[Federal Register Volume 70, Number 89 (Tuesday, May 10, 2005)]
[Notices]
[Pages 24645-24662]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E5-2207]


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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 April 15, 2005 to April 28, 2005. The last 
biweekly notice was published on April 26, 2005 (70 FR 21449).

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

[[Page 24646]]

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 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 e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

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

    Date of amendment request: March 28, 2005.
    Description of amendment request: The licensee proposed to revise 
the licensing bases of OCNGS in the area of radiological dose analyses 
for the design-basis accidents (DBAs). Specifically, the licensee 
proposed to use the alternative source terms (AST) depicted in 
Regulatory Guide 1.183, ``Alternative Radiological Source Terms for 
Evaluating Design Basis Accidents at Nuclear Power Reactors,'' instead 
of the source terms used in the current licensing basis and depicted in 
Technical Information Document 14844, ``Calculation of Distance Factors 
for Power and Test Reactor Sites.'' The acceptance criteria for the 
postulated consequences using AST are set forth in 10 CFR 50.67 and 
General Design Criterion 19, ``Control Room.'' The licensee has 
performed radiological consequence analysis for the most limiting DBAs 
that result in offsite and control room operator exposure to support a 
full-scope implementation of the AST. If approved, the amendment would: 
(1) Revise Section 3.2.A, ``Standby Liquid Control System,'' of the 
Technical Specifications (TSs) to add a specification to require that 
the subject system is operable when the reactor is at or greater than 
212 degrees Fahrenheit; (2) revise various pages of the TS Bases to 
reflect use of the AST methodology. The issuance of the requested 
amendment would also signify the NRC staff's approval to revise the 
OCNGS Updated Final Safety Analysis Report to reflect implementation of 
the AST in the OCNGS licensing basis.
    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

[[Page 24647]]

consideration. The NRC staff's analysis is presented below:
    The first standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated. The AST is an input to calculations used to evaluate the 
consequences of an accident, and does not by itself affect the plant 
response, or the actual pathway of the radiation release. It does, 
however, better represent the physical characteristics of the release, 
so that appropriate mitigation techniques may be applied. The proposed 
amendment does not affect the design of plant systems, structures, or 
components (SSCs), or their operational characteristics or function. As 
a result, implementing the AST would not have any increase on the 
frequency of occurrence for previously analyzed accidents. It may be 
argued that the calculated radiological consequences are different 
because a different set of assumptions, with accompanying acceptance 
criteria, are used. However, since there is no design or operational 
change associated with the proposed amendment, the actual consequences 
of the same accident would not be changed regardless of what 
methodology was used before the accident to arrive at postulated 
consequences. As a result, implementing the AST would not increase the 
consequences of any previously evaluated accident.
    The second standard requires that operation of the unit in 
accordance with the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. The proposed amendment does not alter the design, 
configuration, or method of operation of any SSC. Therefore, no new 
initiators or precursors of a new or different kind of accident are 
created that could result in a new or different kind of accident.
    The third standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
reduction in a margin of safety. Margins of safety are established in 
the design of components, the configuration of components to meet 
certain performance parameters, and in the establishment of setpoints 
to initiate alarms or actions. These are principally documented in the 
OCNGS licensing basis documents such as the Updated Final Safety 
Analysis Report, and none of these would be changed by the amendment. 
Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.
    Based on the NRC staff's analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the proposed amendment involves no 
significant hazards consideration.
    Attorney for licensee: Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LCC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendments request: March 4, 2005.
    Description of amendments request: The proposed amendments would 
delete Section 2.F (2.G in Unit 3) of the Operating License which 
requires reporting violations of the requirements in Section 2.C of the 
Operating License. The amendments will also make administrative and 
editorial changes to the Technical Specifications (TSs). Changes to TS 
1.4, ``Frequency,'' and TS 3.4.3, ``RCS Pressure and Temperature (P/T) 
Limits,'' will correct editorial errors. The changes to TS 2.1.1, 
``Reactor Core SLs,'' and TS 3.3.1, ``Reactor Protective System (RPS) 
Instrumentation--Operating,'' will remove the reference to departure 
from nucleate boiling ratios (DNBR) based on operating cycle, since 
only one of the listed DNBR values is now valid. TS 3.1.10, ``Special 
Test Exceptions (STE)--MODES 1 and 2,'' will be changed to correct an 
inconsistency between the limiting condition for operation and the TS 
Bases. The changes to TS 3.7.2, ``Main Steam Isolation Valves (MSIVs)'' 
and TS 3.7.3, ``Main Feedwater Isolation Valves (MFIVs)'' will correct 
the applicability for these specifications. The change to TS 3.8.1, 
``AC Sources--Operating'' will add a note to a surveillance 
requirement. Changes to TS 3.8.4, ``DC Sources--Operating'' and TS 
3.8.6, ``Battery Cell Parameter'' will remove the reference to AT&T 
batteries. The changes to TS 5.5.9, ``Steam Generator (SG) Tube 
Surveillance Program'' will correct the reference for NRC notification.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Response: No.
    The proposed amendment includes [the] following changes that are 
considered to be administrative and/or editorial changes:
    The reporting requirement in License Condition 2.F (2.G in Unit 
3) is adequately addressed by the requirements identified in 10 CFR 
50.72, ``Immediate notification requirements for operating nuclear 
power reactors'' and 10 CFR 50.73, ``Licensee event report system.'' 
Since Condition 2.F (2.G in Unit 3) is adequately addressed by the 
requirements in 10 CFR 50.72 and 10 CFR 50.73, the Condition is not 
required. Therefore, this is considered an administrative change 
that eliminates regulatory requirements that are adequately 
addressed by the requirements in 10 CFR 50.72 and 10 CFR 50.73.
    The changes to Technical Specifications (TS) 1.4 and 3.4.3 are 
editorial changes only. These changes maintain the format of the 
Technical Specifications and correct editorial errors in the 
Technical Specifications.
    The changes to Technical Specifications 2.1.1 and 3.3.1 remove 
requirements that are no longer applicable to the Palo Verde Nuclear 
Generating Station (PVNGS) units. As part of Amendment 133 to the 
PVNGS Operating License, the minimum DNBR was revised based on Unit 
operating cycle, >=1.30 (through operating cycle 10)'' and >=1.34 
(operating cycle 11 and later).'' All three PVNGS units have 
completed operating cycle 10. Therefore, the reference to the 
minimum d[e]parture from nucleate boiling ratio (DNBR) through 
operating cycle 10 (>=1.30) is no longer required.
    The changes to Technical Specification 3.1.10 correct an 
inconsistency between the Technical Specification limiting condition 
for operation (LCO) and Bases. The Bases for this specification 
states that ``Even if an accident occurs during PHYSICS TESTS with 
one or more LCOs suspended, fuel damage criteria are preserved 
because the limits on power distribution and shutdown capability are 
maintained during PHYSICS TESTS.'' The limits on power distribution 
are maintained by TSs 3.2.1, ``Linear Heat Rate (LHR)'' and 3.2.4 
``Departure from Nucleate Boiling Ratio (DNBR).'' These changes 
ensure that shutdown capability is maintained during physics tests.
    The changes to Technical Specifications Section 3.7.2, ``Main 
Steam Isolation Valves (MSIVs)'' and Section 3.7.3, ``Main Feedwater 
Isolation Valves (MFIVs)'' correct an inconsistency between the 
applicability and the required actions. The changes are consistent 
with the guidance in NUREG-1432, ``Standard Technical 
Specifications, Combustion Engineering Plants.'' Therefore, this is 
considered an administrative change that corrects an inconsistency 
in the Technical Specifications.
    The changes to Technical Specifications Section 3.8.1, ``AC 
Sources--Operating,'' correct an inconsistency in the surveillance 
requirements that were revised in Amendment 129 to the PVNGS 
Operating License. A note was not included with the change to one of 
the surveillance requirements. This change adds the note to

