[Federal Register Volume 69, Number 244 (Tuesday, December 21, 2004)]
[Notices]
[Pages 76486-76498]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-27614]



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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 November 25, 2004, through December 9, 2004. 
The last biweekly notice was published on December 7, 2004 (69 FR 
70712).

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

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

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applicant on a material issue of law or fact. Contentions shall be 
limited to matters within the scope of the amendment under 
consideration. The contention must be one which, if proven, would 
entitle the petitioner/requestor to relief. A petitioner/requestor who 
fails to satisfy these requirements with respect to at least one 
contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: June 10, 2004.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 3.6.3, ``Containment Isolation Valves,'' 
to allow the surveillance frequencies for leakage rate testing to be 
specified in the Catawba Nuclear Station Containment Leak Rate Testing 
Program.
    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:

First Standard

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No.
    This amendment will not change any previously evaluated 
accidents such as the postulated ``Fuel Handling Accident (FHA) in 
Containment''. No credit is assumed for VP containment isolation in 
the FHA within containment. The Containment Purge (VP) System and 
Hydrogen Purge (VY) System containment isolation valves are sealed 
closed during modes 1 through 4. The Containment Air Release and 
Addition (VQ) System containment isolation valves are designed to 
close within 5 seconds of a containment phase ``A'' isolation 
signal. The prevention and mitigation of these accidents is not 
affected by this change.
    Test data demonstrates that the likelihood of a malfunction of a 
resilient seal in one of the VP, VY, or VQ valves is not increased 
by this change in the surveillances. The systems will continue to be 
able to perform their design functions of isolating containment 
during the evaluated accidents. Test procedures will continue to 
monitor the leakage of these valves to ensure the design function 
will continue to be met. There is no impact on previously evaluated 
accidents since the valves will continue to close and seal or remain 
closed as originally assumed in the accident scenarios.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

Second Standard

    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No.
    This change does not involve a physical alteration to the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing any normal plant operation. The 
change does not alter assumptions made in the safety analyses or 
licensing basis. This change will not affect or degrade the ability 
of the Containment Purge System, Hydrogen Purge System, or 
Containment Air Release and Addition System valves to perform their 
specified safety functions. Therefore, the change does not create 
the possibility of a new or different kind of credible accident from 
any accident previously evaluated.

Third Standard

    Does the proposed change involve a significant reduction in a 
margin of safety?
    No.
    SR 3.6.3.6 currently states: ``The measured leakage rate for 
Containment Purge System and Hydrogen Purge System valves must be < 
0.05 La (Design Leakage Rate) when pressurized to Pa 
(Design Containment Pressure). The measured leakage rate for 
Containment Air Release and Addition valves must be < 0.01 
La when pressurized to Pa. These required 
maximum leak rates will not be changed by this amendment. Testing of 
these valves to measure leakage through the valve seats will 
continue, only at a different frequency based on past test results. 
This will be a nominal frequency of 18 months for the VP System and 
in accordance with 10 CFR 50, Appendix J, Option B for the VQ and VY 
Systems. Therefore, the proposed changes listed above do not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

[[Page 76488]]

    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: July 19, 2004.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 3.8.4, ``DC Sources--Operating'' and TS 
3.8.6, ``Battery Cell Parameters'' to allow for the replacement of the 
existing nickel cadmium diesel generator batteries with conventional 
lead acid batteries.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed license amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The DG batteries are not accident initiating equipment; they are 
accident mitigating equipment. As such, they cannot affect the 
probability of any accident being initiated. The performance of the 
replacement batteries will exceed that of the existing batteries. 
Therefore, no accident consequences will be adversely impacted.
    (2) The proposed license amendments do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The DG batteries are not capable by themselves of initiating any 
accident. Other than the replacement of the batteries themselves and 
the associated modification work (e.g., installation of the battery 
HVAC system), no physical changes to the overall plant are being 
proposed. No changes to the overall manner in which the plant is 
operated are being proposed. Therefore, no potential for new 
accident types is generated.
    (3) The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their intended functions. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment. The modification to replace the DG 
batteries will not have any impact on these barriers. In addition, 
no accident mitigating equipment will be adversely impacted as a 
result of the battery replacement. The replacement batteries will 
have overall performance capabilities equal to or greater than those 
for the existing batteries. Therefore, existing safety margins will 
be preserved.

