[Federal Register Volume 70, Number 20 (Tuesday, February 1, 2005)]
[Notices]
[Pages 5233-5254]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-1574]


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NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility Operating 
Licenses


Involving No Significant Hazards Considerations

I. Background

    Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from January 7, 2005, through January 19, 2005. 
The last biweekly notice was published on January 18, 2005 (70 FR 
2886).

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

[[Page 5234]]

operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be

[[Page 5235]]

transmitted either by means of facsimile transmission to (301) 415-3725 
or by e-mail to [email protected]. A copy of the request for 
hearing and petition for leave to intervene should also be sent to the 
attorney for the licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of amendment request: September 15, 2004.
    Description of amendment request: The proposed amendment would 
delete requirements from the Technical Specifications (TSs) to maintain 
hydrogen recombiners and hydrogen and oxygen monitors. A notice of 
availability for this TS improvement using the consolidated line item 
improvement process was published in the Federal Register on September 
25, 2003 (68 FR 55416). Licensees were generally required to implement 
upgrades as described in NUREG-0737, ``Clarification of TMI [Three Mile 
Island] Action Plan Requirements,'' and Regulatory Guide (RG) 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
combustible gas control were imposed by order for many facilities and 
were added to, or included, in the TSs for nuclear power reactors 
currently licensed to operate. The revised Title 10 of the Code of 
Federal Regulations (10 CFR) Section 50.44, ``Standards for combustible 
gas control system in light-water-cooled power reactors,'' eliminated 
the requirements for hydrogen recombiners and relaxed safety 
classifications and licensee commitments to certain design and 
qualification criteria for hydrogen and oxygen monitors.
    The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice 
of availability of a model no significant hazards consideration (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 September 15, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:

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

    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The NRC has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen and oxygen monitors are no longer required to mitigate 
design-basis accidents and, therefore, the hydrogen monitors do not 
meet the definition of a safety-related component as defined in 10 
CFR 50.2. RG 1.97 Category 1, is intended for key variables that 
most directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen and oxygen monitors no 
longer meet the definition of Category 1 in RG 1.97. As part of the 
rulemaking to revise 10 CFR 50.44 the NRC found that Category 3, as 
defined in RG 1.97, is an appropriate categorization for the 
hydrogen monitors because the monitors are required to diagnose the 
course of beyond design-basis accidents. Also, as part of the 
rulemaking to revise 10 CFR 50.44, the NRC found that Category 2, as 
defined in RG 1.97, is an appropriate categorization for the oxygen 
monitors, because the monitors are required to verify the status of 
the inert containment.
    The regulatory requirements for the hydrogen and oxygen monitors 
can be relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, 
[classification of the oxygen monitors as Category 2,] and removal 
of the hydrogen and oxygen monitors from TS will not prevent an 
accident management strategy through the use of the severe accident 
management guidelines, the emergency plan, the emergency operating 
procedures, and site survey monitoring that support modification of 
emergency plan protective action recommendations.
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen and oxygen 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 and oxygen monitor requirements, 
including removal of these requirements from TS, will not result in 
any failure mode not previously analyzed. The hydrogen recombiner 
and hydrogen and oxygen monitor equipment was intended to mitigate a 
design-basis hydrogen release. The hydrogen recombiner and hydrogen 
and oxygen monitor equipment are not considered accident precursors, 
nor does their existence or elimination have any adverse impact on 
the pre-accident state of the reactor core or post accident 
confinement of radionuclides within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen and oxygen monitor requirements, 
including removal of these requirements from TS, in light of 
existing plant equipment, instrumentation, procedures, and programs 
that provide effective mitigation of and recovery from reactor 
accidents, results in a neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The NRC has found that this hydrogen 
release is not risk-significant

[[Page 5236]]

because the design-basis LOCA hydrogen release does not contribute 
to the conditional probability of a large release up to 
approximately 24 hours after the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI, Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Category 2 oxygen monitors are adequate to verify the status of 
an inerted containment.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. The intent of the requirements established 
as a result of the TMI, Unit 2 accident can be adequately met 
without reliance on safety-related oxygen monitors. Removal of 
hydrogen and oxygen monitoring from TS will not result in a 
significant reduction in their functionality, reliability, and 
availability.

    The NRC staff proposes to determine that the amendment request 
involves NSHC.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60666.
    NRC Section Chief: Gene Y. Suh.

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

    Date of amendments request: December 16, 2004.
    Description of amendments request: The requested change will delete 
Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure 
Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated December 16, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    Based upon the reasoning presented above, the requested change does 
not involve significance hazards consideration.
    Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona 
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix, 
Arizona 85072-2034.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendments request: December 1, 2004.
    Description of amendments request: The requested change will delete 
Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure 
Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated December 1, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    Based upon the reasoning presented above, the requested change does 
not involve significance hazards consideration.
    Attorney for licensee: James M. Petro, Jr., Esquire, Counsel, 
Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor, 
Baltimore, MD 21202.
    NRC Section Chief: Richard J. Laufer.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: December 6, 2004.
    Description of amendment request: The requested change will delete

[[Page 5237]]

Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure 
Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated December 6, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    Based upon the reasoning presented above, the requested change does 
not involve significant hazards consideration.
    Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279 .
    NRC Section Chief: M. Kotzalas (Acting).