[[Page 24648]]

the surveillance requirement. Therefore, this is considered an 
administrative change that corrects an inconsistency in the 
Technical Specifications.
    The changes to Technical Specifications Section 3.8.4, ``DC 
Sources--Operating'' and Section 3.8.6, ``Battery Cell Parameters'' 
removes the requirements and references to the AT&T batteries. APS 
has replaced the AT&T batteries with low specific gravity batteries 
in all three units. Therefore, this is considered an administrative 
change that removes unnecessary requirements and references.
    The changes to Technical Specifications Section 5.5.9, ``Steam 
Generator (SG) Tube Surveillance Program,'' updates the requirement 
to notify the NRC based on the January 23, 2001 rule change to 10 
CFR 50.72. Therefore, this change corrects NRC notification 
requirements in Technical Specifications, based on the January 23, 
2001 rule change to 10 CFR 50.72 (65 FR 63786, 10/25/00).
    As discussed above the proposed amendment involves 
administrative and/or editorial changes only. The proposed amendment 
does not impact any accident initiators, analyzed events, or assumed 
mitigation of accident or transient events. The proposed changes do 
not involve the addition or removal of any equipment or any design 
changes to the facility. The proposed changes do not affect plant 
operations, any design function or an analysis that verifies the 
capability of structures, systems, and components (SSCs) of the 
plant. The proposed changes do not change any of the previously 
evaluated accidents in the updated final safety analysis report 
(UFSAR). The proposed changes do not affect SSCs, operating 
procedures, and administrative controls that have the function of 
preventing or mitigating any of these accidents.
    Therefore, the proposed changes do not represent 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.
    As discussed in standard 1, the proposed amendment only involves 
administrative and/or editorial changes. No actual plant equipment 
or accident analysis will be affected by the proposed changes. The 
proposed changes will not change the design function or operation of 
any SSCs. The proposed changes will not result in any new failure 
mechanisms, malfunctions, or accident initiators not considered in 
the design and licensing bases. The proposed amendment does not 
impact any accident initiators, analyzed events, or assumed 
mitigation of accident or transient events.
    Therefore, this proposed change does not create the possibility 
of an accident of a different kind than previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety.
    Response: No.
    As discussed in standard 1, the proposed amendment only involves 
administrative and/or editorial changes. Margin of safety is 
associated with confidence in the ability of the fission product 
barriers (i.e., fuel and fuel cladding, reactor coolant system 
pressure boundary, and containment structure) to limit the level of 
radiation dose to the public. This request involves administrative 
and/or editorial changes only. No actual plant equipment or accident 
analysis will be affected by the proposed changes. Additionally, the 
proposed changes will not relax any criteria used to establish 
safety limits, will not relax any safety system settings, or will 
not relax the bases for any limiting conditions for operation.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona 
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix, 
Arizona 85072-2034.
    NRC Section Chief: Robert A. Gramm.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of amendments request: January 27, 2005.
    Description of amendments request: The proposed amendment would 
allow 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 TSs, 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 would be 
revised to reflect the related changes to LCO 3.0.4, and Surveillance 
Requirement (SR) 3.0.4 would be revised to reflect the LCO 3.0.4 
allowance.
    This change was proposed by the industry's TS 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 January 27, 2005.
    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

[[Page 24649]]

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: Carey Fleming, Esquire, Counsel, 
Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor, 
Baltimore, MD 21202.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: March 14, 2005.
    Description of amendment request: The proposed amendments would 
delete Technical Specification (TS) Section 5.5.4, ``Post Accident 
Sampling,'' 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 NRC's 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 March 14, 2005.
    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 
Specifications (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: Anne W. Cottingham, Winston and Strawn LPP, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: John A. Nakoski.

[[Page 24650]]

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

    Date of amendment request: September 23, 2004, as supplemented by 
letter dated April 19, 2005.
    Description of amendment request: The amendment would revise the 
reactor operational limits, as specified in the River Bend Station Core 
Operating Limits Report (COLR), to compensate for the inoperability of 
the End of Cycle Recirculation Pump Trip (EOC-RPT) instrumentation. 
This will provide an alternative to the existing Limiting Condition for 
Operation for the EOC-RPT instrumentation. The revised Technical 
Specification will require that either the EOC-RPT instrumentation be 
operable or that Minimum Critical Power Ratio and Linear Heat 
Generation Rate limits for the inoperable EOC-RPT be placed in effect 
as specified in the COLR.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The End of Cycle Recirculation Pump Trip (EOC-RPT) functions to 
insert negative reactivity in response to certain anticipated 
transients. The EOC-RPT is a mitigation function and not the 
initiator of any evaluated accident or transient. Operation with 
inoperable EOC-RPT instrumentation and compliance with new 
restrictive Minimum Critical Power Ratio (MCPR) and Linear Heat 
Generation Rate (LHGR) operating limits establish sufficient margin 
to the core thermal MCPR safety limit (SL) and the thermal 
mechanical design limits as would be the case with operable EOC-RPT 
instrumentation and existing MCPR and LHGR limits.
    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 will not create any new modes of plant or 
equipment operation. The proposed change allows the option to apply 
an additional penalty factor to the MCPR and LHGR when the EOC-RPT 
is inoperable. With the addition of the penalty factor, the margin 
to the MCPR SL and the thermal mechanical design limits are 
maintained. Therefore, operating the plant with the proposed change 
will not create the possibility of a new or different kind of 
accident from any previously analyzed.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    By establishing a new restrictive MCPR and LHGR operating limit, 
there are no changes to the plant design and safety analysis. There 
are no changes to the reactor core design instrument setpoints. The 
margin of safety assumed in the safety analysis is not affected. 
Applicable regulatory requirements will continue to be met and 
adequate defense-in[-]depth will be maintained. Sufficient safety 
margins will be maintained.
    The analytical methods used to determine the revised core 
operating limits were reviewed and approved by the NRC, and are 
described in Technical Specification 5.6.5. Specific analyses were 
prepared by the RBS fuel vendor to develop core operating limits 
without crediting the EOC-RPT. Therefore, implementation of the 
proposed changes will not involve a significant reduction in the 
margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: February 15, 2005.
    Description of amendment request: The proposed amendment would 
approve application of an alternative source term methodology with the 
exception that Technical Information Document 14844, ``Calculation of 
Distance Factors for Power Test Reactor Sites,'' will continue to be 
used as the radiation dose basis for equipment qualification.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The implementation of AST assumptions has been evaluated in 
revisions to the analyses of the following limiting DBAs at the 
Byron Station and Braidwood Station.

