[Federal Register Volume 69, Number 196 (Tuesday, October 12, 2004)]
[Notices]
[Pages 60677-60692]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-22544]


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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 September 17, 2004, through September 30, 
2004. The last biweekly notice was published on September 28, 2004 (69 
FR 57978).

[[Page 60678]]

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 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

[[Page 60679]]

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 Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by email to 
[email protected].

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

    Date of amendment request: April 14, 2004.
    Description of amendment request: The proposed amendment would make 
changes to the Technical Specifications (TSs) that will eliminate 
secondary containment operability requirements when handling 
sufficiently decayed irradiated fuel and performing core alterations, 
and will clarify requirements associated with operations with potential 
to drain the reactor vessel. This proposed amendment also uses 
Alternate Source Term (AST) methodology in accordance with 10 CFR 50.67 
for calculating Fuel Handling Accident (FHA) consequences. The proposed 
amendment also removes TSs operability requirements for engineered 
safety features (ESF) (e.g. primary/secondary containment, standby gas 
treatment, and isolation capability) after the sufficient decay of 
``recently'' irradiated fuel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The Nuclear Regulatory Commission (NRC) staff has 
reviewed the licensee's analysis against the standards of 10 CFR 
50.92(c). The NRC staff's review is presented below.
1. Do the Proposed Changes Involve a Significant Increase in the 
Probability or Consequence of an Accident Previously Evaluated?
    The proposed changes do not modify the design or operation of 
equipment used to handle and move new and spent fuel or to perform core 
alterations. The proposed amendment does not modify the design of the 
ESF equipment. The proposed changes, therefore, will not increase the 
probability of accidents previously evaluated.
    AST analysis does not affect the performance of the systems or 
components used to mitigate the consequences of accidents previously 
evaluated. While a direct comparison between current methodologies used 
in the current Pilgrim design basis analysis and Regulatory Guide (RG) 
1.183 is not possible due to different acceptance criteria, the AST 
calculations demonstrate that the radiological consequences to the 
accidents previously evaluated will still remain below the regulatory 
limits. Therefore, any potential change in the radiological 
consequences are not considered significant. Since the radiological 
consequences are below the regulatory limits and the probability of an 
accident is unchanged, the proposed changes do not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
2. Do the Proposed Changes Create the Possibility of a New or Different 
Kind of Accident From Any Accident Previously Analyzed?
    There are no new plant operation modes or physical modifications 
being proposed. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any previously 
analyzed.
3. Do the Proposed Changes Involve a Significant Reduction in the 
Margin of Safety?
    The licensee performed a comprehensive analysis and evaluation of 
the FHA using AST methodology and dose consequence analysis in 
accordance with 10 CFR 50.67. While direct comparison between 
methodologies used in the current Pilgrim design basis analysis and RG 
1.183 is not possible due to different acceptance criteria, the revised 
doses will, however, remain below the total effective dose equivalent 
dose regulatory limits for the control room, exclusion area boundary, 
and low population zone as specified in 10 CFR 50.67. Therefore, by 
meeting the applicatory regulatory limits for AST, any potential 
decrease in a margin of safety would not be considered significant. The 
changes are, therefore, not considered a significant reduction in a 
margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts, 02360-5599.
    NRC Acting Section Chief: Daniel S. Collins.

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

    Date of amendment request: June 17, 2004.
    Description of amendment request: The proposed amendment would 
delete

[[Page 60680]]

entries from Technical Specification (TS) Tables 3.2.6 and 4.2.6 
related to the post-accident hydrogen and oxygen 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, ``Combustible gas control system for 
nuclear power reactors,'' eliminated the requirements for hydrogen 
recombiners (not installed at Vermont Yankee and therefore not 
addressed by this proposed amendment) 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 June 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:

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 and oxygen 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. RG 1.97 Category 1, is intended for key variables that 
most directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen and oxygen 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. Also, as part 
of the rulemaking to revise 10 CFR 50.44, the Commission found that 
Category 2, as defined in RG 1.97, is an appropriate categorization 
for the oxygen monitors, because the monitors are required to verify 
the status of the inert containment.
    The regulatory requirements for the hydrogen and oxygen 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, 
classification of the oxygen monitors as Category 2, and removal of 
the hydrogen and oxygen monitors from TSs 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 relaxation of the hydrogen and oxygen monitor 
requirements, including removal of these requirements from TSs, 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 relaxation of the hydrogen and oxygen monitor requirements, 
including removal of these requirements from TSs, will not result in 
any failure mode not previously analyzed. The hydrogen and oxygen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen and oxygen 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 relaxation of the hydrogen and oxygen monitor requirements, 
including removal of these requirements from TSs, 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.

Category 2 Oxygen Monitors Are Adequate To Verify the Status of an 
Inerted Containment.

    Therefore, this change does not involve a significant reduction 
in the margin of safety. The intent of the requirements established 
as a result of the TMI, Unit 2 accident can be adequately met 
without reliance on safety-related oxygen monitors. Removal of 
hydrogen and oxygen monitoring from TSs will not result in a 
significant reduction in their functionality, reliability, and 
availability.

    Based on the reasoning presented above and the previous discussion 
of the amendment request, the requested change does not involve a 
significant hazards consideration.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: September 1, 2004.
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) 6.6.A, ``Occupational Radiation 
Exposure Report,'' and TS 6.6.B, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant

[[Page 60681]]

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 September 1, 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 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: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: Allen G. Howe.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
and Lancaster Counties, Pennsylvania

    Date of amendment request: April 30, 2004.
    Description of amendment request: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
technical specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of Title 10 of the Code of Federal 
Regulations (10 CFR), Part 50, Section 50.65(a)(4). Limiting Condition 
for Operation (LCO) 3.0.4 exceptions in individual TSs would be 
eliminated, several notes or specific exceptions are revised to reflect 
the related changes to LCO 3.0.4, and Surveillance Requirement (SR) 
3.0.4 is revised to reflect the LCO 3.0.4 allowance.

    This change was proposed by the industry's Technical 
Specification Task Force (TSTF) and is designated TSTF-359. The NRC 
staff issued a notice of opportunity for comment in the Federal 
Register on August 2, 2002 (67 FR 50475), on possible amendments 
concerning TSTF-359, including a model safety evaluation and model 
no significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal 
Register on April 4, 2003 (68 FR 16579). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated April 30, 2004.