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

    Date of amendment request: June 6, 2004.
    Description of amendment request: The proposed change would modify 
the Millstone Power Station, Unit No. 2 Technical Specifications (TSs) 
to extend the 10-year test interval for the Integrated Leakage Rate 
Test program to 15 years from the last Type A test. Specifically, the 
proposed change would revise TS 6.19, ``Containment Leakage Rate 
Testing [CLRT] Program,'' and permit a one-time, 5-year extension of 
the 10-year performance-based Type A test interval. In addition, the 
testing would be in accordance with the CLRT Program, Regulatory Guide 
(RG) 1.163, ``Performance-Based Containment Leak-Test Program'' and 
surveillance testing requirements as proposed in Nuclear Energy 
Institute 94-01 for Type A testing.
    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 extension to Type A testing cannot increase the 
probability of an accident previously evaluated since extension of 
the containment Type A testing is not a physical plant modification 
that could alter the probability of accident occurrence, nor is it 
an activity or modification that by itself could lead to equipment 
failure or accident initiation.
    The proposed one-time, five-year extension to Type A testing 
does not result in a significant increase in the consequences of an 
accident as documented in NUREG-1493. The NUREG notes that very few 
potential containment leakage paths are not identified by Type B and 
C tests. It concludes that even reducing the Type A (ILRT 
[integrated leak rate test]) testing frequency to once per twenty 
years leads to an imperceptible increase in risk.
    DNC (the licensee) provides a high degree of assurance through 
indirect testing and inspection that the containment will not 
degrade in a manner detectable only by Type A testing. The last two 
Type A tests identified containment leakage within acceptance 
criteria, indicating a very leak-tight containment. Inspections 
required by the ASME Code [American Society of Mechanical Engineers 
Boiler and Pressure Vessel Code] are also performed in order to 
identify indications of containment degradation that could affect 
leak-tightness. Separately, Type B and C testing required by 
Technical Specifications, identifies any containment opening from 
design penetrations, such as valves, that would otherwise be 
detected by a Type A test. These factors establish that a one-time, 
five-year extension to the Millstone Unit 2 Type A test interval 
will not represent a significant increase in the consequences of an 
accident.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    The proposed revision to the Technical Specifications adds a 
one-time extension to the current interval for Type A testing for 
Millstone Unit 2. The current test interval of ten years, based on 
past performance, would be extended on a one-time basis to fifteen 
years from the last Type A test. The proposed extension to Type A 
testing does not create the possibility of a new or different type 
of accident since there are no physical changes being made to the 
plant and there are no changes to the operation of the plant that 
could introduce a new failure.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The proposed revision to Millstone Unit 2 Technical 
Specifications adds a one-time extension to the current interval for 
Type A testing. The current test interval of ten years, based on 
past performance, would be extended on a one-time basis to fifteen 
years from the last Type A test for Millstone Unit 2. RG 1.174 
provides guidance for determining the risk impact of plant-specific 
changes to the licensing basis. RG 1.174 defines very small changes 
in risk as resulting in increases of CDF [core damage frequency] 
below 10-\6\/yr and increases in LERF [large early 
release frequency] below 10-\7\/yr. Since the ILRT does 
not impact CDF, the relevant criterion is LERF. The increase in 
LERF, resulting from a change in the Type A ILRT test interval from 
a once-per-ten-years to a once-per-fifteen-years is 0.83 x 
10-\8\/yr, based on internal events. Since guidance in 
Reg. Guide 1.174 defines very small changes in LERF as below 
10-\7\/yr, increasing the ILRT interval from ten to 
fifteen years is, therefore, considered non-risk significant and 
will not significantly reduce the margin of safety. The NUREG-1493 
generic study of the effects of extending containment leakage 
testing found that a 20-year interval in Type A leakage testing 
resulted in an imperceptible increase in risk to the public. NUREG-
1493 generically concludes that the design containment leakage rate 
contributes about 0.1 percent of the overall risk. Decreasing the 
Type A testing frequency would have a minimal effect on this risk 
since 95% of the Type A detectable leakage paths would already be 
detected by Type B and C testing. Given that the proposed change 
will continue to meet the current design basis, any reduction in a 
margin of safety would not be significant.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three

[[Page 5238]]

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: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: Darrell J. Roberts.

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

    Date of amendment request: December 16, 2004.
    Description of amendment request: The proposed amendment would 
revise the current fuel rod average licensing basis burnup limit for 
one lead test assembly (LTA) containing advanced zirconium based alloys 
to a limit not exceeding 71,000 MWD/MTU.
    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.
    The Westinghouse LTA is very similar in design to the 
Westinghouse fuel that comprises the remainder of the core. The 
reload core design for Millstone Unit 3 Cycle 12, where one LTA will 
operate to high burnup, will meet all applicable design criteria. 
The performance of the Emergency Core Cooling System will not be 
affected by the operation of the LTA and operation of the LTA to 
high burnup will not result in a change to the Millstone Unit 3 
reload design and safety analysis limits. Operation of one 
Westinghouse LTA to high burnup will not result in a measurable 
impact on normal operating releases, and will not increase the 
predicted radiological consequences of accidents postulated in 
Chapter 15 of the Millstone FSAR [final safety analysis report]. 
Therefore, neither the probability of occurrence nor the 
consequences of any accident previously evaluated is significantly 
increased.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The Westinghouse LTA is very similar in design (both mechanical 
and composition of materials) to the resident Westinghouse fuel. All 
design and performance criteria will continue to be met and no new 
single failure mechanisms will be created. The irradiation of one 
LTA to high burnup does not involve any alteration to plant 
equipment or procedures, which would introduce any new or unique 
operational modes or accident precursors. Therefore, the possibility 
for a new or different kind of accident from any accident previously 
evaluated is not created.
    3. Involve a significant reduction in a margin of safety.
    The operation of one Westinghouse LTA to high burnup does not 
change the performance requirements of any system or component such 
that any design criteria will be exceeded. The normal limits on core 
operation defined in the Millstone Unit 3 Technical Specifications 
will remain applicable for the core in which the high burnup 
assembly is irradiated. Therefore, the margin of safety as defined 
in the Bases to the Millstone Unit 3 Technical Specifications is not 
significantly reduced.

    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: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
    NRC Section Chief: Darrell Roberts.

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423, 
Millstone Power Station, Unit Nos. 2 and 3, New London County, 
Connecticut

    Date of amendment request: September 8, 2004.
    Description of amendment request: The proposed amendment deletes 
the requirements from the technical specifications (TSs) to maintain 
hydrogen recombiners and hydrogen monitors. Licensees were generally 
required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' 
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' Implementation of these upgrades 
was an outcome of the lessons learned from the accident that occurred 
at TMI Unit 2. Requirements related to combustible gas control were 
imposed by Order for many facilities and were added to or included in 
the TSs for nuclear power reactors currently licensed to operate. The 
revised 10 CFR 50.44, ``Standards for Combustible Gas Control System in 
Light-Water-Cooled Power Reactors,'' eliminated the requirements for 
hydrogen recombiners and relaxed safety classifications and licensee 
commitments to certain design and qualification criteria for hydrogen 
and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration determination for referencing in 
license amendment applications in the Federal Register on September 25, 
2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated September 8, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

[[Page 5239]]

    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TSs, does not involve a 
significant increase in the probability or the consequences of any 
accident previously evaluated.