Loss-of-Coolant Accident
Fuel Handling Accident
Control Rod Ejection Accident
Locked Rotor Accident
Main Steam Line Break Accident
Steam Generator Tube Rupture Accident

    Based upon the results of these analyses, it has been 
demonstrated that, with the requested changes, the dose consequences 
of these limiting events are within the regulatory guidance provided 
by the NRC for use with the AST methodology. This guidance is 
presented in RG 1.183, and Standard Review Plan Section 15.0.1. The 
AST is an input to calculations used to evaluate the consequences of 
an accident and does not by itself affect the plant response or the 
actual pathway of the activity released from the fuel. It does, 
however, better represent the physical characteristics of the 
release such that appropriate mitigation techniques may be applied.
    The AST methodology follows the guidance provided in RG 1.183 
and satisfies the dose limits in 10 CFR 50.67. Even though these 
limits are not directly comparable to the previously specified whole 
body and thyroid requirements of 10 CFR 50, Appendix A, General 
Design Criteria (GDC) 19, ``Control room,'' and 10 CFR 100.11, 
``Determination of exclusion area, low population zone, and 
population center distance,'' the results of the AST analyses have 
demonstrated that the 10 CFR 50.67 limits are satisfied. Therefore, 
it is concluded that AST does not involve a significant increase in 
the consequences of an accident previously evaluated.
    Implementation of AST provides increased operating margins for 
the control room ventilation system filtration efficiencies. It also 
relaxes containment integrity requirements while handling irradiated 
fuel that has decayed for greater than 48 hours and during core 
alterations. Automatic initiation of the radiation isolation mode 
for the control room is not credited in the accident analysis which 
allows relaxation of certain Technical Specification surveillance 
requirements.
    The equipment affected by the proposed changes is mitigative in 
nature and relied upon after an accident has been initiated. 
Application of the AST does result in changes to the functions and 
operation of various filtration systems as described in the Updated 
Final Safety Analysis Report (UFSAR). These effects have been 
considered in the evaluations for these proposed changes. While the 
operation of various systems does change with the implementation of 
AST, the affected systems are not accident initiators; and 
application of the AST methodology, itself, is not an initiator of a 
design basis accident. The proposed changes to the TS revise certain 
equipment performance requirements but do not require any physical 
changes to the plant.
    As a result, the proposed changes do not affect any of the 
parameters or conditions

[[Page 24651]]

that could contribute to the initiation of any accidents. Relaxation 
of operability requirements during the specified conditions will not 
significantly increase the probability of occurrence of an accident 
previously analyzed. Since design basis accident initiators are not 
being altered by adoption of the AST, the probability of an accident 
previously evaluated is not affected.
    Based on the above discussion, the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not involve a physical change to the 
plant. Implementation of AST provides increased operating margins 
for filtration system efficiencies. Application of AST also allows 
for the relaxation of containment integrity requirements while 
handling irradiated fuel that has decayed for greater than 48 hours 
and during core alterations. Automatic initiation of the radiation 
isolation mode for the control room is no longer credited in the 
accident analysis.
    Similarly, the proposed changes do not require any physical 
changes to any structures, systems or components involved in the 
mitigation of any accidents. Therefore, no new initiators or 
precursors of a new or different kind of accident are created. New 
equipment or personnel failure modes that might initiate a new type 
of accident are not created as a result of the proposed changes.
    Based on the above discussion, the proposed changes do 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?
    Approval of a change from the original source term methodology 
(i.e., TID 14844) to an AST methodology, consistent with the 
guidance in RG 1.183, will not result in a significant reduction in 
the margin of safety. The safety margins and analytical 
conservatisms associated with the AST methodology have been 
evaluated and were found acceptable. The results of the revised DBA 
analyses, performed in support of the proposed changes, are subject 
to specific acceptance criteria as specified in RG 1.183. The dose 
consequences of these DBAs remain within the acceptance criteria 
presented in 10 CFR 50.67 and RG 1.183.
    The proposed changes continue to ensure that the doses at the 
exclusion area boundary (EAB) and low population zone boundary 
(LPZ), as well as the control room, are within the specified 
regulatory limits.
    Therefore, based on the above discussion, the proposed changes 
do not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Gene Y. Suh.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, 
Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio; Docket 
Nos. 50-334 and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 
(BVPS-1 and 2), Beaver County, Pennsylvania; Docket No. 50-440, Perry 
Nuclear Power Plant, Unit 1, Lake County, Ohio

    Date of amendment request: February 22, 2005.
    Description of amendment request: The requested change will delete 
Technical Specification requirements related to Occupational Radiation 
Exposure Reports and 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 February 22, 2005.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates the Technical Specifications 
(TSs) reporting requirements to provide a monthly operating report 
of shutdown experience and operating statistics if the equivalent 
data is submitted using an industry electronic database. It also 
eliminates the TS reporting requirement for an annual occupational 
radiation exposure report, which provides information beyond that 
specified in NRC regulations. The proposed change involves no 
changes to plant systems or accident analyses. As such, the change 
is administrative in nature and does not affect initiators of 
analyzed events or assumed mitigation of accidents or transients. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant, add any new equipment, or require any existing equipment 
to be operated in a manner different from the present design. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chiefs: Gene Y. Suh, Richard J. Laufer.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of amendment request: March 22, 2005.
    Description of amendment request: The proposed amendments revise 
the Technical Specifications (TS) for several Reactor Protection System 
functional units. The steam/feedwater flow mismatch coincident with 
steam generator water level--low reactor trip is being deleted, the 
reactor trip on turbine trip interlock is being changed from P-7 to P-
8, the value of the P-8 interlock setpoint is being changed from 45 
percent rated thermal power (RTP) to 40 percent RTP, and the value of 
the P-8 interlock allowable value is being changed from 48 percent RTP 
to 43 percent RTP.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes revise the operability requirements, 
surveillance requirements and the interlock setpoint for two Reactor 
Trip System functional units. The affected trip functional units are 
not initiators of any accident previously evaluated. The proposed 
changes to the affected trip functional units do not adversely 
affect the initiators of any accident previously evaluated. A best 
estimate

[[Page 24652]]

