[Federal Register Volume 69, Number 61 (Tuesday, March 30, 2004)]
[Notices]
[Pages 16615-16627]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 04-6682]


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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

[[Page 16616]]

(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 March 5, 2004 through March 18, 2004. The 
last biweekly notice was published on March 16, 2004 (69 FR 12361).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.

[[Page 16617]]

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

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck 
Plant, Middlesex County, Connecticut

    Date of amendment request: January 9, 2004.
    Description of amendment requests: The Haddam Neck Plant (HNP) is 
currently undergoing active decommissioning. The proposed amendment 
would revise Technical Specifications (TS) 6.6.4, 6.7.1, and 6.8 in 
accordance with Technical Specification Task Force (TSTF) travelers 
152, 258 and 308 to reflect changes to Title 10 Part 20 of the Code of 
Federal Regulations (CFR). The proposed amendment would also revise TS 
6.1, 6.2.1, 6.4, 6.5, and 6.6 to reflect the use of generic 
organizational titles, modeled after TSTF 65 revision 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. Will the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed changes are proposed to reflect the current 
version of 10 CFR 20 and to eliminate the need for a TS change each 
time there is a organizational change. These changes do not impact 
any design basis accidents described in the updated Final Safety 
Analysis Report (FSAR) for the HNP. Since the proposed changes are 
administrative or editorial, they cannot affect the likelihood or 
consequences of accidents.
    Therefore, the proposed administrative changes to the Operating 
License and Technical Specifications will not increase the 
probability or consequences of an accident previously evaluated.
    2. Will the proposed change create the possibility of a new or 
different type of accident from any accident previously evaluated?
    No. The proposed changes do not affect plant system operation. 
The proposed changes do not involve a physical alteration to the 
plant or any change in plant configuration. The proposed changes do 
not require any new operator actions. The changes do not alter the 
way any structure, system, or component functions. The changes do 
not introduce any new failure modes.
    Therefore, this proposed administrative change will not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. Will the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed changes will make the HNP Operating License and 
Defueled Technical Specifications consistent with the current 10 CFR 
20, and eliminate the need for a TS change each time there is an 
organizational change. The proposed changes will not result in any 
technical changes to current requirements. The proposed changes have 
no effect on assumptions and any acceptance criteria for the design 
basis accidents described in the updated FSAR for the HNP.
    Therefore, the proposed administrative changes will not result 
in a 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.
    NRC Section Chief: Claudia Craig.

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

    Date of amendment request: February 9, 2004.
    Description of amendment request: The proposed amendment deletes 
requirements from the technical specifications (TS) to maintain 
hydrogen recombiners and hydrogen monitors. Licensees were generally 
required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' 
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' Implementation of these upgrades 
was an outcome of the lessons learned from the accident that occurred 
at TMI Unit 2. Requirements related to combustible gas control were 
imposed by Order for many facilities and were added to or

[[Page 16618]]

included in the TS for nuclear power reactors currently licensed to 
operate. The revised 10 CFR 50.44, ``Standards for Combustible Gas 
Control System in Light-Water-Cooled Power Reactors,'' eliminated the 
requirements for hydrogen recombiners and relaxed safety 
classifications and licensee commitments to certain design and 
qualification criteria for hydrogen and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on September 
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated February 9, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
RG 1.97 Category 1, is intended for key variables that most directly 
indicate the accomplishment of a safety function for design-basis 
accident events. The hydrogen monitors no longer meet the definition 
of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 
50.44 the Commission found that Category 3, as defined in RG 1.97, 
is an appropriate categorization for the hydrogen monitors because 
the monitors are required to diagnose the course of beyond design-
basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations (PARs) to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3 and removal of 
the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the severe accident 
management guidelines (SAMGs), the emergency plan (EP), the 
emergency operating procedures (EOPs), and site survey monitoring 
that support modification of emergency plan PARs.
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TS, does not involve a 
significant increase in the probability or the consequences of any 
accident previously evaluated.

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

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

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

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

    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: February 9, 2004.
    Description of amendment request: The proposed amendment deletes 
requirements from the technical specifications (TS) to maintain 
hydrogen recombiners and hydrogen monitors. Licensees were generally 
required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' 
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' Implementation of these upgrades 
was an outcome of the lessons learned from the accident that occurred 
at TMI Unit 2. Requirements related to combustible gas control were 
imposed by Order for many facilities and were added to or included in 
the TS for nuclear power reactors currently licensed to operate. The 
revised 10 CFR 50.44, ``Standards for Combustible Gas Control System in 
Light-Water-Cooled Power Reactors,'' eliminated the requirements for 
hydrogen recombiners and relaxed safety classifications and licensee 
commitments to certain design and qualification criteria for hydrogen 
and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NHSC) determination for referencing 
in license amendment applications in the Federal Register on September 
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC

[[Page 16619]]

determination in its application dated February 9, 2004. Basis for 
proposed no significant hazards consideration determination: As 
required by 10 CFR 50.91(a), an analysis of the issue of no significant 
hazards consideration is presented below:

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

    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
RG 1.97 Category 1, is intended for key variables that most directly 
indicate the accomplishment of a safety function for design-basis 
accident events. The hydrogen monitors no longer meet the definition 
of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 
50.44 the Commission found that Category 3, as defined in RG 1.97, 
is an appropriate categorization for the hydrogen monitors because 
the monitors are required to diagnose the course of beyond design-
basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations (PARs) to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3 and removal of 
the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the severe accident 
management guidelines (SAMGs), the emergency plan (EP), the 
emergency operating procedures (EOPs), and site survey monitoring 
that support modification of emergency plan PARs.
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TS, does not involve a 
significant increase in the probability or the consequences of any 
accident previously evaluated.