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

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

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

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

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

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

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

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for Licensee: Thomas S. O'Neill, Associate and General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Daniel Collins, Acting.

[[Page 60682]]

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

    Date of amendment request: August 2, 2004.
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) Section 6.8.4, ``Post Accident 
Sampling,'' and the related requirements to maintain a Post-Accident 
Sampling System (PASS). Licensees were generally required to implement 
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three 
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97, 
Revision 3, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Access Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades was an outcome of the U.S. 
Nuclear Regulatory Commission's (NRC) lessons learned from the accident 
that occurred at TMI Unit 2. Requirements related to PASS were imposed 
by Order for many facilities and were added to or included in the TS 
for nuclear power reactors currently licensed to operate. Lessons 
learned and improvements implemented over the last 20 years have shown 
that the information obtained from PASS can be readily obtained through 
other means or is of little use in the assessment and mitigation of 
accident conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on March 3, 2003 (68 FR 10052) on possible amendments 
to eliminate PASS, including a model safety evaluation and model no 
significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in a 
license amendment application in the Federal Register on May 13, 2003 
(68 FR 25664). The licensee affirmed the applicability of the following 
NSHC determination in its application dated August 2, 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 PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
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. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the 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 PASS requirements from Technical 
Specification (TS) (and other elements of the licensing bases) does 
not involve a significant increase in 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 PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radioisotopes 
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 PASS, 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. Methodologies that are 
not reliant on PASS are designed 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 use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    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: Anthony J. Mendiola.

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

    Date of amendment requests: April 13, 2004.
    Description of amendment requests: The proposed amendments would 
change the licensing basis as described in the Updated Final Safety 
Analysis Report to allow the use of a reinforcing bar (rebar) yield 
strength value based on measured material properties, as documented in 
the licensee rebar acceptance tests, in control rod drive missile 
shield structural calculations.
    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 of occurrence or consequences of an accident 
previously evaluated?

[[Page 60683]]

    Response: No

Probability of Occurrence of an Accident Previously Evaluated

    This is a change in the method of determining the acceptability 
of accommodating the pressure load following a loss-of-coolant 
accident. No physical changes are being made to the plant and no 
potential accident initiators are introduced by this change. Thus, 
the probability of the occurrence of any accident previously 
evaluated is not significantly increased.

Consequences of an Accident Previously Evaluated

    There is reasonable assurance that the ability of control rod 
drive missile shields (missile shields) to maintain their structural 
capability and continue to function as a part of the divider barrier 
separating the lower containment from the upper containment is not 
impacted by this change. The data obtained from rebar acceptance 
test reports demonstrate that the missile shields have adequate 
strength to accommodate the load that would be imposed under assumed 
accident conditions. As a result, the consequences of any accident 
previously evaluated are not significantly increased.
    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 use of increased missile shield rebar yield strength for the 
missile shield structural capability under accident conditions does 
not alter the evaluation of the missile shields' structural 
capability during normal operation, the operational condition in 
which a new or different kind of accident would be initiated. The 
change does not physically alter plant components nor does it alter 
plant operation. The change does not adversely affect current system 
interfaces or create new interfaces that could result in an accident 
or malfunction of a different kind than previously evaluated.
    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 margin of safety for the missile shields is provided by the 
factors that are applied to the individual loads determining the 
load imposed on the missile shields under accident conditions. These 
code safety factors are sufficient to ensure that both anticipated 
and unanticipated loads can be withstood by the concrete structures. 
The use of yield strengths based on measured material properties as 
documented in the I&M [Indiana Michigan Power Company] rebar 
acceptance tests for the missile shield structural evaluation has no 
effect on the margin of safety provided by the load safety factors. 
I&M continues to use the same load factors that were used to license 
the Donald C. Cook Nuclear Plant.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: September 7, 2004.
    Description of amendment request: The proposed amendment revises 
Fort Calhoun Station (FCS) Technical Specification (TS) 5.9.5, ``Core 
Operating Limits Report,'' such that it will read consistent with TS 
5.6.5 of NUREG-1432, Standard Technical Specifications-Combustion 
Engineering Plants. In addition, the list of core reload analysis 
methodologies contained in TS 5.9.5b used to determine the core 
operating limits is updated to move many of these references to Omaha 
Public Power District (OPPD) core reload analysis methodology documents 
OPPD-NA-8301, 8302, and 8303. Several analytical method references that 
are no longer applicable to FCS are deleted from TS 5.9.5b; several 
references will remain, as they are not suitable for incorporation into 
the core reload analysis documents.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed changes are primarily administrative in nature to 
achieve consistency with Standard Technical Specifications and to 
update the list of NRC reviewed and approved analytical methods used 
to develop core operating limits. Several of the analytical methods 
are no longer applicable to Fort Calhoun Station, Unit No. 1, (FCS) 
and thus are deleted from the Technical Specifications (TSs). Many 
of the topical reports currently referenced in TS 5.9.5b are more 
suitably referenced in the OPPD core reload methodology documents 
where they have been relocated.
    The OPPD core reload methodology documents remain referenced in 
TS 5.9.5b and as such are subject to NRC review and approval. The 
relocation of the topical reports referenced in TS 5.9.5b to OPPD 
core reload methodology documents is an administrative change. In 
addition to the incorporation of references currently found in TS 
5.9.5b, OPPD core reload methodology documents OPPD-NA-8301, 8302, 
and 8303 are revised to remove characters designating them as 
proprietary, and approved. This is an administrative change, as OPPD 
no longer considers the documents to be proprietary or topical 
reports. OPPD core reload methodology documents OPPD-NA-8301, 8302, 
and 8303 are enclosed for NRC review and approval [attached to the 
licensee's September 7, 2004, letter] of the changes noted above and 
incorporation of the CASMO-4 (C-4) computer code, which is described 
below.
    OPPD is adding the C-4 code to OPPD-NA-8302, Reload Core 
Analysis Methodology, Neutronics Design Methods and Verification and 
will use the code for nuclear design analysis. This will allow the 
use of the C-4 and SIMULATE-3 (S-3) methodology to perform all 
steady-state pressurized water reactor (PWR) core physics analyses. 
The probability of occurrence of an accident previously evaluated 
will not be increased by the proposed change in the particular codes 
used for physics calculations for nuclear design analysis. The 
results of nuclear design analyses are used as inputs to the 
analysis of accidents that are evaluated in the Updated Safety 
Analysis Report (USAR). These inputs do not alter the physical 
characteristics or modes of operation of any system, structure, or 
component involved in the initiation of an accident. Thus, there is 
no significant increase in the probability of an accident previously 
evaluated as a result of this change.
    The consequences of an accident evaluated in the USAR are 
affected by the value of inputs to the transient safety analysis. An 
extensive benchmark of C-4/S-3 predictions was performed with 
measured data using a variety of fuel designs and operating 
conditions in power reactors and critical experiments. The accuracy 
of C-4/S-3 is similar to, and sometimes better than, the accuracy of 
C-3/S-3. Furthermore, there is always the potential for the value of 
the nuclear design parameters to change solely as a result of the 
new core reload fuel core loading pattern. Regardless of the source 
of a change, an assessment is always made of changes to the nuclear 
design parameters with respect to their effects on the consequences 
of accidents previously evaluated in the USAR. Refueling is an 
anticipated activity, which is described in the USAR. If increased 
consequences are anticipated, compensatory actions are implemented 
to neutralize any expected increase in consequences. These 
compensatory actions include, but are not limited to, crediting any 
existing margins in the analysis or redefining the operating 
envelope to avoid increased consequences. Thus, the nuclear design 
parameters are intermediate results and by themselves will not 
result in an increase in the consequence of an accident evaluated in 
the USAR.