analysis has shown that a turbine trip without a reactor trip below 
40% power does not challenge the pressurizer PORVs [power operated 
relief valves] or the steam generator safety valves; thereby, not 
adversely affecting the probability of a small break LOCA [loss of 
coolant accident] due to a stuck open PORV, or an excessive cooldown 
event due to a stuck open steam generator safety valve. As a result, 
the probability of any accident previously evaluated is not 
significantly increased by the proposed changes.
    The steam/feedwater flow mismatch coincident with steam 
generator water level--low reactor trip is not credited as a primary 
trip in any previously evaluated accidents. The reactor trip on 
turbine trip below the P-8 interlock is not credited as a primary 
trip in any previously evaluated accidents. Therefore, the 
mitigation functions that have been assumed in the accident analyses 
will continue to be performed by the systems and components 
currently credited in the analyses; and the accident analysis 
results are not affected by the changes to the affected trip 
functional units. The P-8 setpoint is not an initial condition of 
any accident previously evaluated. Therefore, the accident analysis 
results are not affected by changes to the P-8 setpoint. No safety 
analyses previously performed in the Turkey Point Units 3 and 4 
UFSAR [Updated Final Safety Analysis Report] required reanalysis for 
these proposed changes. All accident analyses acceptance criteria 
continue to be met. The proposed changes do not create any new 
credible limiting single failure. As a result, the consequences of 
any accident previously evaluated are not significantly increased by 
the proposed changes.
    In conclusion, operation of the facility in accordance with the 
proposed amendments does not involve a significant increase in the 
probability or consequences of any accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    No changes are being made to the plant that would introduce any 
new accident causal mechanisms. The proposed changes do not 
adversely affect previously identified accident initiators and do 
not create any new accident initiators. No new limiting single 
failures or accident scenarios are created by the proposed changes. 
No new challenges to any installed safety system are created by 
these proposed changes. The proposed changes do not result in any 
event previously deemed incredible being made credible.
    The steam/feedwater flow mismatch coincident with steam 
generator water level--low reactor trip is not credited as an 
inhibitor of any potential or actual accident initiators. So, 
deletion of this reactor trip functional unit will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    Changing the interlock for the reactor trip on turbine trip from 
P-7 to P-8 changes the power level associated with enabling and 
disabling the reactor trip on turbine trip function. The turbine 
pressure input to the reactor protection system permissives is not 
an accident initiator and is not credited in the accident analyses. 
Changing the P-8 allowable and trip setpoint values changes the 
power level associated with enabling and disabling the reactor trip 
functions currently associated with P-8. The change does not affect 
how the associated trip functional units operate or function. Since 
these interlock changes do not affect the way that the associated 
trip functional units operate or function, the changes do not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    Therefore, operation of the facility in accordance with the 
proposed amendments does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    No UFSAR safety analyses were changed or modified as a result of 
these proposed changes. Therefore, all margins associated with the 
current UFSAR safety analyses acceptance criteria are unaffected. 
The current UFSAR safety analyses remain bounding. No UFSAR Chapter 
14 events explicitly credit the steam/feedwater flow mismatch 
reactor trip function and the reactor trip on turbine trip function 
below the P-8 setpoint value. The safety systems credited in the 
safety analyses will continue to be available to perform their 
mitigation functions. Changing the P-8 setpoint from 45% to 40% is 
in the conservative direction for the Reactor Coolant Flow--Low 
Reactor Trip and the Reactor Coolant Pump Breaker Position Reactor 
Trip. Therefore, the proposed changes do not result in a significant 
reduction in a margin of safety; and operation of the facility in 
accordance with the proposed amendments would not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: March 28, 2005.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to allow the option of not 
measuring the moderator temperature coefficient within 7 effective 
full-power days after reaching an equilibrium boron concentration of 
300 parts per million. This option would be available if the benchmark 
criteria in WCAP-13749-P-A and the revised prediction specified in the 
core operating limits report are satisfied.
    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[s] do[es] not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The probability or consequences of accidents previously 
evaluated in the UFSAR [updated final safety analysis report] are 
unaffected by this proposed change. There is no change to any 
equipment response or accident mitigation scenario, and this change 
results in no additional challenges to fission product barrier 
integrity. The proposed change does not alter the design, 
configuration, operation, or function of any plant system, 
structure, or component. Further, the existing limits on moderator 
temperature coefficient (MTC) established by the Technical 
Specifications (TS), based on assumptions in the safety analyses, 
remain unchanged and continue to be satisfied. As a result, the 
outcomes of previously evaluated accidents are unaffected. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change[s] do[es] not create the possibility of a 
new or different kind of accident from any previously evaluated.
    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. The proposed change neither installs 
or removes any plant equipment, nor alters the design, physical 
configuration, or mode of operation of any plant structure, system, 
or component. The MTC is a variable that must remain within 
prescribed limits, but it is not an accident initiator. No physical 
changes are being made to the plant, so no new accident causal 
mechanisms are being introduced. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. The proposed change[s] do[es] not involve a significant 
reduction in the margin of safety.
    The margin of safety associated with the acceptance criteria of 
any accident is unchanged. The proposed change will have no affect 
on the availability, operability, or performance of the safety-
related systems and components. The proposed change does not alter 
the design, configuration, operation, or function of any plant 
system, structure, or component. The ability of any operable 
structure, system, or component to perform its designated safety 
function is unaffected by

[[Page 24653]]

this change. A change to a surveillance requirement is proposed 
based on an alternative method of confirming that the surveillance 
is met. The Technical Specifications establish limits for the 
moderator temperature coefficient (MTC) based on assumptions in the 
accident analyses. Applying the conditional exemption from the MTC 
measurement changes the method of meeting the surveillance 
requirement; however, this change does not modify the TS values and 
ensures adherence to the current TS limits. The basis for the 
derivation of the MTC limits from the moderator density coefficient 
(MDC) assumed in the accident analysis is unchanged. Further, the 
safety analysis assumption of a constant MDC and its assumed value 
will not change. Therefore, the margin of safety as defined in the 
TS is not reduced and the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Section Chief: Darrell J. Roberts.

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

    Date of amendment request: March 28, 2005.
    Description of amendment request: The proposed amendment would 
revise Seabrook Station, Unit No. 1 (Seabrook) Technical Specification 
(TS) 3/4.9.13, ``Spent Fuel Assembly Storage.'' This revision would 
reflect a revised criticality safety analysis supporting a two-zone 
spent fuel pool consisting of BORAFLEX[reg] and BORAL[reg] fuel 
assembly storage racks. Additionally, the proposed change would create 
TS 3/4.9.15, ``Spent Fuel Pool Boron Concentration.''
    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 changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed license amendment incorporates the results of a 
revised criticality analysis for the spent fuel pool without making 
any physical changes to the facility. The revised criticality 
analysis for the spent fuel pool (1) credits boron during movement 
of fuel in the spent fuel pool, (2) assumes no neutron-absorbing 
material in the BORAFLEX [reg] storage racks, and (3) applies a 
conservative penalty in the analysis of BORAL [reg] racks. These 
changes do not increase the probability of a fuel assembly being 
misplaced within the spent fuel pool. The movement of fuel 
assemblies will continue to be controlled by approved procedures, 
and the placement of spent fuel will be controlled by the revised 
Technical Specifications. The proposed changes do not alter or 
prevent the ability of structures, systems, or components (SSCs) to 
perform their intended function to mitigate the consequences of an 
initiating event within the acceptance limits assumed in the Updated 
Final Safety Analysis Report (UFSAR).
    The proposed changes do not affect the source term, containment 
isolation or radiological release assumptions used in evaluating the 
radiological consequences of an accident previously evaluated in the 
Seabrook Station UFSAR. The consequences of a misplaced fuel 
assembly are not increased because the analysis demonstrates that 
the fuel will remain sub-critical with a minimum of 872 ppm [part 
per million] boron in the spent fuel pool. The new technical 
specification included in this proposed change will ensure that the 
minimum boron concentration is established during the movement of 
fuel in the spent fuel pool. Further, the proposed changes neither 
increase the types and amounts of radioactivity released offsite nor 
increase occupational or public radiation exposures.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes to the TS do not alter the operation of the 
spent fuel storage system or its ability to perform its design 
function. The proposed changes do not include any physical changes 
to the plant and do not introduce a new or different accident from 
any type previously evaluated. A misplaced fuel assembly does not 
represent a new or different type [of] accident, and the analysis 
shows that the fuel remains sub-critical for the limiting case of a 
misplaced fuel assembly. Similarly, continuing to take credit for 
boron in the spent fuel under accident conditions does not create 
the possibility of a new or different kind of accident. The previous 
criticality analyses took credit for soluble boron in the spent fuel 
pool water to show acceptable results in the analyses of fuel 
misloading events.
    Therefore the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The changes proposed by this license amendment ensure that the 
spent fuel will remain sub-critical under normal and accident 
conditions. The controlled placement of fuel assemblies within the 
spent fuel pool will maintain Keff less than or equal to 
0.95 as required by TS 5.6.1.1 for spent fuel storage. The proposed 
amendment maintains the 0.95 limit on Keff by restricting 
the placement of spent fuel and by crediting soluble boron in the 
fuel pool water.
    To assure that the true reactivity will be less than the 
calculated reactivity, the analyses contain conservative assumptions 
for calculating the safety limits for the spent fuel rack. With this 
proposed change, Keff will be less than or equal to 0.95 
with a 95% probability at a 95% confidence level.
    Therefore, the proposed amendment does not result in a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Section Chief: Darrell J. Roberts.