    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: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: September 21, 2004.
    Description of amendments request: The proposed amendment would 
delete the requirements from the technical specifications (TS) 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 TS for nuclear power reactors currently licensed to operate. The 
revised 10 CFR 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 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 no significant hazards consideration determination 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 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

[[Page 76489]]

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 C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: April 14, 2004.
    Description of amendment request: The proposed amendment would add 
a new section to the Technical Specifications (TSs) and two new 
Limiting Conditions for Operations (LCOs) to allow certain reactor 
coolant system (RCS) hydrostatic and system leakage pressure tests to 
be performed with the reactor pressure vessel temperature above 
212[deg] Fahrenheit (F). The first LCO would allow specified TS 
requirements to be changed to permit performance of special tests and 
operations, which otherwise could not be performed if required to 
comply with the requirements of the TSs. The second LCO would require 
reactor low water level instrumentation, standby gas treatment system, 
and secondary containment to be OPERABLE to allow certain RCS pressure 
tests to be performed with the reactor pressure vessel temperature 
above 212[deg] F, and provides for an exemption from the requirements 
for OPERABILITY for other systems that currently go into effect when in 
Hot Shutdown or when RCS temperature is greater than 212[deg] F. It 
will also update the Table of Contents to reflect the proposed changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The Nuclear Regulatory Commission (NRC) staff has 
reviewed the licensee's analysis against the standards of 10 CFR 
50.92(c). The NRC staff's review is presented below.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated. The 
probability of an accident previously evaluated is not significantly 
increased because the proposed change will not alter the method by 
which RCS hydrostatic pressure and leak testing is performed. Under 
this proposed change the secondary containment, standby gas treatment 
system and associated initiation instrumentation are required to be 
operable during the performance of RCS hydrostatic pressure and leak 
testing and would be capable of handling any airborne radioactivity or 
steam leaks that could occur. The required pressure testing conditions 
provide adequate assurance that the consequences of a steam leak will 
be conservatively bounded by the consequences of a main steamline break 
(MSLB) outside the primary containment. Accordingly, the consequences 
of previously evaluated accidents are not increased significantly.
    The proposed update to the Table of Contents is editorial in 
nature. Since this update is administrative in nature, it cannot 
increase the probability or consequences of a previously analyzed 
accident.
    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 previously evaluated?
    The proposed amendment change will not alter the way that 
hydrostatic pressure and leak testing is performed. Therefore, the 
proposed change will 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?
    The proposed amendment will not involve a significant reduction in 
a margin of safety for a postulated MSLB outside of primary 
containment. The proposed changes and additions result in increased 
system operability requirements above those that currently exist during 
the performance of RCS hydrostatic pressure and leak testing. The 
incremental increase in stored energy in the vessel during testing will 
be conservatively bounded by the consequences of the postulated MSLB 
outside of primary containment. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.
    The proposed update to the Table of Contents is editorial in 
nature. Since this update is administrative in nature, the proposed 
change does not involve a significant reduction in a margin of safety
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c)) are satisfied. Therefore, the NRC staff proposes to determine 
that the

[[Page 76490]]

amendment request involves no significant hazards consideration.
    Attorney for licensee: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts, 02360-5599.
    NRC Section Chief: Darrell Roberts.

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

    Date of amendment request: September 2, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 4.5.B.2.2 (TS) to change the 
surveillance requirement frequency for air testing the drywell and 
suppression pool (torus) spray headers and nozzles from ``once every 5 
years'' to ``following maintenance that could result in nozzle 
blockage.''
    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 drywell and torus headers and spray nozzles are not assumed 
to be initiators of any accidents previously evaluated. Maintenance 
practices and normal environmental conditions to which the system is 
subjected are adequate to ensure operability of the systems. Since 
the system will be able to perform its accident mitigation function, 
the consequences of accident previously evaluated are not increased. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident [from any accident] previously 
[evaluated]?
    Response: No.
    The revised surveillance does not introduce any new mode of 
plant operation, does not involve physical modification of the 
plant, or any new operating modes, and cannot introduce new accident 
initiators. 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.
    Maintenance practices and normal environmental conditions to 
which the system is subjected are adequate to ensure operability of 
the systems. As the spray nozzles are expected to remain fully 
capable of performing their post-accident mitigation function, 
margin of safety is not reduced. 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts, 02360-5599.
    NRC Section Chief: Darrell J. Roberts.

