[Federal Register Volume 70, Number 30 (Tuesday, February 15, 2005)]
[Notices]
[Pages 7762-7777]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-2788]


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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 January 20, 2005, through February 3, 2005. 
The last biweekly notice was published on February 1, 2005 (70 FR 
5233).

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 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 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

[[Page 7763]]

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, [email protected]; 
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 
[email protected]. 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, 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 PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by email to [email protected].

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina; Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 
1, Wake and Chatham Counties, 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: November 17, 2004.
    Description of amendments request: The requested change would 
delete Technical Specification (TS) 5.6.1, ``Occupational Radiation 
Exposure Report,'' and TS 5.6.4, ``Monthly Operating Reports,'' for the 
Brunswick and H. B. Robinson plants. The equivalent change is being 
requested for the Shearon Harris facility by deleting TS 6.9.1.2.a and 
TS 6.9.1.2.c under ``Annual Reports'' and TS 6.9.1.5, ``Monthly 
Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated November 17, 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:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates the Technical Specifications (TS) 
reporting requirements to provide a monthly operating report of 
shutdown experience and operating statistics if the equivalent data 
is submitted using an industry electronic database. It also 
eliminates the TS reporting requirement for an annual occupational 
radiation exposure report, which provides information beyond that 
specified in NRC regulations. The proposed change involves no 
changes to plant systems or accident analyses. As such, the change 
is administrative in nature and does not affect initiators of 
analyzed events or assumed mitigation of accidents or transients. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    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 physical alteration of 
the plant, add any new

[[Page 7764]]

equipment, or require any existing equipment to be operated in a 
manner different from the present design. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in the margin of safety.

    Based upon the reasoning presented above, the requested change does 
not involve a significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Michael L. Marshall.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of amendment request: January 5, 2005.
    Description of amendment request: The proposed amendments would 
revise the Technical Specification (TS) 5.5.19 associated with the Lee 
Combustion Turbine (LCT) testing program. TS 5.5.19.b currently 
requires verification that an LCT can supply the equivalent of one 
Unit's maximum safeguard loads, plus two Units' Mode 3 loads, when 
connected to the system grid every 12 months. In the proposed 
amendments, this requirement would be more clearly specified as ``plus 
two Units' safe shutdown loads.''
    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) Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    Duke proposes to revise TS 5.5.19.b to clarify the Lee 
Combustion Turbine (LCT) testing requirements. The proposed change 
makes the wording of the test requirement consistent with the UFSAR 
[Updated Final Safety Analysis Report] and the original wording of 
the TS requirement before administrative changes were made in 
Amendment 232, 232, 231, and Amendment 300, 300, and 300. LCT 
testing has no impact on the probability of an accident analyzed in 
the UFSAR. The LCT can be credited to mitigate the consequences of 
an accident analyzed in the UFSAR. However, this clarification of 
LCT testing requirements has no impact on its ability to mitigate 
the consequences of an accident. As such, the proposed LAR [license 
amendment request] does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated:
    Duke proposes the revise TS 5.5.19.b to clarify the Lee 
Combustion Turbine (LCT) testing requirements. The proposed change 
makes wording of the test requirement consistent with the UFSAR and 
the original wording of the TS requirement before administrative 
changes were made in Amendment 232, 232, 231, and changes were made 
in Amendment 300, 300, and 300. These changes do not alter the 
nature of events postulated in the Safety Analysis Report nor do 
they introduce any unique precursor mechanisms. Therefore, the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Involve a significant reduction in a margin of safety.
    The proposed TS change does not unfavorably affect any plant 
safety limits, set points, or design parameters. The changes also do 
not unfavorably affect the fuel, fuel cladding, RCS [reactor coolant 
system], or containment integrity. Therefore, the proposed TS 
change, which clarifies TS requirements associated with the LCT 
testing program, 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: Anne W. Cottingham, Winston and Strawn LPP, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: John A. Nakoski.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: July 29, 2004.
    Description of amendment request: The proposed amendment would 
delete the requirements from the technical specifications (TS) to 
maintain a hydrogen dilution system, a hydrogen purge system, and 
hydrogen monitors. Licensees were generally required to implement 
upgrades as described in NUREG-0737, ``Clarification of TMI [Three Mile 
Island] Action Plan Requirements,'' and Regulatory Guide (RG) 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
combustible gas control were imposed by order for many facilities and 
were added to or included in the TS for nuclear power reactors 
currently licensed to operate. The revised Title 10 of the Code of 
Federal Regulations (10 CFR) section 50.44, ``Combustible gas control 
for nuclear power reactors,'' eliminated the requirements for hydrogen 
recombiners and related vent and purge systems and relaxed safety 
classifications and licensee commitments to certain design and 
qualification criteria for hydrogen and oxygen monitors.
    The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice 
of availability of a model no significant hazards consideration 
determination for referencing in license amendment applications in the 
Federal Register on September 25, 2003 (68 FR 55416). The licensee 
affirmed the applicability of the model no significant hazards 
consideration determination in its application dated July 29, 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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The NRC has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
Category 1 in RG 1.97 is intended for key variables that most 
directly indicate