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

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

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

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

    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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: February 3, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.1.8, ``Scram Discharge Volume (SDV) 
Vent and Drain Valves,'' to allow a vent or drain line with one 
inoperable valve to be isolated instead of requiring the valve to be 
restored to operable status within seven days.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on February 24, 2003 (68 FR 8637), on possible 
amendments to revise the action for one or more SDV vent or drain lines 
with an inoperable valve, 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 15, 
2003 (68 FR 18294). The licensee affirmed the applicability of the 
model NSHC determination in its application dated February 3, 2004. 
Basis for proposed no significant hazards consideration determination: 
As required by 10 CFR 50.91(a), an analysis of the issue of no 
significant hazards consideration is presented below:

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

    A change is proposed to allow the affected SDV vent and drain 
line to be isolated when there are one or more SDV vent or drain 
lines with one valve inoperable instead of requiring the valve to be 
restored to operable status within 7 days. With one SDV vent or 
drain valve inoperable in one or more lines, the isolation function 
would be maintained since the redundant valve in the affected line 
would perform its safety function of isolating the SDV. Following 
the completion of the required action, the isolation function is 
fulfilled since the associated line is isolated. The ability to vent 
and drain the SDVs is maintained and controlled through 
administrative controls. This requirement assures the reactor 
protection system is not adversely affected by the inoperable 
valves. With the safety functions of the valves being maintained, 
the probability or consequences of an accident previously evaluated 
are not significantly increased.

[[Page 16620]]

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

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. Thus, 
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 proposed change ensures that the safety functions of the SDV 
vent and drain valves are fulfilled. The isolation function is 
maintained by redundant valves and by the required action to isolate 
the affected line. The ability to vent and drain the SDVs is 
maintained through administrative controls. In addition, the reactor 
protection system will prevent filling of an SDV to the point that 
it has insufficient volume to accept a full scram. Maintaining the 
safety functions related to isolation of the SDV and insertion of 
control rods ensures that the proposed change does not involve a 
significant reduction in the margin of safety.

    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: Robert A. Gramm.

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

    Date of amendment request: December 9, 2003.
    Description of amendment request: The amendment would allow a one-
time increase in the completion time for restoring an inoperable 
emergency feedwater (EFW) system train to operable status to allow the 
realignment of the diesel-driven EFW pump during power operations.
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed license amendment extends, on a one-time basis, the 
Completion Time for restoring an inoperable Emergency Feedwater 
System train to Operable status. The Emergency Feedwater System is 
designed to provide cooling for components essential to the 
mitigation of plant transients and accidents. The system is not an 
initiator of design basis accidents. During the requested extended 
time period of 14 days, the redundant Emergency Feedwater Pump (EFP) 
will be available and capable of providing cooling to the Once-
Through Steam Generators (OTSGs) during emergency conditions. In 
addition, a safety-grade motor driven pump (EFP-1) is available for 
manual initiation and is capable of providing adequate EFW flow for 
OTSG cooling during all design basis events. EFP-1 is also capable 
of being supplied by the ``A'' train emergency diesel generator if 
sufficient electrical loading capacity is available during a loss of 
offsite power condition. Although Feedwater (FW) pump FWP-7 is non-
safety related and its motor is non-seismic, it will also be 
available and capable of providing OTSG cooling during all but the 
most limiting design basis events. FWP-7 also has a non-safety 
diesel backup power supply in the event normal power is not 
available.
    A Probabilistic Safety Assessment (PSA) has been performed to 
assess the risk impact of an increase in Completion Time. Although 
the proposed one-time change results in an increase in Core Damage 
Frequency and Large Early Release Frequency, the value of these 
increases are considered as very small in the current regulatory 
guidance.
    Therefore, granting this LAR [License Amendment Request] does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does not create the possibility of a new or different type of 
accident from any accident previously evaluated.
    The proposed license amendment extends, on a one-time basis, the 
Completion Time for restoring an inoperable Emergency Feedwater 
System train to Operable status.
    The proposed LAR will not result in changes to the design, 
physical configuration of the plant or the assumptions made in the 
safety analysis. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does not involve a significant reduction in the margin of 
safety.
    The proposed license amendment extends, on a one-time basis, the 
Completion Time for restoring an inoperable Emergency Feedwater 
System train to Operable status. The proposed change will allow 
online alignment of one of the Emergency Feedwater pumps to improve 
its reliability, thus increasing the long-term margin of safety of 
the system.
    The proposed LAR will reduce the probability (and associated 
risk) of a plant shutdown to repair an Emergency Feedwater pump. To 
ensure defense-in-depth capabilities and the assumptions in the risk 
assessment are maintained during the proposed one-time extended 
Completion Time, CR-3 [Crystal River Unit 3] will continue the 
performance of 10 CFR 50.65(a)(4) assessments before performing 
maintenance or surveillance activities. Other compensatory actions 
that may be implemented include: use of pre-job briefings and 
periodic operator walkdowns to assess the status of risk sensitive 
equipment in the redundant train, use of operator walkdowns to 
assess and limit transient combustibles in risk significant fire 
areas, and no elective maintenance to be scheduled in the switchyard 
that would challenge the availability of offsite power to the ES 
[engineered safeguards] buses.
    As described above in Item 1, a PSA has been performed to assess 
the risk impact of an increase in Completion Time. Although the 
proposed one-time change results in an increase in Core Damage 
Frequency and Large Early Release Frequency, the value of these 
increases are considered as very small in the current regulatory 
guidance.
    Therefore, granting this LAR does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: William F. Burton, Acting.