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

    Date of amendment request: April 1, 2005.
    Description of amendment request: The licensee proposed to revise 
Section 3.8.7, ``Inverters--Operating,'' of the Technical 
Specifications (TSs), extending the time allowed to fix inoperable 
emergency uninterruptible power supply (UPS) inverters from the current 
24 hours to 7 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff's analysis is presented below:
    The first standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated. The proposed amendment does not affect the design of the 
emergency UPS inverters, the operational characteristics or function of 
the inverters, the interfaces between the inverters and other plant 
systems, or the reliability of the inverters. An inoperable emergency 
UPS inverter was not considered an initiator of a previously analyzed 
event. In addition, the required actions and the associated completion 
times specified by the TSs are not initiators of previously evaluated 
accidents. As a result, extending the completion time

[[Page 24654]]

for an inoperable emergency UPS inverter would not have a significant 
impact on the frequency of occurrence for a previously analyzed 
accident. Furthermore, the proposed amendment will not result in 
modifications to plant activities associated with inverter maintenance, 
but rather, provides operational flexibility by allowing additional 
time to perform inverter corrective maintenance and post-maintenance 
testing on-line. The proposed extension of inoperable time will not 
significantly affect the capability of inverters to perform their 
safety function, which is to ensure an uninterruptible supply of 120-
volt alternating current (ac) electrical power to the associated power 
distribution subsystems. The licensee performed a probabilistic risk 
assessment which concluded that the increase in plant risk is small. 
Therefore, the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    The second standard requires that operation of the unit in 
accordance with the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. The proposed amendment does not alter the design, 
configuration, or method of operation of the emergency UPS inverters or 
their associated 120-volt ac uninterruptible power distribution 
subsystems, nor does the amendment alter any safety analyses inputs and 
assumptions. The proposed extended emergency UPS inverter completion 
time does not reduce the number of emergency UPS inverters below the 
minimum required for safe shutdown or accident mitigation, and does not 
affect the parameters within which NMP2 is operated or the setpoints at 
which protective or mitigative actions are initiated. The use of the 
alternate safety-related maintenance supply to power the 120-volt ac 
uninterruptible power distribution subsystem is consistent with the 
NMP2 design. If a station blackout event were to occur while an 
emergency UPS inverter is out of service, a dedicated portable power 
supply would be connected to provide a continuous source of power to 
the connected systems. Accordingly, no new failure modes, system 
interactions, or accident responses will be created that could result 
in a new or different kind of accident.
    The third standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
reduction in a margin of safety. Margins of safety are established in 
the design of components, the configuration of components to meet 
certain performance parameters, and in the establishment of setpoints 
to initiate alarms or actions. The proposed amendment will not affect 
any margin of safety as defined in the NMP2 Updated Safety Analysis 
Report. The amendment does not change the design or operational 
parameters of the UPS inverters as compared to original plant design. 
Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.
    Based on the NRC staff's analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the proposed amendment involves no 
significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: April 1, 2005.
    Description of amendment request: The proposed amendment would 
provide one-time extension to the completion time for restoration of a 
service water train to operable status in Technical Specification (TS) 
3.7.8, ``Service Water System (SWS).''
    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 amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment does not involve a significant increase 
in the probability of an accident previously evaluated because the 
extended Technical Specification action completion time is not an 
accident initiator. Therefore the probability is not increased 
significantly.
    The proposed amendment does not involve a significant increase 
in the consequences of an accident previously evaluated. With 
service water pump P-7C inoperable, 100% of the required post-
accident SWS cooling capability remains available with the redundant 
train maintained operable. A risk analysis was performed to show 
that the consequences are not significantly increased. The 
compensatory measures provide additional assurance that there is no 
significant increase in the consequences of an accident associated 
with extending the Technical Specification action completion time 
for the service water system for an additional 96 hours.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed amendment only extends the Technical 
Specification action completion time and does not involve a physical 
alteration of any system, structure or component (SSC), or change in 
the way any SSC is operated. The proposed amendment does not involve 
operation of any required SSCs in a manner or configuration 
different from those previously recognized or evaluated. No new 
failure mechanisms will be introduced by the changes being 
requested.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment does not involve a significant reduction 
in a margin of safety. With service water pump P-7C inoperable, 100% 
of the required post-accident service water system cooling 
capability remains available with the redundant train maintained 
operable. Therefore, there is no significant reduction in the margin 
of safety.
    Based on the availability of redundant systems, the compensatory 
measures that will be taken, and the low probability of an accident 
that could not be mitigated by the available systems, the proposed 
amendment would not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR Part 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: January 28, 2005.
    Description of amendment request: The proposed amendment would 
revise

[[Page 24655]]

the SSES 2 Technical Specification (TS) Table 3.3.5.1-1 ``Emergency 
Core Cooling System Instrumentation,'' to change Function 3.e ``HPCI 
[High-Pressure Coolant Injection] System,'' conditions referenced from 
Required Action A.1 from ``D'' to ``C.'' This is an editorial revision 
to correct a typographical error that has been present since PPL 
converted to the Improved Technical Specifications in 1998.
    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?
    Response: No.
    The proposed change to the Unit 2 TS Table 3.3.5.1 provides a 
correction to a typographical error that occurred when preparing a 
change to Unit 2 Technical Specification Table 3.3.5.1-1 in the 
response to an NRC Request for Additional Information (RAI). The 
request was initiated during NRC review of documents submitted by 
PPL for the conversion to the Improved Technical Specifications. 
This proposed change is considered to be administrative in nature 
because it was originally submitted correctly and was inadvertently 
changed in response to the RAI.
    Therefore, this 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.
    As stated above, the proposed change to the Unit 2 TS Table 
3.3.5.1 provides a correction to a typographical error that occurred 
when preparing the response to an NRC Request for Additional 
Information. The request was initiated by the NRC during its review 
of documents submitted by PPL for the conversion to the Improved 
Technical Specifications. This proposed change is administrative in 
nature.
    Therefore, these 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.
    Again, the proposed change to the Unit 2 TS Table 3.3.5.1 
provides a correction to a typographical error that occurred when 
preparing the response to an NRC Request for Additional Information. 
The request was initiated by the NRC during its review of documents 
submitted by PPL for the conversion to the Improved Technical 
Specifications. This proposed change is administrative in nature.
    Therefore, these proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Richard J. Laufer.

PSEG Nuclear LLC, Docket Nos. 50-272, Salem Nuclear Generating Station, 
Unit No. 1 Salem County, New Jersey