[[Page 7765]]

the accomplishment of a safety function for design-basis accident 
events. The hydrogen monitors no longer meet the definition of 
Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 
50.44, the NRC found that Category 3, as defined in RG 1.97, is an 
appropriate categorization for the hydrogen monitors because the 
monitors are required to diagnose the course of beyond design-basis 
accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the severe accident 
management guidelines, the emergency plan, the emergency operating 
procedures, and site survey monitoring that support modification of 
emergency plan protective action recommendations.
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from the TS, does not 
involve a significant increase in the probability or the 
consequences of any 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 elimination of the hydrogen recombiner [dilution/purge 
system for Davis Besse] requirements and relaxation of the hydrogen 
monitor requirements, including removal of these requirements from 
TS, will not result in any failure mode not previously analyzed. The 
hydrogen recombiner [dilution/purge system for Davis Besse] and 
hydrogen monitor equipment was intended to mitigate a design-basis 
hydrogen release. The hydrogen recombiner [dilution/purge system for 
Davis Besse] and hydrogen monitor equipment are not considered 
accident precursors, nor does their existence or elimination have 
any adverse impact on the pre-accident state of the reactor core or 
post accident confinement of radionuclides within the containment 
building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the hydrogen recombiner [dilution/purge 
system for Davis Besse] requirements and relaxation of the hydrogen 
monitor requirements, including removal of these requirements from 
TS, in light of existing plant equipment, instrumentation, 
procedures, and programs that provide effective mitigation of and 
recovery from reactor accidents, results in a neutral impact to the 
margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The NRC has found that this hydrogen 
release is not risk-significant because the design-basis LOCA 
hydrogen release does not contribute to the conditional probability 
of a large release up to approximately 24 hours after the onset of 
core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI, Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TS will 
not result in a significant reduction in their functionality, 
reliability, and availability.
    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.

    Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Gene Y. Suh.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: December 20, 2004.
    Description of amendment request: The proposed change would revise 
Technical Specification (TS) 3/4.9.2, ``Refueling Operations--
Instrumentation,'' concerning source range neutron flux monitors to be 
consistent with Improved Standard Technical Specifications.
    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 changes affect the Limiting Condition for Operation [LCO] 
for Refueling Operations--Instrumentation, in particular, the LCO 
sections pertaining to the source range neutron flux detectors will 
be changed to be more like the corresponding sections in the 
Improved Standard Technical Specifications. The source range neutron 
flux detectors have no control functions and are therefore not 
accident initiators. Consequently, the proposed changes will have no 
impact on the probability of any accident previously evaluated. The 
detectors are not credited in mitigating the consequences of any 
accident; therefore, the proposed changes will have no impact on the 
consequences of any accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The changes affect the Limiting Condition for Operation for 
Refueling Operations--Instrumentation, in particular, the source 
range neutron flux detectors. The source range neutron flux 
detectors will continue to operate in the same manner as previously 
considered. Accident initial conditions and assumptions remain as 
previously analyzed.
    The proposed changes do not introduce any new or different 
accident initiators. Therefore, the proposed 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 changes affect the Limiting Condition for Operation for 
Refueling Operations--Instrumentation; in particular, the source 
range neutron detectors. These detectors have no control functions, 
and are not credited in mitigating the consequences of any accident. 
The source range neutron detectors are not associated with a safety 
limit. In addition, the proposed changes to TS will not result in 
design changes to the source range neutron detectors or in changes 
to how the source range detectors are used. Therefore, the proposed 
changes will not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Gene Y. Suh.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: January 5, 2005.
    Description of amendment request: The license amendment would 
revise Technical Specification 3/4.3.2.1, ``Safety Features Actuation 
System [SFAS] Instrumentation,'' to permit a single inoperable SFAS 
functional unit to be placed in a bypassed condition indefinitely.

[[Page 7766]]