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

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

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: Darrell J. Roberts.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina; Docket Nos. 
50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, 
Oconee County, South Carolina

    Date of amendment request: September 20, 2004.
    Description of amendment request: The proposed amendment deletes 
the requirements from the technical specifications (TS) to maintain 
hydrogen recombiners (McGuire only) and hydrogen monitors (McGuire and 
Oconee). Licensees were generally required to implement upgrades as 
described in NUREG-0737, ``Clarification of TMI [Three Mile Island] 
Action Plan Requirements,'' and Regulatory Guide 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 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 September 20, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

[[Page 5240]]

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

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

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

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

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: October 22, 2004.
    Description of amendment request: The proposed amendments would 
delete the requirements from the Technical Specifications (TSs) to 
maintain hydrogen recombiners and hydrogen monitors. Licensees were 
generally required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' 
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' Implementation of these upgrades 
was an outcome of the lessons learned from the accident that occurred 
at TMI Unit 2. Requirements related to combustible gas control were 
imposed by Order for many facilities and were added to or included in 
the TSs for nuclear power reactors currently licensed to operate. The 
revised 10 CFR 50.44, ``Standards for Combustible Gas Control System in 
Light-Water-Cooled Power Reactors,'' eliminated the requirements for 
hydrogen recombiners and relaxed safety classifications and licensee 
commitments to certain design and qualification criteria for hydrogen 
and oxygen monitors.
    The Nuclear Regulatory Commission (NRC) staff issued a notice of 
availability of a model no significant hazards consideration (NSHC) 
determination for referencing in license amendment applications in the 
Federal Register on September 25, 2003 (68 FR 55416). The licensee 
affirmed the applicability of the model NSHC determination in its 
application dated October 22, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

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

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

[[Page 5241]]

of radionuclides within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

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

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: October 25, 2004.
    Description of amendment request: The requested change will delete 
Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure 
Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated October 25, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    Based upon the reasoning presented above, the requested change does 
not involve significance hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: December 30, 2004.
    Description of amendment request: The proposed amendment would 
revise a Technical Specification (TS) surveillance requirement (SR) in 
TS 3.1.4, ``Control Rod Scram Times.'' Specifically, the proposed 
change would revise the frequency for SR 3.1.4.2, ``Control Rod Scram 
Time Testing,'' from ``120 days cumulative operation in MODE 1'' to 
``200 days cumulative operation in MODE 1.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in licensing amendment applications in the Federal Register on August 
23, 2004 (69 FR 51864). The licensee affirmed the applicability of the 
model NSHC determination in its application dated December 30, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.

    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The frequency 
of surveillance testing is not an initiator of any accident 
previously evaluated. The frequency of surveillance testing does not 
affect the ability to mitigate any accident previously evaluated, as 
the tested component is still required to be operable. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change does not result in any new or different modes of plant 
operation. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time

[[Page 5242]]

testing from every 120 days of cumulative Mode 1 operation to 200 
days of cumulative Mode 1 operation. The proposed change continues 
to test the control rod scram time to ensure the assumptions in the 
safety analysis are protected. Therefore, the proposed change does 
not involve a significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: December 20, 2004.
    Description of amendment request: The proposed amendment would 
increase the lifting tripod's rating from 150 tons to 190 tons. This 
would allow for additional flexibility when lifting the new reactor 
vessel head during refueling outages.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The ANO-1 [Arkansas Nuclear One, Unit 1] Tripod does not perform 
a safety function required by 10 CFR [Part] 50. The Tripod serves to 
perform heavy load movements during refueling outages[,] including 
[movement of] the reactor vessel head. Safe load paths have been 
established in accordance with NUREG-0612[, ``Control of Heavy Loads 
at Nuclear Power Plants,''] to ensure that the fuel and safety[-
]related equipment required to be inservice are protected. Use of 
actual Tripod eyelet Certified Material Test Reports (CMTRs) 
demonstrates that a safety factor of 3 to yield is maintained and 
that the lifting devices will perform their design function under 
maximum lifted loads. The Tripod does not serve any mitigative 
functions to lessen accidents.
    Therefore, the proposed change does not affect the probability 
or consequences of any ANO-1 analyzed accidents.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The only time that the Tripod is performing heavy loads 
movements is during Refueling operations. Safe load paths and load 
drop analyses have been performed to assure that heavy loads 
movements will not cause fuel damage or cause safety[-]related 
equipment to become inoperable. The proposed use of CMTRs instead of 
minimum yield strength of the material still assures that the Tripod 
will perform its required function to not create an accident. In 
addition, there is no change to the operation of the Tripod that 
would create a new failure mode or possible accident.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The design margin for the Tripod is established by NUREG-0612 
and ANSI [American National Standards Institute] N14.6-1978[, 
``Special Lifting Devices for Shipping Containers Weighing 10,000 
Pounds or More for Nuclear Materials'']. A factor of safety of 3 for 
yield strength and 5 for ultimate strength for both the static and 
dynamic load factors is required to be met. These factors of safety 
provide sufficient margin to assure that the Tripod will perform its 
design function of maximum lifted loads. In addition, the use [of] a 
dynamic load factor of 1.15 above the static load is well above the 
actual dynamic factor to be experienced from the design lift speed 
of the polar crane. The use of CMTRs does not result in a 
significant reduction in the margin of safety of the Tripod. In 
addition, the Tripod will be load tested to 150% [percent] of its 
design static and dynamic loading which will further assure adequate 
safety margin.
    Therefore, the margin of safety is not changed by the proposed 
change to the ANO-1 SAR [Safety Analysis Report].

    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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of amendment request: December 17, 2004.
    Description of amendment request: The proposed change will revise 
the air lock surveillance test acceptance criteria to be consistent 
with the NRC approved Industry Technical Specification Task Force 
(TSTF) change to the Standard Technical Specifications (STS), TSTF-52, 
entitled ``Implement 10 CFR [Part] 50, Appendix J, Option B.'' By 
letter dated April 6, 1998, the NRC Staff issued amendment number 135 
to the GGNS license permitting the implementation of the containment 
leak rate testing provisions of 10 CFR Part 50, Appendix J, Option B.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Primary containment air lock leak rate testing can have no 
effect on the probability of any postulated accident. The proposed 
change will increase the allowed containment air lock leakage rate 
and convert it from an absolute leakage rate to a percentage of the 
overall primary containment leakage rate. No change to the overall 
leakage rate of the containment is being proposed, therefore there 
is no change to the consequences of any postulated accident. The 
change in air lock leakage rate will not impact the design or 
operation of any plant system or component nor will they affect 
initiation or mitigation of any accidents previously analyzed.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The primary containment air locks form part of the primary 
containment pressure boundary. The periodic containment air lock 
leakage rate tests specified in SR 3.6.1.2.1 verifies that the air 
lock leakage does not exceed the allowed fraction of the overall 
primary containment leakage rate. This request involves a change in 
the allowable leakage rate of the primary containment air locks 
without increasing the overall allowed leakage rate of the 
containment. Changing the allowable leakage rate has no influence 
on, nor does it contribute in any way to, the possibility of a new 
or different kind of accident or malfunction from those previously 
analyzed. There will be no effect on the types and amounts of 
overall leakage from the primary containment boundary. The proposed 
amendment will not produce any changes to the design or operation of 
the plant. The method of performing the test is not changed. No new 
accident modes are created by changing the allowable leakage in this 
manner. No safety-related equipment or safety functions are altered 
as a result of this change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?

