nuclear Plant Units 1 and 2 
for the Power Uprate and License Renewal Program.''
    Date of issuance: June 6, 2005.
    Effective date: As of the date of issuance and shall be implemented 
with the next update of the UFSAR in accordance with 10 CFR 50.71(e).
    Amendment Nos.: 219, 224.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the License.
    Date of initial notice in Federal Register: February 7, 2005 (70 FR 
6466). The supplement dated April 22, 2004, provided clarifying 
information that did not change the scope of the amendment, application 
nor the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 6, 2005.
    No significant hazards consideration comments received: No.

[[Page 35743]]

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of application for amendments: October 15, 2004.
    Brief description of amendments: The amendments revised Technical 
Specifications related to the reactor coolant pump flywheel inspection 
program by increasing the inspection interval to 20 years.
    Date of issuance: June 6, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 218, 223.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 29, 2005 (70 FR 
15945).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 6, 2005.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of application for amendments: October 15, 2004.
    Brief description of amendments: The amendments revise Technical 
Specifications related to the reactor coolant pump flywheel inspection 
program by increasing the inspection interval to 20 years.
    Date of issuance: June 7, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 170, 160.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 15