    Date of amendment request: February 23, 2005.
    Description of amendment request: The proposed changes will revise 
Technical Specification (TS) Steam Generator (SG) requirements for 
Salem Nuclear Generating Station, Unit No. 1. The proposed changes 
would replace TS 3/4.4.5 ``Steam Generator (SG)'' with ``Steam 
Generator Tube Integrity;'' add a new TS 6.8.4.i, ``Steam Generator 
Program;'' and add a new reporting requirement TS 6.9.1.10 ``Steam 
Generator Tube Inspection Report.'' Additionally, the proposed changes 
would revise TS 3/4.4.6.2, ``Reactor Coolant System Operational 
Leakage.'' Specifically, the Limiting Condition for Operation and 
ACTION and Surveillance Requirements of TS 3/4.4.6.2 would be revised 
to clarify the requirements related to primary-to-secondary leakage. 
These changes would facilitate implementation of industry initiative 
Nuclear Energy Institute (NEI) 97-08, ``Steam Generator Program 
Guidelines,'' to allow a comprehensive, performance-based approach to 
managing SG performance at Salem Nuclear Generating Station, Unit No. 
1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change[s] require[s] a Steam Generator Program that 
includes performance criteria that will provide reasonable assurance 
that the steam generator (SG) tubing will retain integrity over the 
full range of operating conditions (including startup, operation in 
the power range, hot standby, cool down and all anticipated 
transients included in the design specification). The SG performance 
criteria are based on tube structural integrity, accident induced 
leakage, and operational leakage.
    The structural integrity performance criterion is:
    All in-service steam generator tubes shall retain structural 
integrity over the full range of normal operating conditions 
(including startup, operation in the power range, hot standby, and 
cool down and all anticipated transients included in the design 
specification) and design basis accidents. This includes retaining a 
safety factor of 3.0 against burst under normal steady state full 
power operation primary-to-secondary pressure differential and a 
safety factor of 1.4 against burst applied to the design basis 
accident primary-to-secondary pressure differentials. Apart from the 
above requirements, additional loading conditions associated with 
the design basis accidents, or combination of accidents in 
accordance with the design and licensing basis, shall also be 
evaluated to determine if the associated loads contribute 
significantly to burst or collapse. In the assessment of tube 
integrity, those loads that do significantly affect burst or 
collapse shall be determined and assessed in combination with the 
loads due to pressure with a safety factor of 1.2 on the combined 
primary loads and 1.0 on axial secondary loads.
    The accident induced leakage performance criterion is:
    The primary-to-secondary accident induced leakage rate for any 
design basis accidents, other than a SG tube rupture, shall not 
exceed the leakage rate assumed in the accident analysis in terms of 
total leakage rate for all SGs and leakage rate for an individual 
SG. Leakage is not to exceed 1 gpm per SG.
    The operational leakage performance criterion is:
    The reactor coolant system operational primary-to-secondary 
leakage through any one SG shall be limited to 150 gallons per day.
    A steam generator tube rupture (SGTR) event is one of the design 
basis accidents that are analyzed as part of a plant's licensing 
basis. In the analysis of a[n] SGTR event, a bounding primary-to-
secondary leakage rate equal to the operational leakage rate limits 
in the licensing basis plus the leakage rate associated with a 
double-ended rupture of a single tube is assumed.
    For other design basis accidents such as main steam line break 
(MSLB), rod ejection, and reactor coolant pump locked rotor the 
tubes are assumed to retain their structural integrity (i.e., they 
are assumed not to rupture). These analyses assume that primary-to-
secondary leakage for all SGs is 1 gallon per minute or increases to 
1 gallon per minute as a result of accident-induced stresses. The 
accident induced leakage criterion retained by the proposed changes 
accounts for tubes that may leak during design basis accidents. The 
accident induced leakage criterion limits this leakage to no more 
than the value assumed in the accident analysis.

[[Page 24656]]

    The SG performance criteria proposed as part of these TS changes 
identify the standards against which tube integrity is to be 
measured. Meeting the performance criteria provides reasonable 
assurance that the SG tubing will remain capable of fulfilling its 
specific safety function of maintaining reactor coolant pressure 
boundary integrity throughout each operating cycle and in the 
unlikely event of a design basis accident. The performance criteria 
are only a part of the Steam Generator Program required by the 
proposed addition of TS 6.8.4.i. The program defined by NEI 97-06 
includes a framework that incorporates a balance of prevention, 
inspection, evaluation, repair, and leakage monitoring.
    The consequences of design basis accidents are, in part, 
functions of the DOSE EQUIVALENT I-131 in the primary coolant and 
the primary-to-secondary leakage rates resulting from an accident. 
Therefore, limits are included in the Salem TS for operational 
leakage and for DOSE EQUIVALENT I-131 in primary coolant to ensure 
the plant is operated within its analyzed condition. The Salem 
analysis of the limiting design basis accident assumes that primary-
to-secondary leak rate after the accident is 1 gallon per minute 
with no more than 500 gallons per day through any one SG, and that 
the reactor coolant activity levels of DOSE EQUIVALENT I-131 are at 
the TS values before the accident.
    The proposed change[s] do[es] not affect the design of the SGs, 
their method of operation, or primary coolant chemistry controls. 
The proposed approach updates the current TS and enhances the 
requirements for SG inspections.
    The proposed change[s] do[es] not adversely impact any other 
previously evaluated design basis accident and [are] an improvement 
over the current TS.
    Therefore, the proposed changes do not affect the consequences 
of a[n] SGTR accident and the probability of such an accident is 
reduced. In addition, the proposed changes do not affect the 
probabilities or consequences of an MSLB, rod ejection, or a reactor 
coolant pump locked rotor event.
    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 performance based requirements are an improvement 
over the requirements imposed by the current TS.
    Implementation of the proposed Steam Generator Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
result of the implementation of the Steam Generator Program will be 
an enhancement of SG tube performance. Primary-to-secondary leakage 
that may be experienced during all plant conditions will be 
monitored to ensure it remains within current accident analysis 
assumptions.
    The proposed changes do not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change[s] do[es] not impact any 
other plant system or component. The change[s] enhance[s] SG 
inspection requirements.
    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 SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
also isolate the radioactive fission products in the primary coolant 
from the secondary system. In summary, the safety function of a SG 
is maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change[s] do[es] not affect tube design or operating environment. 
The proposed change[s] [are] expected to result in an improvement in 
the tube integrity by implementing the Steam Generator Program to 
manage SG tube inspection, assessment, and plugging. The 
requirements established by the Steam Generator Program are 
consistent with those in the applicable design codes and standards 
and are an improvement over the requirements in the current TS.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed changes to the 
TS.

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

Southern California Edison Company (SCE), et al., Docket Nos. 50-361 
and 50-362, San Onofre Nuclear Generating Station, Unit 2 and Unit 3, 
San Diego County, California

    Date of amendment requests: March 24, 2005.
    Description of amendment requests: The proposed change would revise 
the following Technical Specifications (TSs):
     TS 1.1, Definitions, correct the definition of SHUTDOWN 
MARGIN (SDM).
     TS 3.1.1, SHUTDOWN MARGIN (SDM)--Tavg > 2000F, and TS 
3.1.2, SHUTDOWN MARGIN (SDM)--Tavg < 2000F, relocate the numerical 
shutdown margin requirements to the Core Operating Limits Report 
(COLR).
     TS 3.1.3, Reactivity Balance, increase the required action 
time from 72 hours to 7 days when the ``Core reactivity balance not 
within limit.''
     TS 3.1.5, Control Element Assembly (CEA) Alignment, TS 
3.1.6, Shutdown Control Element Assembly (CEA) Insertion Limits, and TS 
3.1.7, Regulating CEA Insertion Limits, remove the requirement to 
verify SDM.
     TS 3.2.4, Departure From Nucleate Boiling Ratio (DNBR), 
relocate to the COLR the power margin that must be accommodated when 
the Core Operating Limit Supervisory System (COLSS) is in service and 
neither CEA calculator is OPERABLE.
     TS 5.7.1.5, CORE OPERATING LIMITS REPORT (COLR), identify 
that the limits for TSs 3.1.1 and 3.1.2 shall be in the COLR.
    The proposed changes are consistent with the Standard Technical 
Specifications for Combustion Engineering Plants, NUREG-1432, Revision 
3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The Limiting Conditions of Operation (LCOs) and Core Operating 
Limits Report (COLR) will continue to restrict operation to within 
the regions that provide acceptable results. The safety analysis 
will continue to be performed in accordance with the Nuclear 
Regulatory Commission (NRC) approved San Onofre Units 2 and 3 reload 
analysis methodology.
    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 add any new equipment, modify any 
interfaces with any existing equipment, alter the equipment's 
function, or change the method of operating the equipment. The 
proposed change does not alter plant conditions in a manner that 
could affect other plant components. The proposed change does not 
cause any existing equipment to become an accident initiator.