    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 would permit a single SFAS instrument string 
functional unit to be placed in bypass indefinitely. The primary 
function of SFAS is to monitor station conditions and actuate the 
engineered safety features when needed in order to prevent or limit 
fission product and energy release from the core, to isolate the 
containment vessel, and to initiate the operation of the Engineered 
Safety Features (ESF) equipment in the event of a loss-of-coolant 
accident (LOCA).
    The SFAS is a possible accident initiator in that an inadvertent 
system level actuation could result in a transient or accident. The 
existing Technical Specification requirements for SFAS allow 
operation indefinitely with a single SFAS functional unit in trip, 
which results in a 1-out-of-3 channel logic. In this condition, the 
spurious actuation in one of the three remaining corresponding 
functional unit would result in an inadvertent system level 
actuation. Under the proposed change, indefinite operation in a 2-
out-of-3, 1-out-of-3, or 1-out-of-2 channel logic would be allowed. 
The likelihood of a spurious system level actuation for any of the 
configurations allowed under the proposed change is no greater than 
the likelihood of spurious actuation under the 1-out-of-3 channel 
logic allowed under the existing Technical Specification 
requirements. Therefore, operation of the SFAS actuation from that 
permitted by the existing Technical Specifications.
    Under the proposed change, the SFAS will continue to perform 
this function with a high level of reliability. The proposed change 
would allow operation of the SFAS in a condition with reduced 
redundancy from what is currently required by the Technical 
Specifications. Operation of the SFAS with reduced redundancy was 
evaluated against the design criteria to which the system was 
designed. The design criteria applicable to the SFAS, including the 
single failure criterion, continue to be met. The proposed change 
does not prevent the SFAS from mitigating the consequences of 
previously analyzed accidents.
    The proposed change would not increase the likelihood of an 
inadvertent SFAS actuation. The proposed change would not prevent 
the SFAS from mitigating the consequences of previously analyzed 
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 does not alter the SFAS design function or 
the manner in which that function is performed. Under the proposed 
change, the SFAS will continue to perform its function with a high 
degree of reliability. No new failure modes or accident initiators 
are created by the proposed change. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change would allow operation of the SFAS in a 
condition with reduced redundancy from what is currently required by 
the Technical Specifications. Operation of the SFAS with reduced 
redundancy was evaluated against the design criteria to which the 
system was designed. This evaluation shows that with the SFAS in the 
conditions permitted by the proposed change, the SFAS still 
satisfies all the applicable design criteria, including the single 
failure criterion. 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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Gene Y. Suh.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: January 11, 2005.
    Description of amendment request: The proposed amendment would 
revise the Updated Safety Analysis Report (USAR) by modifying the 
design requirements for protection from tornado missiles. Specifically, 
the proposed amendment would allow certain structures, systems, and 
components that are not currently provided with physical protection 
from tornado-induced missiles to be evaluated for acceptability based 
on the Electrical Power Research Institute ``Tornado Missile Risk 
Evaluation Methodology'' (TORMIS).
    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 amendment would reflect use of the Electric Power 
Research Institute (EPRI) Topical Report ``Tornado Missile Risk 
Evaluation Methodology'' (EPRI NP-2005), Volumes I and II. As noted 
in the NRC Safety Evaluation on this report dated October 26, 1983, 
``The current licensing criteria governing tornado missile 
protection are contained in Standard Review Plan (SRP) Sections 
3.5.1.4 and 3.5.2. These criteria generally specify that safety-
related systems be provided positive tornado missile protection 
(barriers) from the maximum credible tornado threat. However, SRP 
Section 3.5.1.4 includes acceptance criteria permitting relaxation 
of the above deterministic guidance, if it can be demonstrated that 
the probability of damage to unprotected essential safety-related 
features is sufficiently small.''
    ``Certain Operating License (OL) applicants and operating 
reactor licensees have chosen to demonstrate compliance with tornado 
missile protection criteria for certain portions of the plant * * * 
by providing a probabilistic analysis which is intended to show a 
sufficiently low risk associated with tornado missiles. Some* * * 
have utilized the tornado missile probabilistic risk assessment 
(PRA) methodology developed by'' EPRI in the Topical Report listed 
above. The NRC noted that this report ``can be utilized when 
assessing the need for positive tornado missile protection for 
specific safety-related plant features.'' This methodology has 
subsequently been utilized in nuclear power plant licensing actions.
    As permitted in NRC Standard Review Plan (NUREG-0800) sections, 
the total probability will be maintained below an allowable level, 
i.e., an acceptance criteria threshold, which reflects an extremely 
low probability of occurrence. The DBNPS [Davis-Besse Nuclear Power 
Station] approach assumes that if the probability calculation result 
for the total plant identifies that the cumulative probability of 
tornado missiles striking an unprotected portion of a safety system 
or component required for safe shutdown in the event of a tornado 
exceeds 10-6 per year, then unique missile barriers would 
need to be installed to lower the total probability below the 
acceptance criteria of 10-6 per year.
    With respect to the probability of occurrence of an accident 
previously evaluated in the USAR, the possibility of a tornado 
reaching the DBNPS site and causing damage to plant structures, 
systems, and components is an event considered in the USAR. The 
changes being proposed herein do not affect the probability that the 
natural phenomena (a tornado) will reach the plant, but they do, 
from a licensing basis perspective, affect the probability that 
missiles generated by the winds of the tornado might strike certain 
plant systems or components. As recently determined, there are a 
limited number of safety-related components that could theoretically 
be struck by a tornado generated missile. The

[[Page 7767]]

total (cumulative) probability of a tornado missile striking an 
unprotected component will be maintained below an extremely low 
acceptance criteria to ensure overall plant safety. Due to the 
extremely low probability of a tornado missile impacting an 
essential component, the small increase in the probability of 
accident initiation is not considered significant.
    With respect to the consequences of an accident previously 
evaluated, there is an extremely low probability of a malfunction of 
an unprotected essential component due to tornado missile impact. 
Due to (1) the extremely low probability of a tornado missile 
striking essential equipment as calculated by TORMIS, and (2) the 
low probability that any tornado missile strikes would cause 
sufficient damage to prevent essential equipment from performing its 
accident-mitigating function, a loss of accident mitigation 
capability is not considered credible. Therefore, the radiological 
consequences of accidents are not significantly affected.
    The proposed change is not considered to constitute a 
significant increase in the probability of occurrence or the 
consequences of an accident, due to the extremely low total 
probability of a tornado missile strike and thus an extremely low 
probability of a radiological release. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of previously evaluated accidents.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The possibility of a tornado reaching the DBNPS site is a design 
basis event considered in the USAR. This change involves recognition 
of the acceptability of performing tornado missile probability 
calculations in accordance with established regulatory guidance. The 
change therefore deals with an established design basis event (the 
tornado). Therefore, the proposed change would not contribute to the 
possibility of a new or different kind of accident from those 
previously analyzed.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This request does not involve a significant reduction in a 
margin of safety. The existing licensing basis for the DBNPS with 
respect to the design basis event of a tornado reaching the plant is 
to provide positive missile barriers for all systems and components 
required for safe shutdown in the event of a tornado. With the 
change, it will be recognized that there is an extremely low 
probability, below an established acceptance limit, that a limited 
subset of these systems and components could be struck. The change 
to missile protection based on extremely low probability (less than 
1 x 10-6 per year cumulative strike probability) of 
occurrence of tornado generated missile strikes on portions of these 
systems and components is not considered to constitute a significant 
decrease in the margin of safety due to that extremely low 
probability. Therefore, the changes associated with this license 
amendment 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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Gene Y. Suh.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio

    Date of amendment request: September 10, 2004.
    Description of amendment request: The proposed amendment would 
delete the requirements from the technical specifications (TS) to 
maintain hydrogen recombiners and hydrogen monitors. Licensees were 
generally required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' 
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' Implementation of these upgrades 
was an outcome of the lessons learned from the accident that occurred 
at TMI, Unit 2. Requirements related to combustible gas control were 
imposed by Order for many facilities and were added to or included in 
the TS for nuclear power reactors currently licensed to operate. The 
revised Title 10 of the Code of Federal Regulations (10 CFR) section 
50.44, ``Standards for Combustible Gas Control System in Light-Water-
Cooled Power Reactors,'' eliminated the requirements for hydrogen 
recombiners and related vent and purge systems and relaxed safety 
classifications and licensee commitments to certain design and 
qualification criteria for hydrogen and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration determination for referencing in 
license amendment applications in the Federal Register on September 25, 
2003 (68 FR 55416). The licensee affirmed the applicability of the 
model no significant hazards consideration determination in its 
application dated September 10, 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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
Category 1 in RG 1.97 is intended for key variables that most 
directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen monitors no longer meet 
the definition of Category 1 in RG 1.97. As part of the rulemaking 
to revise 10 CFR 50.44, the Commission found that Category 3, as 
defined in RG 1.97, is an appropriate categorization for the 
hydrogen monitors because the monitors are required to diagnose the 
course of beyond design-basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the severe accident 
management guidelines, the emergency plan, the emergency operating 
procedures, and site survey monitoring that support modification of 
emergency plan protective action recommendations.
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from the TS, does not 
involve a significant increase in the probability or the 
consequences of any accident previously evaluated.

[[Page 7768]]

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen monitor equipment are 
not considered accident precursors, nor does their existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within 
the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a 
neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI, Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TS will 
not result in a significant reduction in their functionality, 
reliability, and availability.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Gene Y. Suh.

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

    Date of amendment request: November 17, 2004.
    Description of amendment request: The requested change would delete 
Technical Specification (TS) 5.7.1.1.a, ``Occupational Radiation 
Exposure Report,'' and TS 5.7.1.2, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated November 17, 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:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates the Technical Specifications (TS) 
reporting requirements to provide a monthly operating report of 
shutdown experience and operating statistics if the equivalent data 
is submitted using an industry electronic database. It also 
eliminates the TS reporting requirement for an annual occupational 
radiation exposure report, which provides information beyond that 
specified in NRC regulations. The proposed change involves no 
changes to plant systems or accident analyses. As such, the change 
is administrative in nature and does not affect initiators of 
analyzed events or assumed mitigation of accidents or transients. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    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 physical alteration of 
the plant, add any new equipment, or require any existing equipment 
to be operated in a manner different from the present design. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in the margin of safety.

    Based upon the reasoning presented above, the requested change does 
not involve a significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Michael L. Marshall.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R. E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: August 6, 2004.
    Description of amendment request: The proposed amendment deletes 
the requirements from the Technical Specifications (TSs) to maintain 
hydrogen recombiners and hydrogen monitors. Licensees were generally 
required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' 
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' Implementation of these upgrades 
was an outcome of the lessons learned from the accident that occurred 
at TMI, Unit 2. Requirements related to combustible gas control were 
imposed by Order for many facilities and were added to or included in 
the TSs for nuclear power reactors currently licensed to operate. The 
revised 10 CFR 50.44, ``Standards for Combustible Gas Control System in 
Light-Water-Cooled Power Reactors,'' eliminated the requirements for 
hydrogen recombiners and relaxed safety classifications and licensee 
commitments to certain design and qualification criteria for hydrogen 
and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration determination for referencing in 
license amendment applications in the Federal Register on September 25, 
2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated August 6, 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:

[[Page 7769]]

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

    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
Category 1 in RG 1.97 is intended for key variables that most 
directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen monitors no longer meet 
the definition of Category 1 in RG 1.97. As part of the rulemaking 
to revise 10 CFR 50.44 the Commission found that Category 3, as 
defined in RG 1.97, is an appropriate categorization for the 
hydrogen monitors because the monitors are required to diagnose the 
course of beyond design-basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the severe accident 
management guidelines (SAMGs), the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TS, does not involve a 
significant increase in the probability or the consequences of any 
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 elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen monitor equipment are 
not considered accident precursors, nor does their existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within 
the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a 
neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI, Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TS will 
not result in a significant reduction in their functionality, 
reliability, and availability.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Daniel F. Stenger, Ballard Spahr Andrews & 
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 
20005.
    NRC Section Chief: Richard J. Laufer.

Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220 and 50-410, 
Nine Mile Point Nuclear Station, Unit Nos. 1 and 2 (NMP1 and NMP2), 
Oswego County, New York

    Date of amendment request: January 24, 2005.
    Description of amendment request: The licensee proposed amendments 
to delete Sections 6.6.1 and 5.6.1, ``Occupational Radiation Exposure 
Report,'' and Sections 6.6.4 and 5.6.4, ``Monthly Operating Reports,'' 
from the NMP1 and NMP2 Technical Specifications, respectively. The NRC 
staff issued a notice of availability of a model no significant hazards 
consideration (NSHC) determination for referencing in license amendment 
applications in the Federal Register on June 23, 2004 (69 FR 35067). 
The licensee affirmed the applicability of the model NSHC determination 
in its application dated January 24, 2005.
    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 by referencing the model NSHC analysis published by the 
NRC staff. The model NSHC analysis is reproduced 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 eliminates the Technical Specifications 
(TSs) reporting requirements to provide a monthly operating report 
of shutdown experience and operating statistics if the equivalent 
data is submitted using an industry electronic database. It also 
eliminates the TS reporting requirement for an annual occupational 
radiation exposure report, which provides information beyond that 
specified in NRC regulations. The proposed change involves no 
changes to plant systems or accident analyses. As such, the change 
is administrative in nature and does not affect initiators of 
analyzed events or assumed mitigation of accidents or transients. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    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 physical alteration of 
the plant, add any new equipment, or require any existing equipment 
to be operated in a manner different from the present design. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.

[[Page 7770]]

    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in the margin of safety.

    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.

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

    Date of amendment request: October 28, 2004.
    Brief description of amendments: The proposed amendment deletes the 
requirements from the Technical Specifications (TS) to maintain 
hydrogen recombiners and hydrogen monitors. Licensees were generally 
required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' 
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' Implementation of these upgrades 
was an outcome of the lessons learned from the accident that occurred 
at TMI Unit 2. Requirements related to combustible gas control were 
imposed by Order for many facilities and were added to or included in 
the TS for nuclear power reactors currently licensed to operate. The 
revised 10 CFR 50.44, ``Standards for Combustible Gas Control System in 
Light-Water-Cooled Power Reactors,'' eliminated the requirements for 
hydrogen recombiners and relaxed safety classifications and licensee 
commitments to certain design and qualification criteria for hydrogen 
and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on September 
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated October 28, 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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
Category 1 in RG 1.97 is intended for key variables that most 
directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen monitors no longer meet 
the definition of Category 1 in RG 1.97. As part of the rulemaking 
to revise 10 CFR 50.44 the Commission found that Category 3, as 
defined in RG 1.97, is an appropriate categorization for the 
hydrogen monitors because the monitors are required to diagnose the 
course of beyond design-basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the severe accident 
management guidelines (SAMGs), the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TS, does not involve a 
significant increase in the probability or the consequences of any 
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 elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen monitor equipment are 
not considered accident precursors, nor does their existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within 
the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a 
neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TS will 
not result in a significant reduction in their functionality, 
reliability, and availability.

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

[[Page 7771]]

    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Michael K. Webb (Acting).

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: November 4, 2004.
    Description of amendment request: The proposed changes would 
relocate the inservice testing requirements, remove the inservice 
inspection requirements, and add a Bases Control Program to the 
Administrative Controls section of the Technical Specifications (TS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of Surry Units 1 and 2 in accordance with the 
proposed Technical Specifications change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change is administrative in nature, and station 
operations are not being affected. The ASME [American Society of 
Mechanical Engineers] Code requirements are established, reviewed 
and approved by ASME, the industry and ultimately endorsed by the 
NRC for inclusion into 10 CFR 50.55a. Updates to the ASME Code 
reflect advances in technology and consider information obtained 
from plant operating experience to provide enhanced inspection and 
testing. Thus, the proposed change only modifies TS to appropriately 
reference the recently NRC approved Inservice Testing Program for 
the fourth interval at Surry Power Station. Consequently, the 
probability or consequences of an accident previously evaluated are 
not increased.
    2. The proposed Technical Specifications change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    As noted above, the proposed change is administrative in nature, 
and no new accident precursors are being introduced. Since the 
inservice testing will continue to be performed in accordance with 
an NRC approved program, adequate assurance is provided to ensure 
the safety-related pumps and valves would operate as required. No 
new testing is required that could create a new or different type of 
accident. Consequently, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed Technical Specifications change does not involve 
a significant reduction in a margin of safety.
    Performing inservice testing of pumps and valves to the NRC 
approved program for the fourth interval at Surry Power Station 
provides adequate assurance that the safety-related pumps and valves 
will continue to perform their intended safety function. This is an 
administrative change in nature and as such 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: Ms. Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Section Chief: John A. Nakoski.

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

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant 
(WBN), Unit 1, Rhea County, Tennessee

    Date of amendment request: September 15, 2004.
    Brief description of amendment request: In accordance with 
Technical Specification Task Force (TSTF) 285, Charging Pump Swap Low-
Temperature Over-Pressurization Allowance, LCO 3.4.12, Cold 
Overpressure Mitigation System (COMS), is being revised to modify and 
relocate two notes in the WBN Technical Specifications. The changes are 
all administrative, except a change which would allow two charging 
pumps to be made capable of injecting into the Reactor Coolant System 
to support pump swap operations for a period not to exceed one hour 
instead of the currently allowed 15 minutes.
    Date of publication of individual notice in Federal Register: 
February 1, 2005 (70 FR 5226).
    Expiration date of individual notice: March 3, 2005 (public 
comments) and April 4, 2005 (hearing requests).