[[Page 60684]]

    Therefore, the use of the C-4/S-3 code package, which will 
perform the same functions as the C-3/S-3 codes with similar 
accuracy, does not significantly increase the consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed changes are primarily administrative in nature. The 
changes achieve consistency with Standard Technical Specifications, 
update the list of NRC reviewed and approved analytical methods used 
to develop core operating limits by deleting certain analytical 
methods no longer applicable to FCS and relocating many of the 
remainder to OPPD core reload analysis methodology documents, and 
make minor administrative changes to OPPD core reload analysis 
documents referenced in TS 5.9.5b. OPPD intends to utilize the C-4/
S-3 code package for nuclear design analysis. The proposed amendment 
would add the C-4 code to OPPD core reload analysis methodology 
document OPPD-NA-8302.
    The possibility for a new or different kind of accident 
evaluated previously in the USAR will not be created by the proposed 
administrative changes or the change to the particular codes used 
for physics calculations for nuclear design analyses. The change 
involves adding the Studsvik C-4 code to OPPD core reload analysis 
methodology document OPPD-NA-8302. The C-4 code is an update to the 
C-3 code currently approved for use at FCS. The results of nuclear 
design analyses are used as inputs to the analysis of accidents that 
are evaluated in the USAR. These inputs do not alter the physical 
characteristics or modes of operation of any system, structure or 
component involved in the initiation of an accident. Therefore, 
these administrative changes and the addition of the C-4 code, which 
will perform the same functions, as the C-3 code with similar 
accuracy, does not increase the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not involve a significant reduction in 
a margin of safety. The margin of safety as defined in the basis for 
any technical specification will not be reduced nor increased by the 
proposed administrative changes or the change to the codes used for 
physics calculations for nuclear design analyses. The changes 
achieve consistency with Standard Technical Specifications, update 
the list of NRC approved analytical methods used to develop core 
operating limits by deleting certain analytical methods no longer 
applicable to FCS and relocating many of the remainder to OPPD core 
reload analysis methodology documents, and make minor administrative 
changes to OPPD core reload analysis documents referenced in TS 
5.9.5b.
    The change involves the addition of the Studsvik C-4 code to 
OPPD core reload analysis methodologies for nuclear design analysis. 
Extensive benchmarking of the C-4/S-3 computer codes has 
demonstrated that the values of those parameters used in the safety 
analysis are not significantly changed relative to the values 
obtained using the NRC approved C-3/S-3 computer codes. For any 
changes in the calculated values that do occur, the application of 
appropriate biases and uncertainties ensures that the current margin 
of safety is maintained. Specifically, use of these code specific 
biases and uncertainties in safety evaluations continues to provide 
the same statistical assurance that the values of the nuclear 
parameters used in the safety analysis are conservative with respect 
to the actual values on at least a 95/95 probability/confidence 
basis.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert Gramm.

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

    Date of amendment request: April 15, 2004, as supplemented August 
11, 2004.
    Description of amendment request: The amendment request proposes to 
revise the Salem Unit No. 1 Technical Specifications (TSs) to reflect 
the addition of the chilled water system to provide cooling water to 
the containment fan cooling units (CFCUs). The amendment request also 
proposes to revise a non-conservative Action Statement for Salem Unit 
Nos. 1 and 2 that allows three containment cooling fans to be 
inoperable under certain conditions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    Response: No.
    Containment cooling fans remove containment heat loads under 
both normal and accident conditions. As such, they have no impact on 
the probability of occurrence of any previously evaluated accidents, 
although they do function to mitigate accident consequences. With 
regard to accident consequences, revised containment response 
analysis has been performed with the proposed changes of this 
license amendment. This analysis demonstrates that containment 
pressure and temperature limits continue to be met as further 
described below.
    The addition of the non-safety related chilled water system does 
not represent an increase in the consequences of an accident since, 
at the onset of the accident, the chilled water supply is 
automatically isolated on the resulting safety injection signal and 
the safety related Service Water System supplies the cooling method 
to remove the containment heat loads, as presently analyzed. 
Analysis has been performed to evaluate any potential failures that 
could prevent the Containment Cooling System to perform [sic] its 
safety related functions. Redundancy in the chilled water system and 
transfer to service water during an accident are incorporated in the 
design. In addition, as a conservative measure, an action statement 
has been added to require prompt action to restore containment 
cooling or commence a unit shutdown in the event of an unexpected 
condition that results in the loss of normal containment cooling 
capability.
    The accidents previously evaluated that are associated with 
containment heat removal are design basis loss-of-coolant accident 
(LOCA) and main steam line break (MSLB) accident. In the case of the 
design basis LOCA, the revised analysis demonstrates that all cases 
resulted in a peak containment pressure that was less than 47 psig. 
In addition, all long-term cases were well below 50% of the peak 
value within 24 hours. Based on the results, applicable criteria for 
Salem Unit 1 have been met and therefore, the consequences of 
previously evaluated accidents are not increased.
    The proposed change to the non-conservative TS 3.6.2.3 Action b, 
maintains that five CFCUs remain operable to ensure that, upon a 
single failure, a minimum of three CFCUs will provide the required 
containment and air mixing which is consistent with the current 
Salem Dose Analysis.
    Consequently, the proposed license amendment does not increase 
the probability of occurrence or the consequences of accidents 
previously evaluated for Salem.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Response: No.
    Containment cooling fans remove containment heat loads under 
both normal and accident conditions. The containment cooling fans 
are presently part of the plant protection equipment and have been 
analyzed and evaluated as to their function and effectiveness. 
Consequently, they cannot create the possibility of any new or 
different kinds of accidents from any previously evaluated. The 
addition of a chilled water system that is isolated on an accident 
condition does not create a new or different kind of accident. The 
accidents analyzed are the LOCA and MSLB, which are part of the 
Salem Design Bases.
    The proposed change to the non-conservative TS 3.6.2.3 Action b, 
maintains that five CFCUs remain operable to ensure that, upon a 
single failure, a minimum of