[[Page 24657]]

    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.
    Safety Limits ensure that Specified Acceptable Fuel Design 
Limits are not exceeded during steady state operation, normal 
operational transients, and anticipated operational occurrences. All 
fuel limits and design criteria will continue to be met, based on 
the NRC approved San Onofre Units 2 and 3 reload analysis 
methodology. Therefore, the proposed change will have no impact on 
the margins as defined in the Technical Specification bases.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Robert A. Gramm.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for a Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) The 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail 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: December 16, 2004.
    Brief description of amendments: The amendments delete TS 5.6.1, 
``Occupational Radiation Exposure Report'' and TS 5.6.4, ``Monthly 
Operating Reports,'' as described in the Notice of Availability 
published in the Federal Register on June 23, 2004 (69 FR 35067).
    Date of issuance: April 27, 2005.
    Effective date: April 27, 2005, and shall be implemented within 90 
days of the date of issuance.
    Amendment Nos.: Unit 1-154, Unit 2-154, Unit 3-154.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5236).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 27, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 17, 2004.
    Brief Description of amendments: The amendments eliminate the 
requirements to submit monthly operating reports and annual 
occupational radiation exposure reports.
    Date of issuance: April 19, 2005.
    Effective date: April 19, 2005.
    Amendment Nos.: 235 and 263.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: February 15, 2005 (70 
FR 7763).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 19, 2005.
    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: November 17, 2004.
    Brief description of amendment: The amendment eliminates the 
requirements to submit monthly operating reports and annual 
occupational radiation exposure reports.
    Date of issuance: April 19, 2005.
    Effective date: April 19, 2005.
    Amendment No.: 204.
    Renewed Facility Operating License No. DPR-23. Amendment revises 
the Technical Specifications.
    Date of initial notice in Federal Register: February 15, 2005 (70 
FR 7763)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 19, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 17, 2004.
    Brief description of amendment: This amendment revises Technical 
Specifications by eliminating the requirements to submit monthly 
operating reports and annual occupational radiation exposure reports.
    Date of issuance: April 19, 2005.
    Effective date: April 19, 2005.
    Amendment No.: 118.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 2004 (70 
FR 7763).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 19, 2005.

[[Page 24658]]

    No significant hazards consideration comments received: No.

Entergy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of application for amendment: September 23, 2004, as 
supplemented by letter dated January 13, 2005.
    Brief description of amendment: The change revises Columbia 
Generating Station's licensing basis by replacing the current plant-
specific reactor pressure vessel material surveillance program with the 
boiling water reactor vessels and internals project (BWRVIP) integrated 
surveillance program (ISP). Specifically, the amendment revises 
Columbia's final safety analysis report to include participation in the 
ISP as described in the program document BWRVIP-86-A, ``BWR [Boiling 
Water Reactor] Vessel and Internals Project Updated BWR Integrated 
Surveillance Program (ISP) Implementation Plan,'' dated October 2002.
    Date of issuance: April 28, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 192.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 26, 2004 (69 FR 
62471).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 28, 2005.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: December 20, 2004.
    Brief description of amendment: The amendment deletes TS 5.6.1, 
``Occupational Radiation Exposure Report,'' and TS 5.6.4, ``Monthly 
Operating Reports,'' as described in the Notice of Availability 
published in the Federal Register on June 23, 2004 (69 FR 35067).
    Date of issuance: April 14, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 223.
    Renewed Facility Operating License No. DPR-51: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2890).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 14, 2005.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Docket Nos. 50-247 and 50-286, Indian Point 
Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York

    Date of application for amendment: October 22, 2004.
    Brief description of amendment: These amendments revise the 
Technical Specifications by eliminating the requirements associated 
with hydrogen recombiners and hydrogen monitors.
    Date of issuance: April 14, 2005.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 243 and 228.
    Facility Operating License Nos. DPR-26 and DPR-64: Amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5240).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 14, 2005.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Docket Nos. 50-247 and 50-286, Indian Point 
Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York

    Date of application for amendment: October 25, 2004.
    Brief description of amendment: These amendments revise the 
Technical Specifications by eliminating the requirements to submit 
monthly operating reports and occupational radiation exposure reports.
    Date of issuance: April 14, 2005.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 242 and 227.
    Facility Operating License Nos. DPR-26 and DPR-64: Amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5241).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 14, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 30, 2004.
    Brief description of amendments: The amendments modify the 
technical specification (TS) requirements to adopt the provisions of 
the industry/TS Task Force (TSTF) change TSTF-359, ``Increased 
Flexibility in Mode Restraints.''
    Date of issuance: April 5, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days.
    Amendment Nos.: 141, 141, 134, 134.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 26, 2004.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 5, 2005.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of application for amendments: September 15, 2004.
    Brief description of amendments: The proposed amendment will delete 
the Technical Specification (TS) requirements related to hydrogen/
oxygen monitors. The proposed TS changes support implementation of the 
revisions to Title 10 of the Code of Federal Regulations (10 CFR), 
Section 50.44, ``Standards for Combustible Gas Control System in Light-
Water-Cooled Power Reactors,'' that became effective on October 16, 
2003. The changes are consistent with Revision 1 of the NRC-approved 
Industry/Technical Specifications Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-447, ``Elimination of Hydrogen 
Recombiners and Change to Hydrogen and Oxygen Monitors.''
    Date of issuance: April 28, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment Nos.: 213/205.
    Facility Operating License Nos. DPR-19, DPR-25: The amendments 
revised the Technical Specifications.

[[Page 24659]]

    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5243).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 28, 2005.
    No significant hazards consideration comments received: No.

Exelon Generating Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: September 15, 2004.
    Brief description of amendments: The amendments delete the 
Technical Specification requirements to maintain hydrogen recombiners 
and hydrogen/oxygen monitors and related Surveillance Requirements. The 
revised Title 10 of the Code of Federal Regulations (10 CFR) Section 
50.44, ``Combustible Gas Control for Nuclear Power Plants,'' eliminated 
the requirements for hydrogen recombiners and relaxed safety 
classifications and licensee commitments to certain design 
qualification criteria for hydrogen and oxygen monitors.
    Date of issuance: April 22, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment Nos.: 172/158.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5243).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 22, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 25, 2003.
    Brief description of amendments: The amendment revised the 
Technical Specifications (TSs) associated with Reactor Coolant System--
CHEMISTRY. Specifically, the amendment relocates Reactor Coolant 
System--CHEMISTRY, in its entirety from the TSs to the Technical 
Requirements Manual (TRM). In addition, the amendment deletes the 
specific activity requirements related to E-Bar, gross beta and gross 
gamma.
    Date of issuance: April 18, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 174 and 136.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the TSs.
    Date of initial notice in Federal Register: February 17, 2004 (69 
FR 7522).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 18, 2005.
    No significant hazards consideration determination comments 
received: No.

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

    Date of application for amendment: November 17, 2004.
    Brief description of amendment: The amendment eliminates the 
requirements to submit monthly operating reports and annual 
occupational radiation exposure reports.
    Date of issuance: April 19, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 217.
    Facility Operating License No. DPR-72: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 2005 (70 
FR 7768).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 19, 2005.
    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: June 28, 2004.
    Description of amendment request: The amendment revised the 
Seabrook Station, Unit No. 1 Technical Specifications (TSs) to align 
the language of Surveillance Requirement 4.9.4 with that of Limiting 
Condition for Operation 3.9.4, ``Containment Building Penetrations.'' 
The amendment changes the requirement from ``during core alterations 
and the movement of irradiated fuel'' to ``during the movement of 
recently irradiated fuel.''
    Date of issuance: April 21, 2005.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 102.
    Facility Operating License No. NPF-86: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: August 31, 2004 (69 FR 
53110).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 21, 2005.
    No significant hazards consideration comments received: No.