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 3, 2004.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) Surveillance Requirement 4.0.5 by 
updating the American Society of Mechanical Engineers (ASME) Boiler and 
Pressure Vessel Code references as the source of inservice testing 
requirements for ASME Code Class 1, 2, and 3 pumps and valves. The 
proposed amendments replace reference to Section XI of the Code with 
reference to ASME Code for Operation and Maintenance of Nuclear Power 
Plants (ASME OM Code).
    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. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    The proposed changes do not involve a significant increase in 
the probability of an accident previously evaluated because no such 
accidents are affected by the proposed changes. The amendments 
application proposes to revise the Turkey Point Units 3

[[Page 16621]]

and 4 Technical Specifications Surveillance Requirement 4.0.5. The 
proposed changes would revise the technical specifications to 
conform to the requirements of 10 CFR 50.55a(f) regarding the 
inservice testing of pumps and valves for the Fourth 10-Year 
interval.
    The current Turkey Point Units 3 and 4 Technical Specifications 
reference the ASME Code, Section XI, requirements for the inservice 
testing of ASME Code Class 1, 2, and 3 pumps and valves. The 
proposed changes would reference the ASME OM Code, which is 
consistent with 10 CFR Section 50.55a(f). In addition, surveillance 
interval definitions for ``biennially or every 2 years'' as used in 
the ASME OM Code would be added to TS surveillance requirement 
4.0.5.b to ensure consistent interpretation of the terms.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because no new or different accident initiators are introduced by 
these proposed changes.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    The proposed changes do not involve a significant reduction in a 
margin of safety because there are no changes to initial conditions 
contributing to accident severity or consequences. Thus, there is 
not 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: William F. Burton, Acting.

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

    Date of amendment request: December 23, 2003.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to eliminate the reactor head 
cooling containment isolation function since the reactor head cooling 
system has been removed from service. In addition, the TS are being 
changed to correct and clarify existing requirements, make wording 
enhancements, and revise an existing limiting condition for operation 
for radiation monitors used to isolate reactor building ventilation and 
initiate the standby gas treatment system (SGTS).
    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.
    One of the proposed changes removes the Reactor Head Cooling 
system primary containment isolation signal from the TS. The 
existing piping will be removed and the existing process pipe 
through the containment penetration will be cut and capped. This 
equipment was only used for the shutdown-cooling (non-safety 
related) mode of operation. This system does not support safe 
shutdown of the facility. The proposed TS change does not introduce 
new equipment or new equipment operating modes, nor does the 
proposed change alter existing system relationships. These proposed 
changes do not increase the likelihood of the malfunction of any 
structure, system or component (SSC) or impact any analyzed 
accident. Consequently, the probability of an accident previously 
evaluated is not increased.
    The other proposed change adds an allowable outage time to the 
radiation monitors described in TS that initiate the SGTS and adds a 
time requirement for placing inoperable channels in a tripped 
condition. The proposed TS change does not introduce new equipment 
or new equipment operating modes, nor does the proposed change alter 
existing system relationships. The change does not affect plant 
operation, design function or any analysis that verifies the 
capability of a SSC to perform a design function. Further, the 
proposed change does not increase the likelihood of the malfunction 
of any structure, system or component (SSC) or impact any analyzed 
accident. Consequently, the probability of an accident previously 
evaluated is not affected.
    Therefore, the proposed amendment does 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.
    One of the proposed changes removes the Reactor Head Cooling 
system primary containment isolation signal from the TS. The 
existing piping will be removed and the existing process pipe 
through the containment penetration will be cut and capped. This 
equipment was only used for the shutdown-cooling (non-safety 
related) mode of operation. The change does not create the 
possibility of new credible failure mechanisms, or malfunctions. The 
proposed change does not introduce new accident initiators. 
Consequently, the changes cannot create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The other proposed change adds an allowable outage time to the 
radiation monitors described in TS that initiate the SGTS and adds a 
time requirement for placing inoperable channels in a tripped 
condition. This change does not modify the design function or 
operation of any SSC. Further the change does not involve physical 
alterations of the plant; no new or different type of equipment will 
be installed. The proposed change is not an indicator of any 
accident previously evaluated. Consequently, the probability of an 
accident previously evaluated is not affected.
    Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously analyzed.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    One of the proposed changes removes the Reactor Head Cooling 
system primary containment isolation signal from the TS. The 
existing piping will be removed and the existing process pipe 
through the containment penetration will be cut and capped. This 
equipment was only used for the shutdown-cooling (non-safety 
related) mode of operation. This system does not support safe 
shutdown of the facility. This change does not exceed or alter a 
design basis or a safety limit for a parameter established in the 
MNGP [Monticello Nuclear Generating Plant] Updated Safety Analysis 
Report (USAR) or the MNGP facility license. Consequently, the change 
does not result in a significant reduction in the margin of safety.
    The other proposed change adds an allowable outage time to the 
radiation monitors described in TS that initiate the SGTS and adds a 
time requirement for placing inoperable channels in a tripped 
condition. This change ensures continued compliance with regulatory 
and licensing requirements. The change does not exceed or alter a 
design basis or safety limit for a parameter established in the MNGP 
USAR or MNGP facility license. Consequently, the proposed amendment 
does not involve a significant reduction in the margin of safety.
    Therefore, the proposed amendment does not involve a significant 
reduction in the margin of safety.