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

    Date of amendment request: September 2, 2004.
    Description of amendment request: The proposed amendment would 
remove a license condition that currently requires the reactor not to 
be operated for more than 24 hours if one recirculation loop is out of 
service. It would revise Technical Specifications (TSs) to allow the 
minimum critical power ratio (MCPR) safety limit to be changed for 
single loop operations (SLOs). It would also revise the current jet 
pump limiting condition for operation and surveillance requirements to 
allow for the conduct of a TS required surveillance during SLOs. The 
proposed amendment would modify the TSs to address SLO operating 
conditions and restrictions, and delete a TS condition related to 
thermal-hydraulic stability. It would update the TSs for average planar 
linear heat generation rate for SLOs, and update the thermal power 
applicability restrictions to be consistent with NUREG-1433, Revision 
3, ``Standard Technical Specifications for General Electric Boiling-
Water Reactors.'' It would also revise the TSs for linear heat 
generation rate and MCPR for thermal power applicability restrictions. 
The proposed amendment makes an administrative change to have MCPR 
recalculated when reactor power is equal to or greater than 25 percent. 
Lastly, it would update the TSs' table of contents and TS pages to 
administratively reflect all of these proposed changes.
    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 license and technical specification changes will 
allow the plant to be operated with one recirculation pump for 
longer than 24 hours provided that appropriate limits are 
instituted. Extended single recirculation loop operation has been 
evaluated and methodologies have been established for determining 
appropriate operating limits. Implementation of the single 
recirculation loop operating limits ensures that system operation is 
in conformance with the conditions established to minimize the 
probability of accidents and the associated consequences. Required 
completion times for implementing the system operating limits and 
restoring out of specification limits minimize the probability that 
an accident occurs when out of specification conditions exist while 
allowing for deliberate operator action.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability of occurrence 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 license and technical specification changes will 
allow plant operation with a single recirculation loop for longer 
than 24 hours. The proposed changes introduce an additional 
recirculation system-operating mode, however, existing system 
component operating equipment or operating characteristics will not 
change. The Pilgrim Station Single Loop Analysis Report identifies 
required operating limits that apply when the system will be 
operated in the single loop operation mode. Implementation of these 
operating limits will ensure that the system is operated in 
accordance with design. Additionally, revised jet pump surveillance 
ensures that loop specific surveillance is performed as required to 
validate the bounding assumptions of existing accident analyses. As 
such, no new failure mechanisms are created and existing design 
evaluations bound system operation.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in 
[a] margin of safety?
    Response: No.
    The proposed license and technical specification changes 
identify the operating limits that apply to single recirculation 
loop operation. These proposed recirculation system limits were 
identified to ensure that system operation would be in conformance 
to the conditions evaluated in applicable accident and transient 
analyses. Implementation of the proposed limits for single 
recirculation loop operation ensures that safety margins are 
maintained. Required completion times for implementing the system 
operating limits minimizes the

[[Page 76491]]

possibility that an accident occurs when out of specification 
conditions exist.
    Therefore, the proposed changes do not involve a significant 
reduction in [a] margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts 02360-5599.
    NRC Section Chief: Darrell Roberts.

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

    Date of amendment request: September 16, 2003 as supplemented by 
letter dated March 15, 2004.
    Description of amendment request: The proposed amendment would 
relocate the current definition of surveillance frequency to new 
Technical Specification (TS) Sections 4.0.2 and 4.0.3, and revise the 
requirements for missed surveillance in Section 4.0.3. This change is 
consistent with NRC-approved Industry/Technical Specification Task 
Force (TSTF) change TSTF-358, Revision 5. The proposed change would 
allow a longer period of time to perform a missed surveillance. The 
time is extended from the current limit of up to 24 hours or up to the 
limit of the specified frequency, whichever is less; to up to 24 hours 
or up to the limit of the specified frequency, whichever is greater. In 
conjunction with the proposed change, the proposed amendment would add 
the requirements for a Bases Control Program which is consistent with 
Section 5.5 of NUREG 1433. In addition, the current definition of 
surveillance interval (definition ``Z'') would be re-worded and 
relocated to new Section 4.0.1 consistent with Surveillance Requirement 
3.0.1 of NUREG 1433. Appropriate Bases, also consistent with NUREG 1433 
would be adopted for the new sections. An editorial change would be 
made to TS 6.7.C which references the current definition of 
surveillance frequency to now reference the new Section 4.0.2.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments 
concerning missed surveillances, 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 September 28, 2001 (66 FR 49714). The licensee affirmed the 
applicability of the model NSHC determination in its application dated 
September 16, 2003. The model NSHC determination analysis for changes 
to the TS associated with missed surveillances, and the NSHC 
determination analysis provided by the licensee for the remaining TS 
changes, is provided herein.
    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