[[Page 60685]]

three CFCUs will provide the required containment and air mixing 
which is consistent with the current Salem Dose Analysis.
    Therefore, the proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The margin of safety pertinent to the proposed changes is the 
dose consequences resulting from a design basis LOCA. Containment 
cooling fans affect potential dose consequences in that they assist 
in maintaining containment pressure and temperature within design 
limits. By maintaining these limits, three critical functions are 
performed. These are:
    a. Containment integrity is assured by maintaining pressure 
below the containment design limit.
    b. By maintaining pressure below 47 psig, leakage of containment 
atmosphere to the surrounding environment is retained within the 
leakage testing results of 10 CFR 50, Appendix J. In this case, the 
Appendix J testing procedures provide the margin of safety, as long 
as the limiting pressure (47 psig) is not exceeded.
    c. By maintaining containment temperature within limits, the 
qualification of vital electrical equipment to function in the post-
accident containment environment is assured. In this case, the 
margin of safety is provided by the testing and evaluation 
procedures implemented by 10 CFR 50.49.
    In addition, as a conservative measure, an action statement has 
been added to require prompt action to restore containment cooling 
or commence a unit shutdown in the event of an unexpected condition 
that results in the loss of normal containment cooling capability.
    The proposed change to the non-conservative TS 3.6.2.3 Action b, 
maintains that five CFCUs remain operable to ensure that, upon a 
single failure, a minimum of three CFCUs will provide the required 
containment and air mixing which is consistent with the current 
Salem Dose Analysis.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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: July 26, 2004.
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) 5.6.1, ``Occupational Radiation 
Exposure Report,'' and TS 5.6.4, ``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 July 26, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

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

    The proposed change does not 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.

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

    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 NRC staff proposes to 
determine that the requested change does not involve significance 
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.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station (VCSNS), 
Unit No. 1, Fairfield County, South Carolina

    Date of amendment request: August 24, 2004.
    Description of amendment request: The proposed change will revise 
Surveillance Requirements (SRs) 4.7. 1.2.a.1 and 4.7 .1.2.a.2 to 
reflect a more representative model of the Emergency Feedwater (EFW) 
System. The new model has established new technical specification (TS) 
acceptance criteria to assure the design requirements of the system are 
met. These required characteristics are more stringent than those 
currently in the VCSNS TSs for this system.
    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?
    This change represents a more restrictive surveillance 
requirement than currently exists for TS Surveillance 4.7. 1.2.a.1 
and 4.7 .1.2.a.2. These proposed surveillance acceptance criteria 
changes will ensure that the motor driven EFW pumps and the turbine 
driven EFW pump can continue to perform their design function. There 
are no changes planned to any plant installed hardware or software 
and normal plant operations will not be impacted.
    The probability or consequences of accidents previously 
evaluated in the VCSNS FSAR [Final Safety Analysis Report] are 
unaffected by this proposed change because there is no change to any 
equipment response or accident mitigation scenario. There are no 
additional challenges to fission product barrier integrity. 
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? 
The proposed change involves the revision of the Surveillance 
Requirements for the EFW system. The revised requirements are more 
restrictive to insure compliance with the design basis of the 
system. Changes to the system model require changes to the SR 
acceptance criteria in order to maintain the performance level 
assumed in the safety analysis.

[[Page 60686]]

    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed change. 
The proposed change does not challenge the performance or integrity 
of any safety-related system. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Does this change involve a significant reduction in margin of 
Safety?
    The proposed change will have no affect on the availability, 
operability, or performance of the safety-related systems and 
components. A change to the SR is proposed, however, the proposed 
change is more restrictive than the current SR. The more restrictive 
criteria inherently include a 5 gpm leak tolerance for the EFW flow 
control valves. This represents a built in margin for the pump head 
requirement when the flow control valve leakage is determined to be 
less than 5 gpm. Therefore, the proposed change does not involve a 
significant reduction in margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three 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: Thomas G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: Mary Jane Ross-Lee, Acting.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-321 
and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling 
County, Georgia; Docket Nos. 50-348 and 50-364, Joseph M. Farley 
Nuclear Plant, Units 1 and 2, Houston County, Alabama; Docket Nos. 50-
424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of amendment request: July 28, 2004.
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) 5.6.1, ``Occupational Radiation 
Exposure Report,'' and TS 5.6.4, ``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 July 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 proposed change eliminates the TS reporting requirements to 
provide a monthly operating report of shutdown experience and 
operating statistics if the equivalent data is submitted using an 
industry electronic database. It also eliminates the 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.

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

    The proposed change does not 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.

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

    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 a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: Mary Jane Ross-Lee, Acting.