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

[[Page 16622]]

    NRC Section Chief: L. Raghavan.

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

    Date of amendment requests: February 13, 2004.
    Description of amendment requests: The amendment would revise 
Technical Specifications (TSs) 3.3.1, ``Reactor Trip System (RTS) 
Instrumentation,'' 3.3.2, ``Engineered Safety Feature Actuation System 
(ESFAS) Instrumentation,'' and 3.3.6, ``Containment Ventilation 
Isolation Instrumentation.'' The purpose of the amendment is to adopt 
the completion time, test bypass time, and surveillance frequency time 
changes approved by the NRC in Topical Reports WCAP-14333-P-A, 
``Probabilistic Risk Analysis of the RPS [reactor protection system] 
and ESFAS Test Times and Completion Times,'' and WCAP-15376-P-A, 
``Risk-Informed Assessment of the RTS and ESFAS Surveillance Test 
Intervals and Reactor Trip Breaker Test and Completion Times.'' The 
proposed changes would revise the required actions for certain action 
conditions; increase the completion times for several required actions 
(including some notes); delete notes in certain required actions; 
increase frequency time intervals (including certain notes) in several 
surveillance requirements (SRs); add an action condition and required 
actions; add or revise notes in certain SRs; and revise Table 3.3.1-1. 
There are also administrative corrections to the format of the TSs 
(e.g., remove the bold appearance of page number 3.3-45).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since no 
hardware changes are proposed. The same RTS and ESFAS 
instrumentation will continue to be used. The protection systems 
will continue to function in a manner consistent with the plant 
design basis. These changes to the TS [in the amendment] do not 
result in a condition where the design, material, and construction 
standards that were applicable prior to the change are altered.
    The proposed changes will not modify any system interface. The 
proposed changes will not affect the probability of any event 
initiators [because the proposed changes are not event initiators]. 
There will be no degradation in the performance of or an increase in 
the number of challenges imposed on safety-related equipment assumed 
to function during an accident situation. There will be no change to 
normal plant operating parameters or accident mitigation 
performance. The proposed changes will not alter any assumptions or 
change any mitigation actions in the radiological consequence 
evaluations in the Updated Final Safety Analysis Report [for Diablo 
Canyon Units 1 and 2].
    The determination that the results of the proposed changes are 
acceptable [to be considered for plant-specific TS] was established 
in the NRC Safety Evaluations prepared for WCAP-14333-P-A, Revision 
1, (issued by letter dated July 15, 1998) and for WCAP-15376-P-A, 
Revision 1, (issued by letter dated December 20, 2002). 
Implementation of the proposed changes will result in an 
insignificant risk impact. Applicability of these conclusions has 
been verified through plant-specific reviews and implementation of 
the generic analysis results in accordance with the respective NRC 
Safety Evaluation conditions [for the two WCAPs].
    The proposed changes to the CTs [completion times], test bypass 
times, and Surveillance Frequencies reduce the potential for 
inadvertent reactor trips and spurious engineered safety features 
actuations, and therefore do not increase the probability of any 
accident previously evaluated. The proposed changes do not change 
the response of the plant to any accidents and have an insignificant 
impact on the reliability of the RTS and ESFAS signals. The RTS and 
ESFAS will remain highly reliable and the proposed changes will not 
result in a significant increase in the risk of plant operation. 
This is demonstrated by showing that the impact on plant safety as 
measured by the increase in core damage frequency (CDF) is less than 
1.0E-06 per year and the increase in large early release frequency 
(LERF) is less than 1.0E-07 per year. In addition, for the CT 
changes, the incremental conditional core damage probabilities 
(ICCDP) and incremental conditional large early release 
probabilities (ICLERP) are less than 5.0E-07 and 5.0E-08, 
respectively. These changes meet the acceptance criteria in 
Regulatory Guides (RGs) 1.174 and 1.177. Therefore, since the RTS 
and ESFAS will continue to perform their [safety] functions with 
high reliability as originally assumed, and the increase in risk as 
measured by [Delta]CDF, [Delta]LERF, ICCDP, ICLERP risk metrics is 
within the acceptance criteria of existing [NRC] regulatory 
guidance, there will not be a significant increase in the 
consequences of any accidents.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components from 
performing their intended [safety] function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed changes do not affect the source term, 
containment isolation, or radiological release assumptions used in 
evaluating the radiological consequences of any accident previously 
evaluated. The proposed changes are consistent with safety analysis 
assumptions and resultant consequences.
    Therefore, [the] change[s do] not increase the probability or 
consequences of any accident previously evaluated.
    2.The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. The proposed changes will not affect the normal method of 
plant operation. No performance requirements will be affected or 
eliminated. The proposed changes will not result in physical 
alteration to any plant system nor will there be any change in the 
method by which any safety-related plant system performs its safety 
function. There will be no setpoint changes or changes to accident 
analysis assumptions.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of these changes. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of these changes.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do not affect the acceptance criteria for 
any analyzed event nor is there a change to any Safety Analysis 
Limit. There will be no effect on the manner in which safety limits, 
limiting safety system settings, or limiting conditions for 
operation are determined nor will there be any effect on those plant 
systems necessary to assure the accomplishment of protection 
functions. There will be no impact on the overpower limit, departure 
from nucleate boiling limits, local power peaking factor 
(FQ), hot channel factor (F[Delta]H), loss-of-coolant 
accident (LOCA) peak cladding temperature, peak local power density, 
or any other margin of safety. The radiological dose consequence 
acceptance criteria listed in the [NRC] Standard Review Plan will 
continue to be met.
    Redundant RTS and ESFAS trains are maintained, and diversity 
with regard [to] the signals that provide reactor trip and 
engineered safety features actuation is also maintained. All signals 
credited as primary or secondary, and all operator actions credited 
in the accident analyses will remain the same. The proposed changes 
will not result in plant operation in a configuration outside the 
design basis. The calculated impact on risk is insignificant and 
meets the acceptance criteria contained in RGs 1.174 and 1.177. 
Although there was no attempt to quantify any positive human factors 
benefit due to increased CTs and bypass test times, it is expected 
that there would be a net benefit due to a reduced potential for