[[Page 5243]]

    Response: No.
    Air lock integrity and leak tightness are essential for 
maintaining primary containment leakage rate to within limits in the 
event of a design basis accident. The periodic containment air lock 
leakage rate tests verify that the air lock leakage does not exceed 
the allowed fraction of the overall primary containment leakage 
rate. Since no changes are proposed to the maximum allowable primary 
containment leakage rate, the design basis radiological analysis is 
not impacted by this change. The license amendment request removes 
unnecessary conservatism from the testing program and allows 
consistency with current industry practice. Since no changes are 
proposed to the maximum allowable primary containment leakage rate, 
the design basis radiological analysis is not impacted by this 
change.
    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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Michael K. Webb.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois; Docket Nos. 50-237 and 50-249, Dresden Nuclear Power 
Station, Units 2 and 3, Grundy County, Illinois; Docket Nos. 50-373 and 
50-374, LaSalle County Station, Units 1 and 2, LaSalle County, 
Illinois; Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power 
Station, Units 1 and 2, Rock Island County, Illinois

    Date of amendment request: September 15, 2004.
    Description of amendment request: The proposed amendment would 
delete requirements from the Technical Specifications (TSs) to maintain 
hydrogen recombiners and hydrogen and oxygen monitors. Licensees were 
generally required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' 
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' Implementation of these upgrades 
was an outcome of the lessons learned from the accident that occurred 
at TMI, Unit 2. Requirements related to combustible gas control were 
imposed by order for many facilities and were added to, or included, in 
the TSs for nuclear power reactors currently licensed to operate. The 
revised Title 10 of the Code of Federal Regulations (10 CFR) Section 
50.44, `` Combustible gas control for nuclear power reactors,'' 
eliminated the requirements for hydrogen recombiners and relaxed safety 
classifications and licensee commitments to certain design and 
qualification criteria for hydrogen and oxygen monitors.
    The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice 
of availability of a model no significant hazards consideration (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 September 15, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:

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

    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen and oxygen monitors are no longer required to mitigate 
design-basis accidents and, therefore, the hydrogen monitors do not 
meet the definition of a safety-related component as defined in 10 
CFR 50.2. RG 1.97 Category 1, is intended for key variables that 
most directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen and oxygen monitors no 
longer meet the definition of Category 1 in RG 1.97. As part of the 
rulemaking to revise 10 CFR 50.44 the Commission found that Category 
3, as defined in RG 1.97, is an appropriate categorization for the 
hydrogen monitors because the monitors are required to diagnose the 
course of beyond design-basis accidents. Also, as part of the 
rulemaking to revise 10 CFR 50.44, the Commission found that 
Category 2, as defined in RG 1.97, is an appropriate categorization 
for the oxygen monitors, because the monitors are required to verify 
the status of the inert containment.
    The regulatory requirements for the hydrogen and oxygen monitors 
can be relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, 
classification of the oxygen monitors as Category 2, and removal of 
the hydrogen and oxygen monitors from TS will not prevent an 
accident management strategy through the use of the SAMGs [severe 
accident management guidelines], the emergency plan (EP), the 
emergency operating procedures (EOPs), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen and oxygen 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 and oxygen monitor requirements, 
including removal of these requirements from TS, will not result in 
any failure mode not previously analyzed. The hydrogen recombiner 
and hydrogen and oxygen monitor equipment was intended to mitigate a 
design-basis hydrogen release. The hydrogen recombiner and hydrogen 
and oxygen monitor equipment are not considered accident precursors, 
nor does their existence or elimination have any adverse impact on 
the pre-accident state of the reactor core or post accident 
confinement of radionuclides within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen and oxygen 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.

[[Page 5244]]

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

    Based on the reasoning presented above and the previous discussion 
of the amendment request, the requested change does not involve a 
significant hazards consideration.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Gene Y. Suh.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois; 
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station (QCNPS), Units 1 and 2, Rock Island 
County, Illinois

    Date of amendment request: November 4, 2004.
    Description of amendment request: The proposed amendments would 
revise the plant technical specification (TS) pressure and temperature 
(P/T) limit curves for 54 effective full power years (EFPY) to support 
a 20-year license extension for both DNPS and QCNPS to 60 years (i.e., 
54 EFPY), and resolves a non-conservative condition for TS Section 
3.4.9, Figure 3.4.9-2, ``Non-Nuclear Heatup/Cooldown Curve,'' for 
QCNPS.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) section 50.91(a), Exelon Generation Company (EGC) 
has provided its analysis of the issue of no significant hazards 
consideration (NSHC), which is presented below:

    According to 10 CFR 50.92, ``Issuance of amendment,'' paragraph 
(c), a proposed amendment to an operating license involves no 
significant hazards consideration if 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.
    In support of this determination, an evaluation of each of the 
three criteria set forth in 10 CFR 50.92 is provided below regarding 
the proposed license amendment.
    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes request that, for DNPS, Units 2 and 3 and 
QCNPS, Units 1 and 2, P/T limit curves in TS 3.4.9, ``RCS Pressure 
and Temperature (P/T) Limits,'' be revised.
    The P/T limits are prescribed during all operational conditions 
to avoid encountering pressure, temperature, and temperature rate-
of-change conditions that might cause undetected flaws to propagate, 
resulting in non-ductile failure of the reactor coolant pressure 
boundary, which is an unanalyzed condition. The methodology used to 
determine the P/T limits has been approved by the NRC [Nuclear 
Regulatory Commission] and thus is an acceptable method for 
determining these limits. Therefore, the proposed changes do not 
affect the probability of an accident previously evaluated.
    There is no specific accident that postulates a non-ductile 
failure of the reactor coolant pressure (RCP) boundary. The loss of 
coolant accident analyzed for the plant assumes a 4.281 square feet 
complete break of the recirculation pump suction line. The revision 
to the P/T limits does not change this assumption. Thus, the 
radiological consequences of any accident previously evaluated are 
not increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not change the response of plant 
equipment to transient conditions. The proposed changes do not 
introduce any new equipment, modes of system operation, or failure 
mechanisms.
    Non-ductile failure of the RCP boundary is not an analyzed 
accident. The proposed changes to the P/T limits were developed 
using an NRC-approved methodology, and thus the revised limits will 
continue to provide protection against non-ductile failure of the 
RCP boundary.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The margin of safety related to the proposed changes is the 
margin between the proposed P/T limits and the pressures and 
temperatures that would produce nonductile failure of the RCP 
boundary. NRC requirements to protect the integrity of the reactor 
coolant pressure boundary in nuclear power plants is established in 
10 CFR 50, Appendix G, ``Fracture Toughness Requirements,'' which 
requires that the P/T limits for an operating plant be at least as 
conservative as those that would be generated if the methods of 
American Society of Mechanical Engineers, Section XI, Appendix G, 
were applied. The use of an NRC-approved methodology, together with 
conservatively chosen plant-specific input parameters, provides an 
acceptable margin of safety. Therefore, the proposed changes do not 
involve a significant reduction in a margin of safety.
    Based upon the above responses, EGC concluded that the proposed 
amendment presents no significant hazards consideration under the 
standards set forth in 10 CFR 50.92 and, accordingly, a finding of 
no significant hazards consideration is justified.