    With regard to the proposed change to the TS associated with 
missed surveillances, the proposed change relaxes the time allowed 
to perform a missed surveillance. The time between surveillances is 
not an initiator of any accident previously evaluated. Consequently, 
the probability of an accident previously evaluated is not 
significantly increased. The equipment being tested is still 
required to be operable and capable of performing the accident 
mitigation functions assumed in the accident analysis. As a result, 
the consequences of any accident previously evaluated are not 
significantly affected. Any reduction in confidence that a standby 
system might fail to perform its safety function due to a missed 
surveillance is small and would not, in the absence of other 
unrelated failures, lead to an increase in consequences beyond those 
estimated by existing analyses. The addition of a requirement to 
assess and manage the risk introduced by the missed surveillance 
will further minimize possible concerns. Therefore, this change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    With regard to the remaining proposed changes to the TSs, the 
proposed changes do not involve physical changes to the plant or 
introduce any new modes of operation. Accordingly, continued 
assurance is provided that the process variables, structures, 
systems, and components are maintained such that there will be no 
degradation of any fission product barrier which could increase the 
radiological consequences of an accident. Therefore, the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

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

    With regard to the proposed changes to the TSs associated with 
missed surveillances, 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. A missed surveillance will not, in and of 
itself, introduce new failure modes or effects and any increased 
chance that a standby system might fail to perform its safety 
function due to a missed surveillance would not, in the absence of 
other unrelated failures, lead to an accident beyond those 
previously evaluated. The addition of a requirement to assess and 
manage the risk introduced by the missed surveillance will further 
minimize possible concerns. Thus, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    With regard to the remaining proposed changes to the TSs, the 
proposed changes do 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, the 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

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

    With regard to the proposed changes to the TSs associated with 
missed surveillances, the extended time allowed to perform a missed 
surveillance does not result in a significant reduction in the 
margin of safety. As supported by the historical data, the likely 
outcome of any surveillance is verification that the LCO [limiting 
condition for operation] is met. Failure to perform a surveillance 
within the prescribed frequency does not cause equipment to become 
inoperable. The only effect of the additional time allowed to 
perform a missed surveillance on the margin of safety is the 
extension of the time until inoperable equipment is discovered to be 
inoperable by the missed surveillance. However, given the rare 
occurrence of inoperable equipment, and the rare occurrence of a 
missed surveillance, a missed surveillance on inoperable equipment 
would be very unlikely. This must be balanced against the real risk 
of manipulating the plant equipment or condition to perform the 
missed surveillance. In addition, parallel trains and alternate 
equipment are typically available to perform the safety function of 
the equipment not tested. Thus, there is confidence that the 
equipment can perform its assumed safety function. Therefore, these 
changes do not involve a significant reduction in a margin of 
safety.
    With regard to the remaining proposed changes to the TSs, the 
administrative changes do not alter the basic operation of process 
variables, systems, or components as described in the safety 
analysis. No new equipment is introduced. Accordingly, the

[[Page 76492]]

proposed changes do not involve a significant reduction in a margin 
of safety.