[[Page 16623]]

spurious reactor trips and actuations associated with testing.
    Implementation of the proposed changes is expected to result in 
an overall improvement in safety, as follows:
    (a) Reduced testing will result in fewer inadvertent reactor 
trips, less frequent actuation of ESFAS components, less frequent 
distraction of operations personnel without significantly affecting 
RTS and ESFAS reliability.
    (b) Improvements in the effectiveness of the operating staff in 
monitoring and controlling plant operation will be realized. This is 
due to less frequent distraction of the operators and shift 
supervisor to attend to instrumentation Required Actions with short 
CTs.
    (c) Longer repair times associated with increased CTs will lead 
to higher quality repairs and improved reliability.
    (d) The CT extensions for the reactor trip breakers will provide 
additional time to complete test and maintenance activities while at 
power, potentially reducing the number of forced outages related to 
compliance with reactor trip breaker CT, and provide consistency 
with the CT for the logic trains.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: January 28, 2004.
    Description of amendment request: Southern California Edison (SCE) 
permanently shutdown San Onofre Nuclear Generating Station (SONGS), 
Unit 1, in November 1992. Active decommissioning of SONGS Unit 1 began 
in June 1999. As part of decommissioning, SCE constructed an 
Independent Spent Fuel Storage Installation (ISFSI) at SONGS for dry 
cask storage of spent fuel. In March 2004, SCE plans to begin moving 
the spent fuel located in the Unit 1 spent fuel pool into the ISFSI. 
SCE has proposed to eliminate License information and technical 
specifications which will no longer be applicable following the 
transfer of the last fuel assembly from the Unit 1 spent fuel pool to 
the ISFSI.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. This proposed change provides the necessary requirements for 
Unit 1 with no spent fuel located in the spent fuel pool. With no 
spent fuel located at Unit 1, the probability and consequence of the 
fuel handling accident are removed.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different type of accident 
from any accident previously evaluated?
    No. These changes provide the necessary requirements for SONGS 
Unit 1 with no spent fuel in the spent fuel pool. With no spent fuel 
located at Unit 1, there is no possibility of a new or different 
kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety?
    No. These changes provide the necessary requirements for SONGS 
Unit 1 with no spent fuel in the spent fuel pool. With no spent fuel 
located at Unit 1, the fuel handling accident is not applicable and 
there is impact on the margin of safety.
    Therefore, 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 
amendment request involves 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: Mark Thaggard.