    The NRC staff has reviewed EGC's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve NSHC.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Gene Y. Suh.

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

    Date of application for amendments: September 15, 2004.
    Description of amendment request: The proposed amendment would 
delete requirements from the Technical Specifications (TSs) to maintain 
containment hydrogen and oxygen monitors. A notice of availability for 
this technical specification improvement using the consolidated line 
item improvement process (CLIIP) was published in the Federal Register 
on September 25, 2003 (68 FR 55416). Licensees were generally required 
to implement upgrades as described in NUREG-0737, ``Clarification of 
TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory 
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Assess Plant and Environs Conditions During and Following an 
Accident.''

[[Page 5245]]

Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
combustible gas control were imposed by Order for many facilities and 
were added to or included in the TSs for nuclear power reactors 
currently licensed to operate. The revised 10 CFR 50.44, ``Standards 
for combustible gas control system in light-water-cooled power 
reactors,'' eliminated the requirements for hydrogen recombiners and 
relaxed safety classifications and licensee commitments to certain 
design and qualification criteria for hydrogen and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (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 
relevant portions of the model NSHC determination (hydrogen and oxygen 
monitors only) in its application dated September 15, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen and oxygen monitors are no longer required to mitigate 
design-basis accidents and, therefore, the hydrogen monitors do not 
meet the definition of a safety-related component as defined in 10 
CFR 50.2. RG [Regulatory Guide] 1.97 Category 1, is intended for key 
variables that most directly indicate the accomplishment of a safety 
function for design-basis accident events. The hydrogen and oxygen 
monitors no longer meet the definition of Category 1 in RG 1.97. As 
part of the rulemaking to revise 10 CFR 50.44 the Commission found 
that Category 3, as defined in RG 1.97, is an appropriate 
categorization for the hydrogen monitors because the monitors are 
required to diagnose the course of beyond design-basis accidents. 
Also, as part of the rulemaking to revise 10 CFR 50.44, the 
Commission found that Category 2, as defined in RG 1.97, is an 
appropriate categorization for the oxygen monitors, because the 
monitors are required to verify the status of the inert containment.
    The regulatory requirements for the hydrogen and oxygen monitors 
can be relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, 
[classification of the oxygen monitors as Category 2,] and removal 
of the hydrogen and oxygen monitors from 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 protective action 
recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen and oxygen 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 and oxygen monitor requirements, 
including removal of these requirements from TS, will not result in 
any failure mode not previously analyzed. The hydrogen recombiner 
and hydrogen and oxygen monitor equipment was intended to mitigate a 
design-basis hydrogen release. The hydrogen recombiner and hydrogen 
and oxygen monitor equipment are not considered accident precursors, 
nor does their existence or elimination have any adverse impact on 
the pre-accident state of the reactor core or post accident 
confinement of radionuclides within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen and oxygen 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.
    Category 2 oxygen monitors are adequate to verify the status of 
an inerted containment.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. The intent of the requirements established 
as a result of the TMI, Unit 2 accident can be adequately met 
without reliance on safety-related oxygen monitors. Removal of 
hydrogen and oxygen monitoring from TS will not result in a 
significant reduction in their functionality, reliability, and 
availability.

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

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

    Date of amendment request: September 21, 2004.
    Description of amendment request: The proposed amendment deletes 
the requirements from the technical specifications (TS) to maintain 
containment 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

[[Page 5246]]

gas control were imposed by Order for many facilities and were added to 
or included in the TS for nuclear power reactors currently licensed to 
operate. The revised 10 CFR 50.44, ``Standards for Combustible Gas 
Control System in Light-Water-Cooled Power Reactors,'' eliminated the 
requirements for hydrogen recombiners and relaxed safety 
classifications and licensee commitments to certain design and 
qualification criteria for hydrogen and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on September 
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the 
relevant portions of the model NSHC determination (hydrogen monitors 
only) in its application dated September 21, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

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

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

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

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

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

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

    Date of amendment request: January 13, 2005.
    Description of amendment request: The proposed change would allow a 
one-time extended allowed outage time (AOT) change to Improved 
Technical Specifications (ITS) 3.5.2, Emergency Core Cooling Systems 
(ECCS)--Operating; 3.6.6, Reactor Building Spray and Containment 
Cooling Systems; 3.7.8, Decay Heat Closed Cycle Cooling Water System 
(DC); and 3.7.10, Decay Heat Seawater System to allow the refurbishment 
of Decay Heat Seawater System Pump RWP-3B online.
    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:

    This request has been evaluated against the standards in 10 CFR 
50.92, and has been determined to not involve a significant hazards 
consideration. In support of this conclusion, the following analysis 
is provided:
    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 the systems described above from 72 hours to 10 
days. These Systems are designed to provide cooling for components 
essential to the mitigation of plant transients and

[[Page 5247]]

accidents. The systems are not initiators of design basis accidents. 
The proposed ITS changes have been evaluated to assess their impact 
on normal operation of the systems affected and to ensure that their 
design basis safety functions are preserved.
    A Probabilistic Safety Assessment (PSA) has been performed to 
assess the risk impact of an increase in Completion Time from 72 
hours to 10 days. Although the proposed one-time change results in 
an increase in Core Damage Frequency (CDF) and Large Early Release 
Frequency (LERF), the value of these increases are considered as 
small (CDF) and very small (LERF) 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 the systems described above from 72 hours to 10 
days.
    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 the systems described above from 72 hours to 10 
days. The proposed change will allow online repair of Decay Heat 
Seawater pump RWP-3B to restore the pump to full qualification which 
will improve its reliability and useful lifetime, 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 a Decay Heat Services Seawater 
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 will continue the performance of 10 
CFR 50.65(a)(4) assessments before performing maintenance or 
surveillance activities and no maintenance activities of other risk 
sensitive equipment beyond that required for the refurbishment 
activity will be scheduled concurrent with the repair activity. 
Other compensatory actions that will be implemented include: 
operator attention to the importance of protecting the operable 
redundant train and support systems will be increased, selection of 
beneficial Makeup Pump configurations, no elective maintenance will 
be scheduled in the switchyard, and the establishment of fire 
watches.
    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 (CDF), and Large Early Release Frequency (LERF), the value 
of these increases is considered as small (CDF) and very small 
(LERF) in the current regulatory guidance.
    Therefore, granting this LAR does not involve a significant 
reduction in the margin of safety.
    Based on the above, Progress Energy Florida, Inc. (PEF) 
concludes that the proposed LAR presents a no significant hazards 
consideration under the standards set forth in 10 CFR 50.92(c), and 
accordingly, a finding of ``no significant hazards consideration'' 
is justified.