    Therefore, this change does not involve a significant reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and its 
endorsement of the model NSHC for missed surveillances 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: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: October 5, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Section 6.7.C ``Primary Containment Leak 
Rate Testing Program,'' to allow a one-time extension to the 10-year 
interval for performing the next Type A containment integrated leak 
rate test (ILRT). Specifically, the change would allow the test to be 
performed within 15 years from the last ILRT, which was performed in 
April 1995.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed revision to Technical Specifications adds a one-
time extension to the current interval for Type A testing. The 
current test interval of 10.6 years, based on past performance, is 
extended on a one-time basis to fifteen years from the last Type A 
test. The proposed extension to Type A testing cannot increase the 
probability of an accident previously evaluated since the 
containment Type A testing extension is not a modification and the 
test extension is not of a type that could lead to equipment failure 
or accident initiation.
    The proposed extension to Type A testing does not involve a 
significant increase in the consequences of an accident since 
research documented in NUREG-1493 has found that, generically, very 
few potential containment leakage paths are not identified by Type B 
and C tests. The NUREG concluded that reducing the Type A (ILRT) 
testing frequency to once per twenty years was found to lead to an 
imperceptible increase in risk. These generic conclusions were 
confirmed by a plant specific risk analysis performed using the 
current Vermont Yankee Probabilistic Safety Assessment (PSA) 
internal events model that concluded the consequences are low to 
negligible.
    Testing and inspection programs in place also provide a high 
degree of assurance that the containment will not degrade in a 
manner detectable only by Type A testing. The last two successful 
Type A tests indicate a very leak tight containment. Type B and C 
testing required by Technical Specifications will identify any 
containment opening such as valves that would otherwise be detected 
by the Type A tests. Inspections, including those required by the 
ASME [American Society of Mechanical Engineers] code and the 
Maintenance Rule are performed in order to identify indications of 
containment degradation that could affect that leak tightness.
    Therefore, the proposed changes do not represent a significant 
increase in the probability or consequences of an accident 
previously analyzed.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed revision to Technical Specifications adds a one 
time extension to the current interval for Type A testing. The 
current test interval of 10.6 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 cannot 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 mode creating an accident or affecting the mitigation of an 
accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    The proposed revision to Technical Specifications adds a one 
time extension to the current interval for Type A testing. The 
current test interval of 10.6 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 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 extension in Type A leakage testing resulted in an 
imperceptible increase in risk to the public. NUREG-1493 found that, 
generically, the design containment leakage rate contributes about 
0.1 percent to the individual risk and that the decrease in Type A 
testing frequency would have a minimal affect on this risk since 95% 
of the potential leakage paths are detected by Type C testing. This 
was further confirmed by a plant specific risk assessment using the 
current Vermont Yankee PSA internal events model that concluded the 
risk associated with this change is negligibly small and/or non-risk 
significant.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128. 
[FEDREG][VOL]*[/VOL][NO]*[/NO][DATE]*[/DATE][NOTICES] 
[NOTICE][PREAMB][AGENCY]*[/AGENCY][SUBJECT]*[/SUBJECT] ?>
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: October 6, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Surveillance Requirement 4.5.B.1 related 
to air testing of the drywell spray headers and nozzles. Specifically, 
the amendment would change the test frequency from once every 5 years 
to following maintenance that could result in nozzle blockage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has

[[Page 76493]]

reviewed the licensee's analysis against the three standards of 10 CFR 
50.92(c). The NRC staff's analysis is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would revise the Technical Specification 
surveillance requirements associated with the air test of the drywell 
spray headers and nozzles. The frequency of the air test would be 
changed from a fixed 5-year frequency to following maintenance that 
could result in nozzle blockage.
    This surveillance test is performed while the plant is in a cold 
shutdown condition and the equipment is not required to be operable. 
The testing is to verify that the spray headers and nozzles are not 
obstructed. The proposed change in the surveillance test frequency will 
not result in any design changes to systems, structures, or components, 
or their method of operation. The drywell spray headers and nozzles are 
not initiators of any accidents previously evaluated. Therefore, the 
proposed change does not involve a significant increase in the 
probability of any accident previously evaluated.
    The drywell spray headers provide a means to control both 
temperature and pressure inside the primary containment, within design 
limits, under post-accident conditions. Due to the system design and 
operation considerations discussed in the licensee's application, the 
potential for corrosion product formation is minimized. In addition, 
the Vermont Yankee foreign material exclusion program has been judged 
to be sufficient to ensure that foreign material is not inadvertently 
introduced into the system. The proposed testing requirements are 
considered sufficient to provide a high degree of confidence that 
containment spray will function when required. Therefore, the proposed 
change does not involve a significant increase in the 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 proposed change in the surveillance test frequency does not 
create the possibility of a new or different kind 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 mode, creating an accident or affecting the mitigation of an 
accident.
    Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the surveillance requirement to verify 
that the drywell spray headers and nozzles are unobstructed. Industry 
experience, Vermont Yankee surveillance history and the environmental 
conditions the system is subjected to are adequate to ensure continued 
system availability. As the spray nozzles are expected to remain 
unobstructed and be able to perform their post-accident function, plant 
safety is not affected.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92 are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: Allen G. Howe.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power 
Plant, Unit 3, Humboldt County, California

