Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 24645-24662 [E5-2207]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 89 (Tuesday, May 10, 2005)]
[Notices]
[Pages 24645-24662]
From the Federal Register Online via the Government Printing 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, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the

[[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, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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 re