    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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Michael L. Marshall.

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

    Date of amendment request: October 29, 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 7 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 October 29, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    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 SDV 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.

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 SDV is 
maintained through administrative controls. In addition, the reactor 
protection system will prevent filling of the 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: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Section Chief: M. Kotzalas (Acting).

[[Page 5248]]

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

    Date of amendment requests: December 27, 2004.
    Description of amendment requests: The requested change will delete 
Technical Specification (TS) 5.7.1.1.a, ``Occupational Radiation 
Exposure Report,'' and TS 5.7.1.4, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated December 27, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    Based upon the reasoning presented above, the requested change does 
not involve significance hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment requests: December 27, 2004.
    Description of amendment requests: The proposed amendments would 
revise the San Onofre Nuclear Generating Station (SONGS), Units 2 and 3 
accident source term used in the design basis radiological consequences 
analyses. These license amendments are requested in accordance with the 
requirements of 10 CFR 50.67, which addresses the use of an Alternative 
Source Term (AST) at operating reactors, and relevant guidance of 
Regulatory Guide 1.183. These license amendments represent full-scope 
implementation of the AST described in Regulatory Guide 1.183.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the Facility Operating Licenses for San 
Onofre Units 2 and 3 credit an Alternative Source Term (AST) for the 
design basis radiological site boundary and control room dose 
analyses. This change represents full scope implementation of the 
AST as described in Regulatory Guide 1.183. The proposed changes to 
the Facility Operating Licenses also expand the allowed use of fuel 
failure estimates by Departure from Nucleate Boiling (DNB) 
statistical convolution methodology from only the reactor coolant 
pump sheared shaft event to the Updated Final Safety Analysis Report 
(UFSAR) Chapter 15 non-Loss-of-Coolant-Accident (LOCA) events that 
assume a loss of flow (i.e., a loss of AC power) and that fail fuel. 
The proposed changes reflect the parameters used in the radiological 
consequences calculations for the LOCA, Fuel Handling Accident 
inside containment (FHA-IC), Fuel Handling Accident in the Fuel 
Handling Building (FHA-FHB) and pre-trip Steam Line Break Outside 
Containment (SLB-OC).
    The purpose of this proposed change is to change the design 
requirements for the Control Room Envelope (CRE). This proposed 
change will allow an increase in the assumed amount of unfiltered 
air inleakage through the CRE. Currently, design basis radiological 
consequence analyses assume CRE inleakage of 0 cfm, plus an assumed 
10 cubic feet per minute (cfm) inleakage due to ingress and egress 
into the Control Room. Analyses to support this change demonstrate 
acceptable post-accident dose consequences in the Control Room 
assuming 990 cfm of CRE inleakage (plus 10 cfm due to ingress and 
egress for a total of 1000 cfm).
    This proposed change does not affect the precursors for 
accidents or transients analyzed in Chapter 15 of the San Onofre 
Units 2 and 3 UFSAR. Therefore, there is no increase in the 
probability of accidents previously evaluated. The probability 
remains the same because the accident analyses performed involve no 
change to a system, component or structure that affects initiating 
events for any UFSAR Chapter 15 accident evaluated.
    A re-analysis of the UFSAR Chapter 15 LOCA, SLB-OC, FHA-IC, and 
FHA-FHB events was conducted with respect to radiological 
consequences. This re-analysis was performed in accordance with AST 
methodology provided in Regulatory Guide (RG) 1.183 and with ARCON96 
atmospheric dispersion methodology provided in RG 1.194. The 
reanalysis consequences were expressed in terms of Total Effective 
Dose Equivalent (TEDE) dose.
    Implementation of the AST methodology, as described in 10 CFR 
50.67, specifies control room, exclusion area boundary (EAB), and 
low population zone (LPZ) dose acceptance criteria in terms of TEDE 
dose. The dose acceptance criteria for specific events are specified 
in RG 1.183. The revised analyses for all evaluated events meet the 
applicable RG 1.183 TEDE dose acceptance criteria for AST 
implementation.
    The previous dose calculations analyzed the dose consequences to 
thyroid and whole body as a result of postulated design basis 
events. The previous control room dose calculations were shown to be 
within the regulatory limits of 10 CFR 50 Appendix A General Design 
Criterion 19 with respect to thyroid, beta-skin and whole body dose. 
The previous LOCA and SLB offsite dose calculations were shown to be 
within the regulatory limits of 10 CFR 100.11 with respect to 
thyroid and whole body dose. The previous FHA-IC and FHA-FHB offsite 
dose calculations were shown to be well within (i.e., less than 25 
percent of) the regulatory limits of 10 CFR 100.11 with respect to 
thyroid and whole body dose. RG 1.183 Footnote 7 provides a means to 
compare the thyroid and whole body dose results of the previous 
calculations with the TEDE results of the AST calculations. This 
methodology requires multiplying the previous thyroid dose by 0.03 
and adding the product to the previous whole body dose. The 
resultant

[[Page 5249]]

``effective'' TEDE is then compared to the AST TEDE result. This 
comparison is presented in Table 5-1.
    The Table 5-1 comparison shows a decrease in dose consequences 
when evaluated using AST methodology for all but the LOCA offsite 
dose receptors. The LOCA EAB dose using AST methodology has 
increased due to the requirement to calculate the maximum 2-hour 
window EAB dose versus the previous requirement to calculate the 0 
to 2 hour window EAB dose. The LOCA LPZ dose using AST methodology 
has increased primarily due to changes in the AST Refueling Water 
Storage Tank (RWST) iodine transport model. Although the LOCA EAB 
and LPZ doses using AST methodology have increased, they remain 
significantly below the 25 Rem TEDE offsite dose acceptance 
criterion.