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, 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 PDR Reference staff at 1 (800) 397-4209,

[[Page 7772]]

(301) 415-4737 or by e-mail to [email protected].

Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-318, Calvert 
Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

    Date of application of amendment: September 30, 2003.
    Description of amendment request: The amendment modifies Technical 
Specification (TS) 4.3.1, ``Criticality,'' adds TS 3.7.16, ``Spent Fuel 
Pool Boron Concentration,'' and adds TS 3.7.17, ``Spent Fuel Pool 
Storage.'' Specifically, the amendment increases the maximum enrichment 
limit of the fuel assemblies that can be stored in the Unit 2 spent 
fuel pool by taking credit for soluble boron, burnup, and configuration 
control in maintaining acceptable margins of subcriticality.
    Date of issuance: January 27, 2005.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 246.
    Renewed License No. DPR-69: Amendment revised the Technical 
Specifications.
    Date of initial notice in Federal Register: January 20, 2004 (69 FR 
2739).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 27, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 26, 2004, as supplemented 
January 26, 2005.
    Brief description of amendments: These amendments revise the 
Technical Specifications by eliminating the requirements associated 
with hydrogen and oxygen monitors.
    Date of issuance: February 2, 2005.
    Effective date: As of its date of issuance, and shall be 
implemented within 120 days.
    Amendment Nos.: 234 and 261.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: August 31, 2004 (69 FR 
53100). The January 26, 2005, supplement contained clarifying 
information only and did not change the initial proposed no significant 
hazards consideration determination or expand the scope of the initial 
application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 2, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 25, 2003, as 
supplemented June 9, and July 30, 2003, and September 13, 2004.
    Brief description of amendments: The amendments revised the 
Technical Specifications to incorporate a Steam Generator (SG) program 
that defines a performance-based approach to maintaining SG tube 
integrity. The SG program includes performance criteria that define the 
basis for tube integrity and provides reasonable assurance that SG 
tubing will remain capable of fulfilling its safety function of 
maintaining reactor coolant system pressure boundary integrity. The 
proposed amendments add a new TS for SG tube integrity (3.4.18) and 
revise the TS for reactor coolant operation leakage (3.4.13), SG tube 
surveillance program (5.5.9), and SG tube inspection report (5.6.8).
    Date of issuance: January 13, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 218 and 212.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2003 (68 FR 
40712).
    The supplements dated June 9, and July 30, 2003, and September 13, 
2004, 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 January 13, 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: August 18, 2004.
    Brief description of amendments: The amendments revised the 
Technical Specifications to remove references to Safety Injection Steam 
Line Pressure-Low.
    Date of issuance:
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 224 and 206.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the Technical Specifications.
    Date of initial notice in  Federal Register: November 9, 2004 (69 
FR 64987).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 27, 2005.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Units 1 and 2, Ogle County, Illinois; Docket Nos. STN 
50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, 
Illinois

    Date of application for amendments: August 15, 2003, as 
supplemented on April 9, 2004.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3.4.15, ``RCS Leakage Detection Instrumentation'', 
to require one containment sump monitor and one containment atmosphere 
particulate radioactivity monitor to be operable in Modes 1, 2, 3, and 
4.
    Date of issuance: January 14, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 140, 133.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in  Federal Register: October 28, 2003 (68 
FR 61477).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 14, 2005.
    No significant hazards consideration comments received: No.

Power and Light Company, et al., Docket No. 50-389, St. Lucie Plant, 
Unit No. 2, St. Lucie County, Florida

    Date of application for amendment: December 2, 2003, as 
supplemented by letters dated September 14 and December 10, 2004, and 
January 7, 2005.
    Brief description of amendment: This amendment revised the 
Technical Specifications (TSs) to permit operation

[[Page 7773]]