Southern Nuclear Operating Company (SNC), Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of amendment request: July 20, 2004.
    Description of amendment request: The proposed amendments would 
make various changes to the technical specifications (TSs) associated 
with the Plant Hatch DC electrical system consistent with Technical 
Specification Task Force (TSTF) change traveler TSTF-360, including 
specific action and increased completion time for an inoperable battery 
charger, increase the completion time for an inoperable station service 
battery from 2 to 12 hours, relocate preventive maintenance 
surveillance requirements (SRs) to licensee controlled programs, 
provide alternate testing criteria for battery charger testing, replace 
battery specific gravity monitoring with float current monitoring, 
relocate and create a Section 5.5 program to reference actions for cell 
voltage and electrolyte level, and provide specific actions and 
increased completion times for out-of-limits conditions for certain 
battery parameters.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This is a proposed change to the DC system Technical 
Specifications. No physical changes are being proposed to any system 
designed to prevent a previously evaluated accident, such as a loss 
of coolant accident with a loss of offsite power.
    This proposed TS change provides specific completion times for 
certain inoperable DC components, and relocates some surveillance 
requirements to owner controlled programs. Additionally, monitoring 
of specific gravity will be replaced with float current monitoring, 
and some Action levels for cell voltage and electrolyte level are 
relocated to owner controlled programs.
    The completion time for battery charger inoperability is 
increased to 7 days; however, only after verification that the 
associated battery is fully operable, without such verification, the 
7 day completion time is not used. Thus, adequate DC to support 
design basis events is ensured.
    Increasing the station service battery out of service time from 
2 to 12 hours will allow more time for proper maintenance to repair 
a faulty battery. However, the 12 hour out of service time is still 
a very restrictive time and so the probability of an event where the 
battery would be needed within this 12 hour time frame is very low. 
In fact, a probability risk assessment of the increased out of 
service time has been performed and it fell within the criteria of 
Reg Guide 1.174 and 1.177.
    The relocation of certain SRs and action levels is done for 
surveillances and parameter action levels that are more intended to 
monitor and maintain long term

[[Page 60687]]

component performance. These relocated items are not meant as clear 
levels at which the DC components can no longer be considered 
operable. Those that are remain in the TS. Additionally, this 
particular owner controlled program will be referenced in proposed 
Section 5.5.13 of the TS. This commitment to the program will insure 
that the DC system will continue to be adequately monitored and 
maintained.
    Therefore, this proposed change to the TS ensures that the DC 
system will be able to provide its safety function. The probability 
and consequences of a previously evaluated accident are thus not 
increased.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    None of the DC components will be physically altered. 
Furthermore, the design bases for the DC distribution systems, 
batteries and chargers is not changing. Although some surveillance 
requirements are being relocated and one (specific gravity 
monitoring) is being eliminated, DC system components will still be 
adequately surveilled and maintained. Therefore, no, new modes of 
operation or failure are introduced by the proposed TS change and 
therefore, the possibility of a new type event is not created.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The design functions of the DC system are unchanged. The 
proposed TS changes relocate many surveillance requirements and 
action levels to owner controlled programs. However, the owner 
controlled program is referenced in the new proposed Section 5.5.13 
of the TS. The SNC commitment to this program will continue to 
ensure that the DC system is adequately monitored, surveilled, and 
maintained to insure that it can perform its safety function when 
called upon.
    The addition of a 7 day completion time for the battery chargers 
can be used only if adequate battery capacity is verified. Thus, the 
DC system is capable of performing its safety function throughout 
the 7 day completion time.
    Increasing the allowed out of service time for the station 
service batteries does not result in a significant reduction in the 
margin of safety since the proposed 12 hour time limit is still a 
very short time. Probabilistic risk analysis shows that the core 
damage frequency and large early release fractions are within the 
guidelines of Reg Guides 1.174 and 1.177.
    Elimination of specific gravity surveillance is acceptable since 
the float current monitoring adequately replaces it.
    For the above 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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: Mary Jane Ross-Lee, Acting.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

    Date of amendment request: July 2, 2004 (TS-449).
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) 5.6.1, ``Occupational Radiation 
Exposure Report,'' and TS 5.6.4, ``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 July 2, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

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

    The proposed change does not 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.

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

    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in the margin of safety.
    Based upon the reasoning presented above, the requested change 
does not involve significance hazards consideration.

    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Michael L. Marshall (Acting).

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: August 12, 2004.
    Description of amendment request: The proposed Technical 
Specification change will revise Surveillance Requirement 4.7.8.d.3 by 
removing the vacuum relief flow portion. The proposed revision removes 
criteria from the surveillance that is not necessary to verify the 
operability of the Auxiliary Building Gas Treatment System (ABGTS). The 
bases associated with the ABGTS will be revised to remove discussions 
regarding the vacuum relief flow portion of this surveillance as part 
of this effort.
    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?
    No. The proposed change removes an overly restrictive criterion 
for vacuum relief flow as part of the ABGTS operability 
verification. This criterion is not required for the verification of 
ABGTS operability and therefore, the removal does not reduce the 
associated safety function. No system modification or operating 
practices are changed by the proposed revision. The accident 
mitigation functions of the ABGTS will not be adversely affected by 
the proposed removal and offsite dose potential is not increased. 
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?

[[Page 60688]]

    No. The proposed change does not result in the alteration of 
plant equipment or components or the modification of operating 
requirements for plant systems. Additionally, the ABGTS functions 
serve to mitigate accident conditions and are not considered a 
source for accident generation. 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?
    No. The proposed removal of an unnecessary criterion from the 
ABGTS surveillance will not result in a change to plant setpoints 
that function to maintain the safety margins. The ABGTS will 
continue to provide the required negative pressure conditions for 
the auxiliary building during accident conditions to maintain 
acceptable dose conditions. The actuation of safety features for 
accident mitigation will not be affected by the proposed changes. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The United States Nuclear Regulatory Commission (NRC) staff has 
reviewed the licensee's analysis and, based on this review, it appears 
that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, 
the NRC staff proposes to determine that the amendment request involves 
no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Acting Section Chief: Michael Marshall, Jr.

Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328, 
Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: August 18, 2004.
    Description of amendment request: The proposed amendment would 
rename the Trip Setpoint column of Technical Specification (TS) Tables 
2.2-1 and 3.3-4, remove the inequality signs for the trip setpoint 
values as appropriate, and revise the inequality representation for the 
allowable values, as needed. This proposed amendment is a revision to a 
previous amendment request dated November 15, 2002 (ADAMS Accession No. 
ML023290477), that supersedes the original request in its entirety.
    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:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed revisions for the nominal trip setpoint 
representation are administrative changes that will not impact the 
application of the reactor trip or ESF [engineered safety feature] 
actuation system instrumentation requirements. This is based on the 
setpoint requirements being applied without change, as well as the 
Avs [allowable values], in accordance with the current setpoint 
methodology. The removal of the inequalities associated with the 
trip setpoint values will be more appropriate for the use of nominal 
setpoint values but will not differ in application from the setpoint 
methodology utilized by TVA. Deletion of the nominal terminology 
associated with overtemperature delta temperature average 
temperature at rated thermal power (T') provides a better 
representation of the limit associated with this value. In addition, 
this change will not alter plant equipment or operating practices. 
Therefore, the implementation of these changes will not increase the 
probability or consequences of an accident.
    The revision of the reactor coolant pump (RCP) underfrequency, 
intermediate range neutron flux P-6, and fuel storage pool area 
radiation monitor trip setpoints and the Avs for the RCP 
underfrequency, intermediate range neutron flux P-6, and 
undervoltage has been evaluated and the results are documented in 
approved calculations. These calculations verify that the revised 
values are acceptable in accordance with appropriate calculation 
methodologies and that they will continue to support the accident 
analysis. These revisions will not require changes to the 
instrumentation settings currently being used or the methods for 
maintaining them. The offsite dose potential will be reduced because 
the proposed TS values are more conservative and will ensure the 
adequacy of designed safety functions to limit the release of 
radioactivity. Therefore, the proposed revision of these values will 
not significantly increase the probability or consequences of an 
accident.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The revision of the nominal trip setpoint representation and 
elimination of the nominal nomenclature, as well as the revised 
setpoint values and Avs will not alter the plant configuration or 
functions. The revised setpoints and the proposed operability limits 
will continue to provide acceptable initiation of safety functions 
for the mitigation of postulated accidents as required by the design 
basis. The primary function of the reactor protection system, the 
ESF actuation system, and the radiation monitoring function is to 
initiate accident mitigation functions. These functions are not 
considered to be initiators of postulated accidents. The proposed 
changes do not create the possibility of a new or different kind of 
accident because the design functions are not altered and the 
proposed values meet the accident analysis requirements for accident 
mitigation.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The setpoint and Av revisions proposed in this request were 
evaluated and found to be acceptable without impact to the safety 
limits required for the associated functions. The nominal trip 
setpoint representation change and the elimination of inappropriate 
nominal indications do not alter the TS functions or their 
application and will not require changes to design settings. Plant 
systems will continue to be actuated for those plant conditions that 
require the initiation of accident mitigation functions. The margin 
of safety is not reduced because the proposed conservative changes 
to the Av and setpoint representations will not change design 
functions and the initiation of accident mitigation functions for 
appropriate plant conditions is ensured.

    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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Michael L. Marshall, Jr. (Acting).

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has

[[Page 60689]]

made a determination based on that assessment, it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by email to [email protected].

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of application for amendments: August 22, 2003.
    Brief description of amendments: The amendments authorize revision 
of the Updated Final Safety Analysis Report to incorporate the 
description of the approved change to the maximum fuel pin 
pressurization criteria used in the evaluation of the design basis 
fuel-handling accident as described in the amendment application of 
August 22, 2003.
    Date of issuance: September 27, 2004.
    Effective date: September 27, 2004, and shall be implemented within 
60 days of the date of issuance.
    Amendment Nos.: Unit 1-153, Unit 2-153, Unit 3-153.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments authorize the revision of the Updated Final Safety Analysis 
Report.
    Date of initial notice in Federal Register: December 9, 2003 (68 FR 
68656). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 27, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 10, 2002, as supplemented 
March 12, 2003, April 10, 2003, March 5, 2004, and July 22, 2004.
    Brief description of amendment: The amendment approves full 
implementation of the alternative source term, with the exception of 
the loss-of-coolant accident.
    Date of issuance: September 24, 2004.
    Effective date: September 24, 2004.
    Amendment No.: 201.
    Renewed Facility Operating License No. DPR-23: Amendment revises 
the Technical Specifications.
    Date of initial notice in Federal Register: April 1, 2003 (68 FR 
15758). The April 10, 2003, March 5, 2004, and July 22, 2004, 
supplements 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 amendment is contained 
in a Safety Evaluation dated September 24, 2004.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: September 26, 2002, as 
supplemented June 2, 2003, May 7, June 18, and August 24, 2004.
    Brief description of amendment: The amendment revised the technical 
specifications (TSs) to allow relaxation of containment operability 
requirements while handling irradiated fuel and core alterations.
    Date of issuance: September 20, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 284.
    Facility Operating License No. DPR-65: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68731). The supplements dated June 2, 2003, May 7, June 18, and 
August 24, 2004 contained clarifying information and did not change the 
staff's proposed finding of no significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 20, 2004.
    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: May 9, 2002, as supplemented by 
letter dated June 11, 2003, and August 18 and September 22, 2004.
    Brief description of amendments: The amendments approve changes to 
the Updated Final Safety Analysis Report for Catawba, Units 1 and 2 to 
eliminate the single failure of either of the 125 VDC Distribution 
Centers, EDE or EDF, from the design-basis steam generator tube rupture 
accident analyses.
    Date of issuance: September 24, 2004.
    Effective date: As of the date of issuance and shall be implemented 
with the next update of the Safety Analysis Report in accordance with 
10 CFR 50.71(e)
    Amendment Nos.: 217, 211.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revise the Licensing Basis.
    Date of initial notice in Federal Register: July 23, 2002 (67 FR 
48215). The supplements dated June 11, 2003, and August 18 and 
September 22, 2004, provided clarifying information that did not change 
the scope of the May 9, 2002, application nor the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 24, 2004.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of application for amendment: June 30, 2003, as supplemented 
by letters dated November 20, 2003, February 27, 2004, and September 
10, 2004.
    Brief description of amendment: The amendment (1) reorganizes the 
Arkansas Nuclear One, Unit No. 2 (ANO-2) Technical Specification (TS) 
Section 6.0, Administrative Controls, (2) modifies the ANO-2 Facility 
Operating License, and Actions and Surveillance Requirements (SRs) of 
various other TSs, to support the reorganization of Section 6.0, and 
(3) modifies several Actions and SRs that are related to systems that 
are shared by ANO-2 and Arkansas Nuclear One, Unit No. 1.
    Date of issuance: September 28, 2004.
    Effective date: As of the date of issuance to be implemented within 
120 days from the date of issuance.