    Date of amendment request: July 9, 2004.
    Description of amendment request: The Humboldt Bay Power Plant 
(HBPP), Unit 3, is a decommissioning nuclear power plant that was 
permanently shutdown in July 1976. In December of 2003, Pacific Gas and 
Electric (PG&E or the licensee) applied for a license to store its 
spent fuel in an onsite dry cask independent spent fuel storage 
installation (ISFSI). Moving the spent fuel to an ISFSI would permit 
the licensee to begin significant decommissioning activities. The 
licensee has chosen to use a Holtec HI-STAR HB spent fuel cask handling 
system involving a spent fuel multipurpose canister and overpack. To 
facilitate spent fuel transfer from the HBPP spent fuel pool to the 
ISFSI, the licensee will also need to install a new crane that can be 
used to lift the cask handling system loaded with spent fuel 
assemblies. The licensee states it will be able to satisfy the 
applicable guidance of NUREG-0612, ``Control of Heavy Loads at Nuclear 
Power Plants,'' and NUREG-0554. ``Single-Failure Proof Cranes for 
Nuclear Power Plants,'' in performing the necessary movement of the 
HBPP spent fuel to dry cask storage. The licensee has requested a 
license amendment that approves the use of the crane and associated 
changes to the HBPP Defueled Safety Analysis Report (DSAR) along with 
analyses, design, and procedural changes required to implement transfer 
of the spent fuel from the spent fuel pool to the ISFSI.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. With the HI-STAR HB System and the associated design and 
handling procedures, all cask drops and other events, which could 
damage other spent fuel, have been precluded through the robust 
handling systems, and mechanical arrangement that preclude crane 
movement over spent fuel, meeting the guidelines of NUREG-0612. 
Revisions of the HBPP procedures implementing the control of heavy 
loads ensures that PG&E will meet the NUREG-0612 guidelines and will 
protect the fuel storage locations and the new HI-STAR HB System 
loading/unloading activities. As a result of this design approach, a 
cask-handling accident that results in a significant offsite 
radiological release is not considered credible as demonstrated by 
the probabilistic evaluation that was performed using the guidelines 
of NUREG-0612 Appendix B and updated information from NUREG-1774 
[``A Survey of Crane Operating Experience at U.S. Nuclear Power 
Plants from 1968 through 2002.'']
    Other HBPP licensing-basis events, such as the drop of a spent 
fuel assembly, have not been affected by these changes and remain 
bounding events for potential radiological consequences.
    The proposed design of the dry cask system, the handling system, 
and associated procedural controls provide assurance that: (1) 
operational errors and mishandling events, and (2) support system 
malfunctions will not result in an increase in the probability or 
consequence of an accident previously analyzed.
    The proposed changes to use the Holtec HI-STAR HB system have 
been evaluated for seismic events and tornado missile impacts and it 
has been determined that these changes will not result in an 
increase in the probability or consequences of an accident 
previously evaluated. The Fire Protection Program will ensure that 
the combustible materials are properly controlled such that the 
total combustibles meet the current program commitments. Therefore, 
the proposed changes do not involve a

[[Page 76494]]

significant increase in the probability or consequences of an 
accident.
    2. Does the proposed amendment create the possibility of a new 
or different type of accident from any accident previously 
evaluated?
    No. The engineering design measures and the handling procedures 
preclude the possibility of new or different kinds of accidents. 
Damage to 10 CFR 50 structures, systems, and components from the 
cask handling and associated activities, and events resulting from 
possible damage to contained fuel have been considered. Both the 
types of accidents and the results remain within the envelope of 
existing HBPP DSAR licensing basis analyses, as demonstrated by the 
PG&E and Holtec analyses.
    The rupture of multipurpose canister (MPC) dewatering, forced 
helium dehydration or related closure system lines or the 
malfunction of equipment during cask handling operations resulting 
in radiological consequences are bounded by the HBPP DSAR fuel-
handling accident analysis.
    Other design considerations, such as spent fuel pool (SFP) 
thermal, water chemistry and clarity, criticality, and structural, 
were evaluated and determined not to introduce the possibility of a 
new or different kind of accident from any previously evaluated.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. With the Holtec HI-STAR HB System, and the associated design 
and handling procedures, cask drops and other events have been 
precluded through robust load handling systems, providing defense-
in-depth as described in NUREG-0612. Cask tipovers, while not 
considered credible, are shown to be below the 60g limit, preventing 
damage to the contained fuel assemblies (and associated structures), 
and meeting the analysis guidelines of NUREG-0612. As the existing 
licensing basis assumes a nonmechanistic drop damaging the SFP and 
all fuel, the result of this design approach with the minimization 
of drops and the associated structural challenges assure the margin 
of safety has been maintained.
    Other HBPP licensing-basis events, such as the drop of a spent 
fuel assembly, have not been affected by these changes and remain 
bounding events. Revision of HBPP procedures implementing the 
control of heavy loads to incorporate the additional restrictions on 
heavy loads movement will not affect the procedures or methodology 
used and will, therefore, not affect margins.
    Adverse effects from seismic events and/or cask drops or 
tipovers have been evaluated, assuring that the fuel, MPC, and 
overpack remain within their design bases. Since design basis 
criteria are fully satisfied, there is no impact on the margin of 
safety.
    The Fire Protection Program will continue to ensure that the 
combustible materials are properly controlled such that the total 
combustibles meet the current program commitments. Thus, there are 
no significant reductions in margin of safety associated with these 
changes.
    Other design considerations, such as SFP thermal, water 
chemistry, criticality, and structural, were evaluated and 
determined to not involve a reduction in a margin of safety.