with a reduced reactor coolant system flow corresponding to a steam 
generator (SG) tube plugging level of 30-percent per SG. This amendment 
also includes the transition to Westinghouse Reload Safety Evaluation 
Methodology (WCAP-9272).
    Date of issuance: January 31, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 138.
    Renewed Facility Operating License No. NPF-16: Amendment revised 
the TSs.
    Date of initial notice in  Federal Register: March 18, 2004 (69 FR 
12873).
    The September 14 and December 10, 2004, and January 7, 2005, 
supplements did not affect the original proposed no significant hazards 
determination, or expand the scope of the request as noticed in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 31, 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: April 19, 2004, as supplemented 
on July 16, 2004.
    Brief description of amendment: The amendment revised Section 3/
4.6.2, ``Protective Instrumentation,'' to establish a 24-month 
operating cycle calibration frequency for the intermediate range 
monitor instrumentation. In addition, the amendment authorized 
relocation of the limiting conditions for operation and surveillance 
requirements for certain control rod withdrawal block instruments from 
Section 3/4.6.2 to the Updated Final Safety Analysis Report.
    Date of issuance: January 25, 2005.
    Effective date: January 25, 2005.
    Amendment No.: 186.
    Facility Operating License No. DPR-63: Amendment revised the 
Technical Specifications.
    Date of initial notice in  Federal Register: May 25, 2004 (69 FR 
29769).
    The July 16, 2004, letter provided clarifying information within 
the scope of the original application and did not change the staff's 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated January 25, 2005.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of application for amendment: October 5, 2004.
    Brief description of amendment: The amendment deletes Technical 
Specification (TS) 5.6.1, ``Occupational Radiation Exposure Report,'' 
and TS 5.6.4 ``Monthly Operating Reports,'' as described in the Notice 
of Availability published in the Federal Register on June 23, 2004 (69 
FR 35067).
    Date of issuance: January 31, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 256.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in  Federal Register: November 9, 2004 (69 
FR 64989).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 31, 2005.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of application for amendment: January 30, 2004.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TSs) to (1) clarify the permissive setpoint for the 
source range monitor detector-not-fully-inserted rod block bypass, (2) 
correct a typographical error in the surveillance requirement for 
suppression pool temperature monitoring, (3) clarify the setpoint for 
the pressure suppression chamber-reactor building vacuum breakers 
instrumentation, (4) clarify the operating force requirements for the 
pressure suppression chamber-drywell vacuum breakers surveillance test, 
and (5) make corrections resulting from license Amendments 130 and 132.
    Date of issuance: January 28, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 141.
    Facility Operating License No. DPR-22. Amendment revised the TSs.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19573).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 28, 2005.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of application for amendment: October 5, 2004.
    Brief description of amendment: The amendment deletes technical 
specification (TS) 6.7.A.2, ``Requirement to submit an Occupational 
Radiation Exposure Report,'' TS 6.7.A.3, ``Requirement to submit a 
Monthly Operating Report,'' and TS 6.7.A.6, ``Requirement to report 
safety/relief valve failures and challenges'' as described in the 
Notice of Availability published in the Federal Register on June 23, 
2004 (69 FR 35067).
    Date of issuance: February 1, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 142.
    Facility Operating License No. DPR-22. Amendment revised the TSs.
    Date of initial notice in Federal Register: November 9, 2004 (69 FR 
64989).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 1, 2005.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant, 
Van Buren County, Michigan

    Date of application for amendment: October 5, 2004.
    Brief description of amendment: The amendment deletes technical 
specification 5.6.1, ``Occupational Radiation Exposure Report,'' and TS 
5.6.4 ``Monthly Operating Reports,'' as described in the Notice of 
Availability published in the Federal Register on June 23, 2004 (69 FR 
35067).
    Date of issuance: January 10, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 220.
    Facility Operating License No. DPR-20: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 9, 2004 (69 FR 
64989).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 10, 2005.
    No significant hazards consideration comments received: No.

[[Page 7774]]

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant, 
Van Buren County, Michigan

    Date of application for amendment: January 30, 2004.
    Brief description of amendment: The amendment eliminates 
requirements for hydrogen recombiners and relocates the requirements 
for hydrogen monitors to the licensee's Commitment Management Program.
    Date of issuance: January 11, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 221.
    Facility Operating License No. DPR-20: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 2, 2004 (69 FR 
9862).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 11, 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 5, 2004.
    Brief description of amendments: The amendments delete technical 
specification (TS) 5.6.1, ``Occupational Radiation Exposure Report,'' 
and TS 5.6.4 ``Monthly Operating Reports,'' as described in the Notice 
of Availability published in the Federal Register on June 23, 2004 (69 
FR 35067).
    Date of issuance: January 31, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 168, 158.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 2004 (69 FR 
64989).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 31, 2005.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of application for amendments: February 13, 2004, as 
supplemented by letters dated November 5 and December 10, 2004.
    Brief description of amendments: The amendments revise Technical 
Specifications (TSs) 3.3.1, ``Reactor Trip System (RTS) 
Instrumentation,'' 3.3.2, ``Engineered Safety Feature Actuation System 
(ESFAS) Instrumentation,'' and 3.3.6, ``Containment Ventilation 
Isolation Instrumentation,'' to adopt the completion time, test bypass 
time, and surveillance frequency time changes approved by the NRC in 
Topical Reports WCAP-14333-P-A, ``Probabilistic Risk Analysis of the 
RPS [reactor protection system] and ESFAS Test Times and Completion 
Times,'' and WCAP-15376-P-A, ``Risk-Informed Assessment of the RTS and 
ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and 
Completion Times.'' The amendments revise the required actions for 
certain action conditions; increase the completion times for several 
required actions (including some notes); delete notes in certain 
required actions; and increase frequency time intervals (including 
certain notes) in several surveillance requirements.
    Date of issuance: January 31, 2005.
    Effective date: January 31, 2005, and shall be implemented within 
180 days of the date of issuance.
    Amendment Nos.: Unit 1--179; Unit 2--181.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 30, 2004 (69 FR 
16622). The supplemental letters dated November 5 and December 10, 
2004, provided clarifying information that did not change the scope of 
the original application as noticed or the NRC staff's original 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 31, 2005.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch Nuclear 
Plant, Unit 2, Appling County, Georgia