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

    Date of amendment request: September 19, 2003.
    Description of amendment request: The proposed change will revise 
the Technical Specifications (TSs) Surveillance Requirement (SR) 
4.2.4.2, to reflect the use of the Power Distribution Monitoring System 
(PDMS) for a core power distribution measurement. This change will also 
result in revising the Bases for 3/4.2.4 to reflect the use of the 
PDMS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed change to TS 4.2.4.2 clarifies the use of the 
PDMS as means of measuring core power distribution with one Power 
Range Channel inoperable to determine if QPTR [Quadrant Power Tilt 
Ratio] is within the limit. The use of PDMS was approved in 
Amendment 142 and added as TS 3.3.3.11. This clarification of its 
use in TS 4.2.4.2 specifies an additional method of performing the 
surveillance requirement and will not increase the probability of an 
accident previously evaluated.
    The probability or consequences of accidents previously 
evaluated in the VCSNS FSAR [Final Safety Analysis Report] are 
unaffected by this proposed change because there is no change to any 
equipment response or accident mitigation scenario. There are no 
additional challenges to fission product barrier integrity. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new 
different kind of accident from any previously evaluated?
    No. The proposed change to TS 4.2.4.2 clarifies the use of the 
PDMS as means of measuring core power distribution with one Power 
Range Channel inoperable to determine if QPTR is within the limit. 
The use of the PDMS was approved in Amendment 142 and added as TS 
3.3.3.11. This clarification of its use in TS 4.2.4.2 specifies an 
additional method of performing the surveillance requirement and 
does not create the possibility of a new different kind of accident 
or malfunction.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed change. 
The proposed change does not challenge the performance or integrity 
of any safety-related system. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Does this change involve a significant reduction in margin of 
safety?
    No. 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. A change to the surveillance 
requirement is proposed; however, this clarification of the use of 
PDMS in TS 4.2.4.2 specifies an additional method of performing the 
surveillance requirement.

    The NRC staff has reviewed the licencee's analysis and based on 
this

[[Page 16624]]

review, it appears that the three standards of 10 CFR 50.92 are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Thomas G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: March 10, 2004.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification allowable value for the spent fuel 
pool area radiation monitors.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed technical specification (TS) change to reduce 
the allowable value for the spent fuel pool area radiation monitors 
does not change any operator actions nor does it change plant 
systems or structures. Therefore, the proposed change does not 
result in a significant increase in the probability of a Fuel 
Handling Accident (FHA). The surveillance requirement radiation 
limit for the spent fuel pool area radiation monitors will be 
lowered to compensate for the change in source terms which resulted 
from the methodology change due to discovery of a modeling error. 
This change ensures the monitors perform their safety function of 
limiting the site boundary dose to a small fraction of the 10 CFR 
part 100 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?
    No. The proposed TS change does not alter the function of the 
spent fuel monitors which is to initiate ABGTS [Auxiliary Building 
Gas Treatment System actuation] upon an FHA. The TS allowable value 
and the associated setpoints for the spent fuel pool area radiation 
monitors will be lowered due to calculation methodology changes 
resulting from discovery of a modeling error. The change will not 
result in the installation of any new equipment or system. No new 
operations procedures, conditions, or modes will be created by this 
proposed change. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in 
margin of safety?
    No. The method for calculating the radiological consequences are 
revised for calculating the safety limit of the spent fuel pool area 
radiation monitors to correctly account for isotopic release 
fractions. The monitors' setpoints are based on 30 rem thyroid at 
the site boundary resulting from an unfiltered release. At the 
monitor setpoint, the monitors initiate ABGTS and thus the release 
is filtered. The radiological dose consequences do not change and 
remain less than a small fraction of the dose limit identified in 10 
CFR 100. The surveillance requirement is being reduced for 
consistency with calculation methodology changes and to ensure the 
monitors perform their intended design function of limiting the site 
boundary dose to less than 30 rem thyroid subsequent to an FHA. 
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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: William F. Burton, Acting.

Virginia Electric and Power Company, Docket No. 50-339, North Anna 
Power Station, Unit No. 2, Louisa County, Virginia

    Date of amendment request: January 23, 2004.
    Description of amendment request: The proposed amendment would 
revise Improved Technical Specifications (TS) Surveillance Requirements 
3.5.1.4, 3.5.4.3, and 3.6.7.3 to delete a note that differentiates 
between the amount of boron concentrations at North Anna Power Station, 
Units 1 and 2.
    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 license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The proposed changes to TS Surveillance Requirements 3.5.1.4, 
3.5.4.3, and 3.6.7.3 delete a note that is no longer necessary and 
do not alter any plant equipment or operating practices in such a 
manner that the probability of an accident is increased. The 
proposed changes will not alter assumptions relative to the 
mitigation of an accident or transient event.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    The proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or 
changes in the methods governing normal plant operation. Therefore, 
the proposed changes do 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?
    The proposed changes do not alter the boron concentrations in 
the safety injection accumulators, RWST [refueling water storage 
tank], and casing cooling tank. The proposed changes to TS 
Surveillance Requirements 3.5.1.4, 3.5.4.3, and 3.6.7.3 are 
considered administrative in nature. Therefore, the proposed changes 
do not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Section Chief: John A. Nakoski.