[[Page 60690]]

    Amendment No.: 255.
    Facility Operating License No. NPF-6: Amendment revises the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 9, 2003 (68 FR 
68663). The supplements dated February 27, 2004, and September 10, 
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 amendment is contained 
in a Safety Evaluation dated September 28, 2004.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant, 
Unit No. 1, St. Lucie County, Florida

    Date of application for amendment: November 25, 2002, as 
supplemented by letters dated May 14, 2003, September 29, 2003, and 
March 25, 2004.
    Brief description of amendment: The proposed amendment would revise 
Technical Specification 3.9.11, ``Storage Pool Water Level'' and TS 
5.6.1, ``Fuel Storage--Criticality.'' This amendment permits St. Lucie 
Unit 1 to credit soluble boron, fuel loading restrictions, and control 
element assemblies in the spent fuel pool criticality analyses and 
eliminate the need to credit Boraflex neutron absorbing material for 
reactivity control.
    Date of Issuance: September 23, 2004.
    Effective Date: As of the date of issuance and shall be implemented 
by September 30, 2005.
    Amendment No.: 193.
    Renewed Facility Operating License No. DPR-67: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: January 7, 2003 (68 FR 
806). The May 14, 2003, September 29, 2003, and March 25, 2004, 
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 September 23, 2004.
    No significant hazards consideration comments received: No.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: August 25, 2003, as supplemented by 
letters dated February 9, February 23, March 25, April 15, May 20, and 
July 29, 2004.
    Description of amendment request: The amendment revised the 
Technical Specifications (TSs) to extend the emergency diesel generator 
allowed outage time from 72 hours to a period of 14 days, and to allow 
extension of the current two-hour time requirement to four hours for 
verification of redundant component operability. These changes are in 
support of installing a non-safety-related supplemental emergency power 
system. The Bases of the affected TSs will be modified to address the 
changes.
    Date of issuance: September 21, 2004.
    Effective date: As of its date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 97.
    Facility Operating License No. NPF-86: The amendment revised the 
TSS.
    Date of initial notice in Federal Register: December 29, 2003 (68 
FR 68669). The supplements dated February 9, February 23, March 25, 
April 15, May 20, and July 29, 2004, did not change the staff's 
proposed finding of no significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 21, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 14, 2004, as 
supplemented July 26, 2004.
    Brief description of amendments: The amendments modify technical 
specification (TS) 3.9.2 limiting condition for operation, delete TS 
surveillance requirements (SRs) 4.9.2.a and 4.9.2.b for the Source 
Range Neutron Flux Monitor channel functional test, revise SR 4.9.2.c 
for the channel check test, and add a requirement to perform a channel 
calibration every 18 months as well as revise TS 4.10.4.2 and 4.10.3.2 
(Units 1 and 2 respectively) for Intermediate and Power Range channel 
functional test.
    Date of issuance: September 23, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 283, 267.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 11, 2004 (69 FR 
26191).
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 23, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 25, 2004, as supplemented 
August 6, 2004.
    Brief description of amendment: The amendment revises technical 
specification (TS) 3.10.f.2 to add an allowed outage time for the 
individual rod position indication (IRPI) system of 24 hours with more 
than one IRPI group inoperable and adds the definition of 
``immediately'' to TS Section 1.0.
    Date of issuance: September 22, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 176.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 6, 2004 (69 FR 
40675).
    The supplement dated August 6, 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. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated September 22, 2004
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 7, 2003, as supplemented 
March 17, May 18, and August 18, 2004.
    Brief description of amendment: The amendment adds Technical 
Specification Section 3.3.e.1.A.3, which provides requirements for 
turbine building service water header isolation logic.

[[Page 60691]]

    Date of issuance: September 24, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 177.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 5, 2003 (68 FR 
46244).
    The supplements dated March 17, May 18, and August 18, 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. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated September 24, 2004.
    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: July 24, 2003.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3.8.4, ``DC Sources--Operating,'' TS 3.8.5, ``DC 
Sources--Shutdown,'' and TS 3.8.6, ``Battery Cell Parameters,'' and add 
a new TS 5.5.17, ``Battery Monitoring and Maintenance Program.'' The 
changes adopt in part the NRC-approved Technical Specification Task 
Force (TSTF-360, Revision 1, ``DC Electrical Rewrite.''
    Date of issuance: September 20, 2004.
    Effective date: September 20, 2004, and shall be implemented within 
120 days from the date of issuance. The licensee shall reflect the 
relocation of TS requirements to licensee-controlled programs and the 
TS Bases, as described in the licensee's letter dated July 24, 2003, 
and the NRC safety evaluation attached to the amendment, in the next 
scheduled update of the Final Safety Analysis Report Update submitted 
pursuant to 10 CFR 50.71(e).
    Amendment Nos.: Unit 1--172; Unit 2--174.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 2, 2003 (68 
FR 52236).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 20, 2004.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric 
Station, Unit 2 (SSES-2), Luzerne County, Pennsylvania