    Date of application for amendments: April 26, 2004, as supplemented 
by letters dated August 17 and September 7, 2004.
    Brief description of amendments: The amendment revised the 
Technical Specification Section 5.5.12, ``Primary Containment Leakage 
Rate Testing Program'' to reflect a one-time deferral of the Type A 
Containment Integrated Leak Rate Test (ILRT). This change extends the 
10-year interval between ILRTs to 15 years from the previous ILRT.
    Date of issuance: February 1, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 187.
    Renewed Facility Operating License No. NPF-5: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 3, 2004 (69 FR 
46591).
    The supplements dated August 17 and September 7, 2004, 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 February 1, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 17, 2003, as 
supplemented by letters dated October 28 and November 16, 2004.
    Brief description of amendment: The amendment revises TSs 3.3.1, 
``Reactor Trip System (RTS) Instrumentation,'' 3.3.2, ``Engineered 
Safety Feature Actuation System (ESFAS) Instrumentation,'' and 3.3.9, 
``Boron Dilution Mitigation System (BDMS)'' to adopt the completion 
time, test bypass time, and surveillance time interval changes in NRC-
approved WCAP-14333-P-A, ``Probabilistic Risk Analysis of the RPS 
[reactor protection system] and ESFAS Test Times and Completion 
Times,'' and WCAP-15376-P-A, ``Risk-Informed Assessment of the RTS and 
ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and 
Completion Times.'' The TS changes revise required actions for certain 
action conditions; increase the completion times for several required 
actions (including some notes); delete notes in certain required 
actions; increase frequency time intervals (including certain notes) in 
several surveillance requirements (SRs); add an action condition and 
required actions; revise notes in certain SRs; and revise Table 3.3.2-
1. There is also an administrative correction to the format of the TSs.

[[Page 7775]]

    Date of issuance: January 31, 2005.
    Effective date: January 31, 2005, and shall be implemented within 
120 days of its date of issuance.
    Amendment No.: 165.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 3, 2004 (69 FR 
5211).
    The supplemental letters dated October 28 and November 16, 2004, 
provided clarifying information that did not change the scope of the 
original application as noticed or the NRC staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 31, 2005.
    No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: December 15, 2003, as supplemented by 
letters dated October 7 and November 12, 2004.
    Brief description of amendment: The amendment revises Technical 
Specifications (TSs) 3.3.1, ``Reactor Trip System (RTS) 
Instrumentation,'' and 3.3.2, ``Engineered Safety Feature Actuation 
System (ESFAS) Instrumentation,'' to adopt the completion time, test 
bypass time, and surveillance frequency time changes approved by the 
NRC in Topical Reports WCAP-14333-P-A, ``Probabilistic Risk Analysis of 
the RPS [reactor protection system] and ESFAS Test Times and Completion 
Times,'' and WCAP-15376-P-A, ``Risk-Informed Assessment of the RTS and 
ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and 
Completion Times.'' The amendment revises the required actions for 
certain action conditions; increase the completion times for several 
required actions (including some notes); delete notes in certain 
required actions; and increase frequency time intervals (including 
certain notes) in several surveillance requirements.
    Date of issuance: January 31, 2005.
    Effective date: January 31, 2005, and shall be implemented within 
180 days of the date of issuance.
    Amendment No.: 156.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 3, 2004 (69 FR 
5212).
    The supplemental letters dated October 7 and November 12, 2004, 
provided clarifying information that did not change the scope of the 
original application as noticed or the NRC staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 31, 2005.
    No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: October 7, 2004.
    Brief description of amendment: The amendment revises Section 5.3, 
``Unit Staff Qualifications,'' of the technical specifications (TSs) to 
add the qualification requirements for the shift manager and the 
control room supervisor. In addition, based on a comparison review 
performed by the NRC and Wolf Creek Nuclear Operating Corporation 
personnel, editorial corrections are being made to the TSs.
    Date of issuance: January 31, 2005.
    Effective date: January 31, 2005, and shall be implemented within 
90 days from the date of issuance.
    Amendment No.: 159.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 23, 2004 (68 
FR 68188).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 31, 2005.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant

[[Page 7776]]

hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 
application for amendment, (2) the amendment to Facility Operating 
License, 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 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 PDR Reference 
staff at 1 (800) 397-4209, (301) 415-4737 or by email to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC Web site, 
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR Reference staff at 
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. 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 requestor's/petitioner's interest. The petition must 
also identify 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 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
intends to rely to establish those facts or expert opinion. The 
petition must include sufficient information to show that a genuine 
dispute exists with the applicant on a material issue of law or 
fact.\1\ Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner to relief. A petitioner/requestor 
who fails to satisfy these requirements with respect to at least one 
contention will not be permitted to participate as a party.
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    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly available because they 
are asserted to contain safeguards or proprietary information, 
petitioners desiring access to this information should contact the 
applicant or applicant's counsel and discuss the need for a 
protective order.
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    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    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. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    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:

[[Page 7777]]

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, [email protected]; 
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 
[email protected]. 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 or 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).

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of amendment request: January 15, 2005.
    Description of amendment request: The amendment revises the 
Operating License to add a license condition to allow a one-time 
extension of the allowed outage time for the west centrifugal charging 
pump.
    Date of issuance: January 16, 2005.
    Effective date: January 16, 2005.
    Amendment No.: 285.
    Facility Operating License No. DPR-58: Amendment revises the 
Operating License.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): No. The Commission's related evaluation of the 
amendment, finding of emergency circumstances, state consultation, and 
final NSHC determination are contained in a safety evaluation dated 
January 16, 2005.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: M. Kotzalas, Acting.

    Dated at Rockville, Maryland, this 7th day of February, 2005.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 05-2788 Filed 2-14-05; 8:45 am]
BILLING CODE 7590-01-P