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

[[Page 16625]]

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, located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by e-mail to pdr@nrc.gov.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2 (HBRSEP2), Darlington County, South Carolina

    Date of application for amendment: December 3, 2003, as 
supplemented January 14 and February 6, 2004.
    Brief description of amendment: The amendment eliminates a license 
condition that limits HBRSEP2 operation to 504 effective full-power 
days. This license condition was added in License Amendment No. 196, 
issued on November 5, 2002.
    Date of issuance: March 10, 2004.
    Effective date: March 10, 2004.
    Amendment No. 200.
    Facility Operating License No. DPR-23: Amendment revises Appendix 
B, ``Additional Conditions,'' to the Facility Operating License.
    Date of initial notice in Federal Register: February 3, 2004 (69 FR 
5201). The February 6, 2004, supplemental letter provided clarifying 
information only and did not change the initial proposed no significant 
hazards consideration determination or expand the scope of the initial 
application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 10, 2004.
    No significant hazards consideration comments received: No.
    Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan
    Date of application for amendment: October 22, 2003.
    Brief description of amendment: The amendment deletes requirements 
from the Technical Specifications to maintain hydrogen recombiners and 
hydrogen and oxygen monitors.
    Date of issuance: March 15, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 159.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 3, 2004 (69 FR 
5202).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 15, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 7, 2003 as supplemented 
September 18, 2003.
    Brief description of amendment: The amendment changed Technical 
Specifications (TSs) affecting cycle-specific parameters that will be 
relocated to the Core Operating Limits Report.
    Date of issuance: March 9, 2004.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days from the date of issuance.
    Amendment No.: 218.
    Facility Operating License No. NPF-49: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28849). The September 18, 2003 supplement contained clarifying 
information and did not change the staff's proposed finding of no 
significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 9, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 14, 2002, supplemented 
by letters dated September 11, 2003, and March 10, 2004.
    Brief description of amendments: The amendments revised the 
Technical Specification 3.3.2, ``Engineered Safety Features Actuation 
System Instrumentation.''
    Date of issuance: March 16, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 220 & 202.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 19, 2003 (68 FR 
49815).
    The supplements dated September 11, 2003, and March 10, 2004, 
provided clarifying information that did not change the scope of the 
November 14, 2003, application nor the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 16, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 19, 2003.
    Brief description of amendment: The amendment deletes the 
requirements from the Technical Specifications to maintain hydrogen 
recombiners and hydrogen analyzers.
    Date of issuance: March 9, 2004.
    Effective date: As of the date of issuance and shall be implemented 
120 days from the date of issuance.
    Amendment No.: 192.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register:  January 20, 2004 (69 
FR 2741).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 9, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 18, 2003.
    Brief description of amendments: The amendments revise Appendix A, 
Technical Specifications (TS), of Facility Operating License Nos. NPF-
11 and NPF-18. Specifically, the change modifies TS Table 3.3.6.1-1, 
``Primary

[[Page 16626]]

Containment Isolation Instrumentation,'' to add the requirement to 
perform a Channel Check in accordance with Surveillance Requirement 
(SR) 3.3.6.1.1 to thirteen listed instrument functions. The change is 
the result of the replacement of existing plant equipment with 
equipment that has the capability of permitting the performance of a 
Channel Check with the plant in MODES 1, 2, and 3. The change is 
consistent with the wording specified in NUREG-1434, ``Standard 
Technical Specifications General Electric Plants, BWR/6,'' Revision 2, 
dated June 2001.
    Date of issuance: March 5, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 166/152.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register:  June 10, 2003 (68 FR 
34667).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 5, 2004.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-352, Limerick Generating 
Station, Unit 1, Montgomery County, Pennsylvania

    Date of application for amendment: December 22, 2003, as 
supplemented by letter dated February 13, 2004. The February 13, 2004, 
submittal provided clarifying information and did not change the 
staff's proposed finding of no significant hazards.
    Brief description of amendment: This amendment revised the safety 
limit minimum critical power ratio value in TS 2.1 with the reactor 
steam dome pressure greater than 785 psig and core flow greater than 
10% of rated core flow from the current specification of 1.10 to 1.07 
for two recirculation-loop operation and from 1.11 to 1.08 for single 
recirculation-loop operation.
    Date of issuance: March 12, 2004.
    Effective date: As of date of issuance and shall be implemented 
within 30 days.
    Amendment No. 170.
    Facility Operating License No. NPF-39. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 3, 2004 (69 FR 
5203).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 12, 2004.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of application for amendments: February 27, 2003, as 
supplemented by letters dated April 11 and August 5, 2003.
    Brief description of amendments: The amendments revise the 
Technical Specifications to allow a one-time change in the containment 
Type A integrated leakage rate test interval that extends the test 
interval from 10 to 15 years.
    Date of issuance: March 8, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 220/214.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 1, 2003 (68 FR 
15759). The April 11 and August 5, 2003, submittals provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 8, 2004.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-265, Quad Cities Nuclear 
Power Station, Unit 2, Rock Island County, Illinois