National Aeronautics and Space Administration, Docket No. 50-30, the 
Plum Brook Test Reactor, Sandusky, Ohio

    Date of application for amendment: January 14, 2005.
    Brief description of amendment: The amendment clarifies the license 
requirements for confirmation of Final Status Survey results prior to 
backfilling or covering of excavated areas.
    Date of issuance: April 21, 2005.
    Effective date: The license amendment is effective as of its date 
of issuance.
    Amendment No.: 12.
    Facility License No. TR-3: This amendment consists of changes to 
the Facility License.
    Date of initial notice in Federal Register: March 15, 2005 (70 FR 
12743).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation enclosed with the amendment dated April 21, 
2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: January 24, 2005.
    Brief description of amendments: The amendments deleted Sections 
6.6.1 and 5.6.1, ``Occupational Radiation Exposure Report,'' and 
Sections 6.6.4 and 5.6.4, ``Monthly Operating Reports,'' from the NMP1 
and NMP2 Technical Specifications.
    Date of issuance: April 19, 2005.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: 188 and 115.
    Facility Operating License Nos. DPR-63 and NPF-69: Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: February 15, 2005 (70 
FR 7769).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 19, 2005.
    No significant hazards consideration comments received: No

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

    Date of application for amendment: September 22, 2004.
    Brief description of amendment: The amendment extended the validity 
of the

[[Page 24660]]

reactor pressure vessel pressure-temperature limit curves from May 1, 
2005, to May 1, 2006.
    Date of issuance: April 25, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 197.
    Facility Operating License No. NPF-22: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 7, 2004 (69 FR 
70721).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 25, 2005.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Dates of application for amendments: February 26 and April 28, 
2004, as supplemented by letters dated July 8 and October 20, 2004.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) Section 5.6.6, Reactor Coolant System (RCS) Pressure 
Temperature Limits Report (PTLR), to facilitate future licensee-
controlled changes to the PTLR. The changes include a revised PTLR that 
provides new heatup and cooldown limits and Cold Overpressure 
Protection System (COPS) setpoints, and to recalculate the minimum size 
of the pressurizer power operated relief valve orifice of the RCS vent. 
In addition, the changes relocate the COPS arming temperature to the 
PTLR, and lower the COPS arming temperature from 350 [deg]F to 220 
[deg]F. The licensee also included TS bases changes to support the 
changes to the TSs.
    Date of issuance: March 28, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 136 (Unit 1) and 115 (Unit 2).
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19575) and April 22, 2004 (69 FR 34707)
    The supplements dated July 8 and October 20, 2004, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 28, 2005.
    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 amendments: December 2, 2004.
    Brief description of amendments: The amendments modify technical 
specification (TS) requirements for mode change limitations in Limiting 
Condition for Operation 3.0.4 and Surveillance Requirement 4.0.4 
consistent with Industry/TS Task Force (TSTF) Standard TS Change 
Traveler, TSTF-359, Revision 9, ``Increased Flexibility in Mode 
Restraints.'' A notice of availability for this TS improvement using 
the Consolidated Line Item Improvement Process was published in the 
Federal Register (FR) on April 4, 2003 (68 FR 16579).
    Date of issuance: April 11, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 301, 290.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2901)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 11, 2005.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: September 23, 2003, as supplemented by 
letter dated June 9, 2004.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TSs) to extend the interval between local 
leak rate tests for the containment purge and vent valves with 
resilient seats (containment purge valves, hydrogen purge valves, and 
containment pressure relief valves).
    Date of issuance: April 13, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 116 and 116.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the TSs.
    Date of initial notice in Federal Register: November 12, 2003 (68 
FR 64140).
    The supplement dated June 9, 2004, provided additional information 
that clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 13, 2005.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 13, 2004.
    Brief description of amendment: The amendment revised Surveillance 
Requirement (SR) 3.8.1.7 (fast-start test), SR 3.8.1.12 (safety 
injection actuation signal test), SR 3.8.1.15 (hot restart test), and 
SR 3.8.1.20 (redundant unit test) to clarify what voltage and frequency 
limits are applicable during the transient and steady state portions of 
the diesel generator start testing performed by these SRs.
    Date of issuance: April 21, 2005.
    Effective date: April 21, 2005, and shall be implemented within 90 
days from the date of issuance.
    Amendment No.: 161.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2904)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 21, 2005.
    No significant hazards consideration comments received: No.

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

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

[[Page 24661]]

which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) The 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room (PDR), located at One White Flint North, Public File Area 01F21, 
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly 
available records will be accessible from the Agencywide Documents 
Access and Management System's (ADAMS) Public Electronic Reading Room 
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems 
in accessing the documents located in ADAMS, contact the PDR Reference 
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. Within 60 days after the date 
of publication of this notice, the licensee may file a request for a 
hearing with respect to issuance of the amendment to the subject 
facility operating license and any person whose interest may be 
affected by this proceeding and who wishes to participate as a party in 
the proceeding must file a written request for a hearing and a petition 
for leave to intervene. Requests for a hearing and a petition for leave 
to intervene shall be filed in accordance with the Commission's ``Rules 
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. 
Interested persons should consult a current copy of 10 CFR 2.309, which 
is available at the Commission's PDR, located at One White Flint North, 
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland, and electronically on the Internet at the NRC Web site, 
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR Reference staff at 
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a 
request for a hearing or petition for leave to intervene is filed by 
the above date, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
intends to rely to establish those facts or expert opinion. The 
petition must include sufficient information to show that a genuine 
dispute exists with the applicant on a material issue of law or 
fact.\1\ Contentions shall be limited to matters within the scope of 
the amendment

[[Page 24662]]

under consideration. The contention must be one which, if proven, would 
entitle the petitioner to relief. A petitioner/requestor who fails to 
satisfy these requirements with respect to at least one contention will 
not be permitted to participate as a party.
---------------------------------------------------------------------------

    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly available because they 
are asserted to contain safeguards or proprietary information, 
petitioners desiring access to this information should contact the 
applicant or applicant's counsel and discuss the need for a 
protective order.
---------------------------------------------------------------------------

    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
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 e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer or the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

    Date of amendment request: April 11, 2005, as supplemented on April 
14, 2005.
    Description of amendment request: The amendments revise Technical 
Specification (TS) 5.5.9, ``Steam Generator (SG) Tube Surveillance 
Program,'' to incorporate changes in the SG inspection scope for 
Braidwood Station, Unit 2 only, during refueling outage 11.
    Date of issuance: April 25, 2005.
    Effective date: April 25, 2005.
    Amendment Nos.: 135, 135.
    Facility Operating License Nos. NPF-72 and NPF-77: Amendment 
revises the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. Joliet Herald News, April 15 and 18, 2005, 
and Morris Daily Herald, April 19, 2005. The announcement provided an 
opportunity to submit comments on the Commission's proposed NSHC 
determination. No comments have been received. The Commission's related 
evaluation of the amendment, finding of exigent circumstances, state 
consultation, and final NSHC determination are contained in a safety 
evaluation dated April 25, 2005.
    Attorney for licensee: Thomas S. O'Neil.
    NRC Section Chief: Gene Y Suh.

    Dated at Rockville, Maryland, this 2nd day of May 2005.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. E5-2207 Filed 5-9-05; 8:45 am]
BILLING CODE 7590-01-P