            Table 5-1.--Comparison of Previous and AST Doses
------------------------------------------------------------------------
                                      ``Effective''
                                     TEDE of previous
        Event-dose receptor           dose analyses      AST TEDE (Rem)
                                          (Rem)
------------------------------------------------------------------------
FHA-IC:
    Control Room..................                1.0           2.7 E-01
    EAB...........................                2.0           8.0 E-01
    LPZ...........................           5.6 E-02           2.3 E-02
FHA-FHB:
    Control Room..................           3.7 E-01           7.3 E-02
    EAB...........................           6.6 E-01           2.1 E-01
    LPZ...........................           1.9 E-02           6.1 E-03
LOCA:
    Control Room..................                4.5                2.7
    EAB...........................                3.7                5.1
    LPZ...........................                1.2                1.8
SLB-OC:
    Control Room..................              (\1\)                2.1
    EAB...........................                8.0                4.1
    LPZ...........................              (\1\)               0.1
------------------------------------------------------------------------
\1\ Not evaluated.

    The proposed changes do not increase the probability of an 
accident previously evaluated. The proposed changes result in dose 
consequences that, if compared to previous ones, are in most cases 
decreased and in other cases only slightly increased (using guidance 
in footnote 7 of RG 1.183). However, the dose consequences of the 
revised analyses are below the AST regulatory acceptance criteria.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The implementation of this proposed change does not create the 
possibility of an accident of a different type than was previously 
evaluated in the UFSAR. The proposed change credits the AST for the 
design basis radiological site boundary and control room dose 
analyses and expands the allowed use of fuel failure estimates by 
DNB statistical convolution methodology from only the reactor 
coolant pump sheared shaft event to the UFSAR Chapter 15 non-LOCA 
events that assume a loss of flow (i.e., a loss of AC power) and 
that fail fuel. The changes proposed do not change how Design Basis 
Accident (DBA) events were postulated nor do the changes themselves 
initiate a new kind of accident with a unique set of conditions. The 
changes proposed are based on a re-analysis of offsite and control 
room doses for four design basis accidents. The revised analyses are 
consistent with the regulatory guidance established in RG 1.183. The 
revised analyses utilize the most current understanding of source 
term timing and chemical forms. Through this re-analysis, no new 
accident initiator or failure mode was identified.
    Therefore, this proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The implementation of this proposed amendment does not reduce 
the margin of safety. The alternative source term radiological dose 
consequence analyses utilize the regulatory acceptance criteria of 
10 CFR 50 Appendix A General Design Criterion (GDC) 19 and 10 CFR 
50.67, as specified in RG 1.183. These acceptance criteria have been 
developed for the purpose of use in design basis accident analyses 
such that meeting these limits demonstrates adequate protection of 
public health and safety. An acceptable margin of safety is inherent 
in these licensing limits. The radiological analyses results remain 
within these regulatory acceptance criteria.
    Therefore, there is no significant reduction in the margin of 
safety as a result of the proposed amendment.
    Based on the above, SCE concludes that the proposed amendments 
present no significant hazards consideration under the standards set 
forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.

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

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 amendment request: November 12, 2004.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications 3.1.7, ``Standby Liquid Control (SLC) 
System,'' for Hatch Units 1 and 2. The proposed amendments would update 
Figure 3.1.7-1 of Units 1 and 2 TS to reflect the increased 
concentration of Boron-10 in the solution. Conforming revisions to 
Bases B 3.1.7, ``Standby Liquid Control (SLC) System'' are also 
included.
    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

[[Page 5250]]

consequences of an accident previously evaluated?
    This is a proposed change to Figure 3.1.7-1 of the Units 1 and 2 
Technical Specifications. This figure is a graph of the weight 
percent of Sodium Pentaborate solution in the Standby Liquid Control 
(SLC) Tank, as a function of the gross volume of solution in the 
tank. The figure is proposed to be changed in order to accommodate 
an injection of Sodium Pentaborate solution into the reactor, 
following an ATWS event, such that the concentration of Boron-10 
atoms in the reactor will be 800 ppm natural Boron equivalent. This 
is necessary to accommodate increased cycle energy requirements for 
the Hatch Units 1 and 2 cores.
    The proposed change to the Figure will not increase the 
probability of an ATWS event because the curve has nothing to do 
with the prevention of an ATWS event. The new requirements will 
ensure that, in the future, the core will have adequate shutdown 
margin to mitigate the consequences of an ATWS event.
    Also, no systems or components designed to ensure the safe 
shutdown of the reactor are being physically changed as a result of 
this proposed TS change. In fact, no safety related systems or 
components designed for the prevention of previously evaluated 
events are being altered by the amendment.
    As a result, the probability and consequences of an ATWS event, 
or any other previously evaluated event, will not increase as a 
result of this amendment.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    This proposed TS revision results in a change to the SLC TS 
figure 3.7.1-1 requirements. However, this does not result in 
physical changes to the SLC system. SLC pump operation, maintenance 
and testing remain the same. Accordingly, no changes to the 
operation, maintenance or surveillance procedures will result from 
this TS revision request. Therefore, no new modes of operation are 
introduced by this TS change.
    Since no new modes of operation are introduced, the proposed 
change does not create the possibility of a new or different type 
event from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    This proposed TS change is being made to increase the boron 
concentration requirements of the sodium pentaborate solution 
injected into the reactor vessel following an Anticipated Transient 
Without Scram (ATWS) event. The change is necessary due to new fuel 
designs and higher energy requirements for fuel cycles. Therefore, 
the change is being made to insure that shutdown requirements can be 
met for the ATWS event. This will insure the margin of safety with 
respect to ATWS will continue to be met.
    Consequently, this proposed TS change will not result in a 
decrease in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: October 14, 2004.
    Description of amendment request: The requested change will delete 
Technical Specification (TS) 6.9.1.5 related to the annual 
``Occupational Radiation Exposure Report,'' and TS 6.9.1.10, ``Monthly 
Reactor Operating Report.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated October 14, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates the Technical Specifications 
(TSs) reporting requirements to provide a monthly operating letter 
report of shutdown experience and operating statistics if the 
equivalent data is submitted using an industry electronic database. 
It also eliminates the TS reporting requirement for an annual 
occupational radiation exposure report, which provides information 
beyond that specified in NRC regulations. The proposed change 
involves no changes to plant systems or accident analyses. As such, 
the change is administrative in nature and does not affect 
initiators of analyzed events or assumed mitigation of accidents or 
transients. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant, add any new equipment, or require any existing equipment 
to be operated in a manner different from the present design. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in the margin of safety.
    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Michael L. Marshall, Jr.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety

[[Page 5251]]

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

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina; Carolina Power & Light Company, Docket No. 50-261, H.B. 
Robinson Steam Electric Plant, Unit No. 2, Darlington County, South 
Carolina