    Date of application for amendment: November 14, 2003, as 
supplemented by letters dated December 23, 2003, and January 7, 2004.
    Brief description of amendment: The amendment revises the values 
and wording of the Technical Specifications safety limit minimum 
critical power ratio (SLMCPR).
    Date of issuance: March 10, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 215.
    Facility Operating License No. DPR-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 20, 2004 (69 FR 
2743). The December 23, 2003, and January 7, 2004, submittals provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 10, 2004.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of application for amendments: November 21, 2003.
    Brief description of amendments: These amendments allow transfer of 
the requirements of Technical Specifications (TSs) 6.5 (Review and 
Audit), 6.8.2 and 6.8.3 (Procedures and Programs Review Specifics), and 
6.10 (Record Retention) to the St. Lucie Plant's Quality Assurance Plan 
(a licensee-controlled document).
    Date of Issuance: March 11, 2004.
    Effective Date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 189 & 133.
    Renewed Facility Operating License Nos. DPR-67 and NPF-16: 
Amendments revised the TSs.
    Date of initial notice in Federal Register: January 6, 2004 (69 FR 
698).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 11, 2004.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket No. 50-387, Susquehanna Steam Electric 
Station, Unit 1, Luzerne County, Pennsylvania

    Date of application for amendments: July 1, 2003, as supplemented 
by letters dated November 17 and December 22, 2003.
    Brief description of amendments: The amendment revised the values 
of the Safety Limit for Minimum Critical Power Ratio in the Unit 1 
Technical Specifications (TSs) 2.1.1.2, clarified fuel design features 
in TS 4.2.1, and updated the references used to determine the core 
operating limits in TS 5.6.5.b.
    Date of issuance: March 9, 2004.
    Effective date: As of the date of issuance and shall be implemented 
upon startup following the thirteenth refueling and inspection outage.
    Amendment Nos.: 216.
    Facility Operating License No. NPF-14: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 5, 2003 (68 FR 
46245).
    The supplements dated November 17 and December 22, 2003, 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.

[[Page 16627]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 9, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 3, 2003, as 
supplemented February 9, 2004.
    Brief description of amendments: The amendments revised the 
Technical Specifications. to add a Limiting Condition for Operation 
(LCO) for the Liner Heat Generation Rate. The new LCO is included in 
Section 3.2, Power Distribution Limits. The proposed amendments would 
also change the recirculation loop LCO, Section 5.6.5, and the 
appropriate Bases.
    Date of issuance: March 8, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 239 / 182.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2003 (68 
FR 64128).
    The supplement dated February 9, 2004, provided clarifying 
information that did not change the scope of the October 3, 2003, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 8, 2004.
    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: December 18, 2003.
    Brief description of amendments: The amendments modified Technical 
Specification 3.9.6 to correct completion times of ACTIONS B.2 and B.3, 
which were overlooked in Amendment No. 105.
    Date of issuance: March 5, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 110 and 110.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 3, 2004 (69 FR 
5209).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 5, 2004.
    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: March 18, 2003, as supplemented by 
letter dated August 14, 2003.
    Brief description of amendments: The amendments modified Technical 
Specifications (TS) to permanently except seven containment isolation 
valves in each unit, in residual heat removal and containment spray 
systems, from local leakage rate testing requirements of 10 CFR Part 
50, Appendix J.
    Date of issuance: March 5, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 111 and 111.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 15, 2003 (68 FR 
8289).
    The August 14, 2003, supplemental letter provided clarifying 
information and did not change the scope of the original Federal 
Register notice or staff's original no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 5, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 27, 2003, as supplemented 
by letter dated December 12, 2003.
    Brief description of amendment: The amendment (1) revises the 
definition of dose equivalent radioiodine 131 (I-131), and (2) 
increases the maximum allowed closure time of each main feedwater 
isolation valve (MFIV) from 5 seconds to 15 seconds. A plant 
modification would replace the electro-hydraulic MFIV actuators with 
system-medium actuators to improve MFIV reliability and reduce 
maintenance requirements, and the MFIV stroke time would be increased. 
A plant modification would also replace swing check valves in each 
auxiliary feedwater (AFW) motor-driven pump discharge line with an 
automatic recirculation control check valve to reduce the potential for 
vibration and increase AFW flow margin. The NRC also approves the re-
analysis of the steam generator tube rupture with overfill accident 
submitted in the application.
    Date of issuance: March 11, 2004.
    Effective date: March 11, 2004, and shall be implemented prior to 
the entry into Mode 3 in the restart of the Callaway Plant from the 
Refueling Outage (RO) 13, which is scheduled for April 2004.
    Amendment No.: 159.
    Facility Operating License No. NPF-30: The amendment revises the 
Technical Specifications and updates the Final Safety Analysis Report.
    Date of initial notice in Federal Register: July 22, 2003 (68 FR 
43394).
    The additional information provided in the supplemental letter does 
not expand the scope of the application as noticed and does not change 
the staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 11, 2004.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 19th day of March, 2004.

    For the Nuclear Regulatory Commission.
Edwin M. Hackett,
Director, Acting, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 04-6682 Filed 3-29-04; 8:45 am]
BILLING CODE 7590-01-P