    Date of application for amendments: September 16, 2003, as 
supplemented by letter dated April 27, 2004.
    Brief description of amendments: The amendment revised the values 
of the Safety Limit for Minimum Critical Power Ratio in TS 2.1.1.2 for 
current SSES-2 Cycle 12 mid-cycle two-recirculation-loop and single-
recirculation-loop operation.
    Date of issuance: September 21, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 191.
    Facility Operating License No. NPF-22: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 28, 2003 (68 FR 
61480). The supplement dated April 27, 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. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated September 21, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 29, 2002, as supplemented 
by letters dated March 28, 2003, May 1, 2003, and August 20, 2004.
    Brief description of amendments: The amendments modify the Salem 
Nuclear Generating Station, Unit Nos. 1 and 2, Technical Specifications 
(TSs) requirements for containment closure associated with the 
equipment hatch and personnel airlocks during Core Alterations and 
movement of irradiated fuel within the containment. The change allows 
the equipment hatch and the personnel airlocks to remain open during 
fuel movement inside containment provided administrative controls are 
in place to ensure the closure of the equipment hatch and personnel 
airlock following a fuel handling accident within the containment 
building. In addition, the associated TS Bases are revised.
    Date of issuance: September 16, 2004.
    Effective date: As of the date of issuance, and shall be 
implemented within 90 days.
    Amendment Nos.: 263 and 245.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the TSs.
    Date of initial notice in Federal Register: August 20, 2002 (67 FR 
53989). The licensee's supplements dated March 28, 2003, May 1, 2003, 
and August 20, 2004, provided clarifying information that did not 
change the scope of the proposed amendments as described in the 
original notice of proposed action published in the Federal Register, 
and did not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated September 16, 
2004.
    No significant hazards consideration comments received: No.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: April 9, 2002, as supplemented 
January 10, 2003, February 24, 2004, and August 27, 2004.
    Brief description of amendment: The amendment revised the Ginna 
Improved Technical Specification with regards to: relocating figures 
associated with Core Safety Limits to the Core Operating Limits Report 
(COLR), relocating Overtemperture [Delta]T and Overpower [Delta]T 
parameters to the COLR, and replacing current trip setpoints for the 
Reactor Protection System and the Engineered Safety Feature Actuation 
System with Limiting Safety System Settings in accordance with the 
Instrument Society of America Standard 67.04, Part 2.
    Date of issuance: September 22, 2004.
    Effective date: As of the date of issuance to be implemented within 
1 year.
    Amendment No.: 85.
    Renewed Facility Operating License No. DPR-18: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: May 28, 2002 (67 FR 
36933). The supplements dated January 10, 2003, February 24, 2004 and 
August 27, 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.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 22, 2004.
    No significant hazards consideration comments received: No.

[[Page 60692]]

Southern California Edison Company, et al., Docket No. 50-206, San 
Onofre Nuclear Generating Station, Unit 1, San Diego County, California

    Date of application for amendment: January 28, 2004, supplemented 
by a letter dated July 23, 2004.
    Brief description of amendment: The amendment revises the SONGS 
Unit 1 License and Permanently Defueled Technical Specifications to 
modify or remove operational and administrative requirements that are 
not applicable upon the transfer of all spent fuel from the spent fuel 
pool into the SONGS dry cask storage Independent Spent Fuel Storage 
Installation.
    Date of issuance: September 21, 2004.
    Effective date: As of the date that all reactor fuel has been 
permanently removed from the spent fuel pool and stored in an 
Independent Spent Fuel Storage Installation. The license amendment 
shall be implemented within 30 days of its effective date.
    Amendment No.: 163.
    Facility Operating License No. DPR-13: This amendment revises both 
the license and the technical specifications.
    Date of initial notice in Federal Register: March 30, 2004 (69 FR 
16623). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 21, 2004.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama

    Date of application for amendments: July 31, 2002 as supplemented 
by letters dated December 9, 2002, February 12, 2003, March 26, 2003, 
July 11, 2003, July 17, 2003, May 17, 2004, July 2, 2004, August 24, 
2004 and September 17, 2004.
    Description of amendment request: The amendments requested full 
implementation of an alternative source term (AST) methodology for the 
Units 1, 2, and 3 operating licenses and design bases. The amendments 
adopt the AST methodology by revising the current accident source term 
and replacing it with an accident source term as prescribed in 10 CFR 
50.67. The submittals also proposed to revise and/or remove the 
Technical Specification (TS) Sections associated with control room 
emergency ventilation (CREV), standby gas treatment (SGT), standby 
liquid control (SLC), and secondary containment systems. Additionally, 
the submittals requested modification of the licensing and design basis 
to reflect the application of the AST methodology and the function of 
the SLC system, and deletion of a license condition for Units 2 and 3.
    The supplements to the original application included the withdrawal 
of the request to delete one of the TS Sections described above, 
associated with the absorption of elemental iodine by the SGT and CREV 
systems charcoal filters. Also the supplements added a new TS Section 
to require verification that the minimum fuel decay period has passed 
prior to moving fuel after the reactor is shut down.
    Date of issuance: September 27, 2004.
    Effective date: Date of issuance, to be implemented prior to 
restart of Unit 1, and within 120 days for Units 2 and 3.
    Amendment Nos.: 251, 290 and 249.
    Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: 
Amendments revised the Operating Licenses and TSs.
    Date of initial notice in Federal Register: October 15, 2002 (67 FR 
63697). The supplements dated December 9, 2002, February 12, March 26, 
July 11, and July 17, 2003, provided information that changed the scope 
of the original request, therefore another Federal Register notice was 
published on April 27, 2004 (69 FR 22883). However, the supplements 
dated May 17, July 2, August 24, and September 17, 2004, provided 
clarifying information that did not expand the scope of the revised 
request or the proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 27, 2004.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendment: March 5, 2004.
    Brief description of amendment: The amendments delete surveillance 
requirements to perform certain channel functional tests of the source 
range, intermediate, and power range neutron flux monitors. These 
amendments eliminate extraneous and unnecessary performance of these 
surveillances.
    Date of issuance: September 20, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 295 and 285.
    Facility Operating License No. DPR-77 and DPR-79: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19576). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 20, 2004.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendment: March 5, 2004.
    Brief description of amendment: The amendments eliminate the 
requirements in the technical specifications associated with hydrogen 
recombiners and hydrogen monitors.
    Date of issuance: September 20, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 296 and 286.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the technical specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19576).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 20, 2004.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 1\st\ day of October, 2004.

    For the Nuclear Regulatory Commission.
William H. Ruland,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 04-22544 Filed 10-8-04; 8:45 am]
BILLING CODE 7590-01-P