    Date of application for amendments: December 19, 2003, as 
supplemented January 14, 2004.
    Brief description of amendments: The amendments allows entry into a 
mode or other specified condition in the applicability of a technical 
specification (TS), while in a condition statement and the associated 
required actions of the TS, provided the licensee performs a risk 
assessment and manages risk consistent with the program as proposed by 
the industry's Technical Specification Task Force (TSTF) and is 
designated TSTF-359.
    Date of issuance: January 11, 2005.
    Effective date: January 11, 2005.
    Amendment Nos.: 233 and 260.
    Facility Operating License Nos. DPR-71, DPR-62, and DPR-23.: 
Amendments change the Technical Specifications.
    Date of initial notice in Federal Register: The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated January 11, 2005.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3, Westchester County, New York

    Date of application for amendment: October 26, 2004, as 
supplemented on December 22, 2004.
    Brief description of amendment: The amendment revises Technical 
Specification 3.7.11, ``Control Room Ventilation System (CRVS),'' to 
allow, on a one-time basis, an extension of the allowed outage time to 
support placement of the CRVS in an alternate configuration for tracer 
gas testing. The proposed amendment would also allow self-contained 
breathing apparatus and potassium iodide pills to be used as 
compensatory measures for the control room operators in the event that 
the tracer gas test results are not bounded by the dose consequence 
evaluations.
    Date of issuance: January 19, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 223.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 8, 2004 (69 FR 
64792).
    The December 22 letter provided information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 19, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 25, 2003.
    Brief description of amendments: The amendments modify the Limerick 
Generating Station, (LGS) Units 1 and 2, Technical Specifications (TSs) 
contained in Appendix A to Operating License Nos. NPF-39 and NPF-85, 
respectively. The amendments add a footnote to the LGS TS 3.4.3.2.e to 
indicate that reactor coolant system (RCS) pressure isolation valve 
leakage is excluded from any other allowable RCS operational leakage 
specified in LGS TS 3.4.3.2.
    Date of issuance: January 18, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 172 and 134.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the TSs.
    Date of initial notice in Federal Register: February 3, 2004 (69 FR 
5203).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 18, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 31, 2004.
    Brief description of amendment: This amendment revised Technical 
Specification (TS) requirements for mode change limitations in Limiting 
Condition for Operation 3.0.4 and Surveillance Requirement 3.0.4 to 
adopt the provisions of Industry TS Task Force (TSTF) change TSTF-359, 
``Increase Flexibility in Mode Restraints.''
    Date of issuance: January 6, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 131.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 6, 2004 (69 FR 
40675).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 6, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 19, 2003.
    Brief description of amendment: The amendment modifies TS 
requirements to adopt the provisions of Industry/TS Task Force (TSTF) 
change TSTF-359, ``Increased Flexibility in Mode Restraints.'' The 
availability of TSTF-359 for adoption by licensees was announced in the 
Federal Register on April 4, 2003 (68 FR 16579).
    Date of issuance: January 11, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days of issuance.
    Amendment No.: 215.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 17, 2004 (69 
FR 7523).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 11, 2005.
    No significant hazards consideration comments received: No.

[[Page 5252]]

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 application for amendments: April 23, 2004.
    Brief description of amendments: The amendments revise several 
Technical Specification (TS) Allowed Outage Times for TS 3.3.3, 
Accident Monitoring Instrumentation, to be consistent with the 
Completion Times in the related Specification in NUREG-1431, Revision 
2, ``Standard Technical Specifications Westinghouse Plants (the 
Improved Standard Technical Specifications, or ISTS).''
    Date of issuance: January 6, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos: 227 and 223.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 25, 2004 (69 FR 
29767).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 6, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 23, 2003.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) requirements to adopt the provisions of the TS Task 
Force (TSTF) change TSTF-359, regarding increased flexibility in mode 
changes. The availability of TSTF-359 for adoption by licensees was 
announced in the Federal Register on April 4, 2003 (68 FR 16579).
    Date of issuance: January 10, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 255.
    Facility Operating License No. DPR-49: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 16, 2004 (69 
FR 55844).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 10, 2005.
    No significant hazards consideration comments received: No.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: September 30, 2004.
    Brief description of amendments: The amendments delete Technical 
Specification (TS) 6.9.1.2, ``Occupational Radiation Exposure Report,'' 
and TS 6.9.1.5, ``Monthly Operating Reports,'' as described in the 
Notice of Availability published in the Federal Register on June 23, 
2004 (69 FR 35067).
    Date of issuance: January 5, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: Unit 1-168; Unit 2-157.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: October 26, 2004 (69 FR 
62478).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 5, 2005.
    No significant hazards consideration comments received: No.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: February 3, 2004 as supplemented by 
letter dated December 1, 2004.
    Brief description of amendments: The amendments modify Technical 
Specifications (TSs) requirements to adopt the provisions of Industry/
TS Task Force (TSTF) change TSTF-359, ``Increase Flexibility in Mode 
Restraints.'' The availability of TSTF-359 for adoption by licensees 
was announced in the Federal Register on April 4, 2003 (68 FR 16579).
    Date of issuance: January 10, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: Unit 1-170; Unit 2-158.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: March 2, 2004 (69 FR 
9865).
    The supplement dated December 1, 2004, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 10, 2005.
    No significant hazards consideration comments received: No.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an

[[Page 5253]]

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

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

    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the

[[Page 5254]]

authority to act for the petitioners/requestors with respect to that 
contention.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) e-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer or the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).

STP Nuclear Operating Company, Docket No. 50-498, South Texas Project, 
Unit 1, Matagorda County, Texas

    Date of amendment request: January 6, 2005.
    Description of amendment request: The amendment revises Technical 
Specification (TS) 3.7.4, ``Essential Cooling Water System,'' and the 
associated TS for systems supported by the Essential Cooling Water 
(ECW), to extend the allowed outage time for an additional 7 days for 
ECW Train B as a one-time change for the purpose of making repairs to 
the Train B ECW pump.
    Date of issuance: January 10, 2005.
    Effective date: Effective as of the date of issuance and shall be 
implemented immediately.
    Amendment No.: 169.
    Facility Operating License No. NPF-76: Amendment revises the 
technical specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, state consultation, and final NSHC 
determination are contained in a safety evaluation dated January 10, 
2005.
    Attorney for licensee: A.H. Gutterman, Morgan, Lewis & Bockius, 
1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Michael K. Webb, Acting.

    Dated at Rockville, Maryland, this 24th day of January 2005.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management Office of Nuclear 
Reactor Regulation.
[FR Doc. 05-1574 Filed 1-31-05; 8:45 am]
BILLING CODE 7590-01-P