    Based on the above evaluations, the licensee concludes that the 
activities associated with the above changes present no significant 
hazards consideration under the standards set forth in 10 CFR 50.92 and 
accordingly, a finding by the NRC 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 request involves no significant hazards consideration.
    Attorney for licensee: Richard F. Locke, Esquire, Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Claudia Craig.

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

    Date of amendment request: October 6, 2004
    Brief description of amendments: The proposed change will revise 
the Technical Specification (TS) 3.8.1, ``AC Sources--Operating,'' to 
allow surveillance testing of the onsite diesel generators (DGs) during 
power operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The licensee's analysis 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 design of plant equipment is not being modified by the 
proposed changes. In addition, the DGs and their associated 
emergency loads are accident mitigating features. As such, testing 
of the diesel generators (DGs) themselves is not associated with any 
potential accident-initiating mechanism. Therefore, there will be no 
significant impact on any accident probabilities by the approval of 
the requested changes.
    The changes include an increase in the online time that a DG 
under test will be paralleled to the grid (for SRs [Surveillance 
Requirements] 3.8.1.10 and 3.8.1.14) or unavailable due to testing 
(per SR 3.8.1.13). However, the overall time that the DG is 
paralleled in all modes (outage/non-outage) should remain unchanged. 
As such, the ability of the tested DG to respond to a design basis 
accident [DBA] could be adversely impacted by the proposed changes. 
However, the impacts are not considered significant based, in part, 
on the ability of the remaining DG to mitigate a DBA or provide safe 
shutdown. With regard to SR 3.8.1.10 and SR 3.8.1.14, experience 
shows that testing per these SRs typically does not perturb the 
electrical distribution system and share the same electrical 
configuration alignment as the current monthly surveillance. In 
addition, operating experience and qualitative evaluation of the 
probability of the DG or bus loads being adversely affected 
concurrent with or due to a significant grid disturbance, while the 
DG is being tested, support the conclusion that the proposed changes 
do not involve any significant increase in the likelihood of a 
safety-related bus blackout or damage to plant loads.
    The SR changes that are consistent with TSTF [Technical 
Specification Task Force]-283 have been approved generically and for 
individual Licensees. The on-line tests allowed by the TSTF are only 
to be performed for the purpose of establishing OPERABILITY. 
Performance of these SRs during restricted MODES will require an 
assessment to assure plant safety is maintained or enhanced.
    Deletion of expired TS LCO [Limiting Condition for Operation] 
3.8.1, Required Action A.3, one-time 21-day Completion Time 
allowance for Startup Transformer XST2 preventive maintenance is an 
administrative change only. 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 accident from any accident previously evaluated?
    Response: No.
    The proposed changes would not create any new accidents since no 
changes are being made to the plant that would introduce any new 
accident causal mechanisms. Equipment will be operated in the same 
configuration as currently allowed for other DG SRs that allow 
testing during at-power operation. Deletion of expired TS LCO 3.8.1, 
Required Action A.3, one-time 21-day Completion Time allowance for 
Startup Transformer XST2 preventive maintenance is an administrative 
change only. This license amendment request does not impact any 
plant systems that are accident initiators; neither does it 
adversely impact any accident mitigating systems.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not involve a significant reduction in 
the margin of safety. The margin of safety is related to the 
confidence in the ability of the fission product barriers to perform 
their design functions during and following an accident situation. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment system. The proposed changes do not 
directly affect these barriers, nor do

[[Page 76495]]

they involve any significant adverse impact on the DGs which serve 
to support these barriers in the event of an accident concurrent 
with a loss of offsite power. The proposed changes to the testing 
requirements for the plant DGs do not affect the OPERABILITY 
requirements for the DGs, as verification of such OPERABILITY will 
continue to be performed as required (except during different 
allowed MODES). The changes have an insignificant impact on DG 
availability, as continued verification of OPERABILITY supports the 
capability of the DGs to perform their required function of 
providing emergency power to plant equipment that supports or 
constitutes the fission product barriers. Only one DG is to be 
tested at a time, so that the remaining DG will be available to 
safely shut down the plant if required. Consequently, performance of 
the fission product barriers will not be impacted by implementation 
of the proposed amendment.
    In addition, the proposed changes involve no changes to 
setpoints or limits established or assumed by the accident analysis. 
On this and the above basis, no safety margins will be impacted.
    Deletion of expired TS LCO 3.8.1, Required Action A.3, one-time 
21-day Completion Time allowance for Startup Transformer XST2 
preventive maintenance is an administrative change only.
    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: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Mike Webb, Acting Chief.

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

    Date of amendment request: October 13, 2004.
    Brief description of amendments: The proposed changes will revise 
the Technical Specifications (TSs) to incorporate two topical reports 
used to determine the core operating limits of Comanche Peak Steam 
Electric Station (CPSES), Units 1 and 2, and delete reference to four 
topical reports and a reference to NUREG-0800 that are no longer 
required to support CPSES, Units 1 and 2, core operating limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The licensee's analysis is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change is administrative in nature and as such does 
not impact the condition or performance of any plant structure, 
system or component. The core operating limits are established to 
support Technical Specifications 3.1, 3.2, 3.3, and 3.4. The core 
operating limits ensure that fuel design limits are not exceeded 
during any conditions of normal operation or in the event of any 
Anticipated Operational Occurrence (AOO). The methods used to 
determine the core operating limits for each operating cycle are 
based on methods previously found acceptable by the NRC and listed 
in TS section 5.6.5.b. Application of these approved methods will 
continue to ensure that acceptable operating limits are established 
to protect the fuel cladding integrity during normal operation and 
AOOs. The requested Technical Specification changes do not involve 
any plant modifications or operational changes that could affect 
system reliability, performance, or possibility of operator error. 
The requested changes do not affect any postulated accident 
precursors, do not affect any accident mitigation systems, and do 
not introduce any new accident initiation mechanisms.
    As a result, the proposed change to the CPSES Technical 
Specifications does not involve any increase in the probability or 
the consequences of any accident or malfunction of equipment 
important to safety previously evaluated since neither accident 
probabilities nor consequences are being affected by this proposed 
administrative change.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change is administrative in nature, and therefore 
does not involve any changes in station operation or physical 
modifications to the plant. In addition, no changes are being made 
in the methods used to respond to plant transients that have been 
previously analyzed. No changes are being made to plant parameters 
within which the plant is normally operated or in the setpoints, 
which initiate protective or mitigative actions, and no new failure 
modes are being introduced.
    Therefore, the proposed administrative change to the CPSES 
Technical Specifications does not create the possibility of a new or 
different kind of accident or malfunction of equipment important to 
safety from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change is administrative in nature and does not 
impact station operation or any plant structure, system or component 
that is relied upon for accident mitigation. Furthermore, the margin 
of safety assumed in the plant safety analysis is not affected in 
any way by the proposed administrative change.
    Therefore, the proposed change to the CPSES Technical 
Specifications does not involve any 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: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Michael Webb, Acting Chief.

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

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No. 2, St. Lucie County, Florida

    Date of amendment request: November 8, 2004.
    Description of amendment request: To revise Technical Specification 
Section 4.4.5.4 to modify the definitions of steam generator (SG) tube 
``Plugging Limit'' and ``Tube Inspection.''
    Date of publication of individual notice in the Federal Register: 
November 24, 2004 (69 FR 68408).
    Expiration date of individual notice: January 24, 2005.

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.

[[Page 76496]]

The Commission has made appropriate findings as required by the Act and 
the Commission's rules and regulations in 10 CFR Chapter I, which are 
set forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan
    Date of application for amendment: April 1, 2004.
    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: November 29, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 163.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 24, 2004 (69 FR 
52037).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 29, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 7, 2004, as supplemented by 
letters dated November 12 and 18, 2004.
    Description of amendment request: The amendment revised the Safety 
Limit Minimum Critical Power Ratio in Technical Specification 2.1.1.2 
to reflect the results of cycle-specific calculations performed for 
Fermi 2 operating Cycles 10 and 11.
    Date of issuance: November 30, 2004.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup for Fermi 2 Cycle 11 operation.
    Amendment No.: 164.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. (November 9, 2005; 69 FR 64986) The notice 
provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. No comments have been received. The notice also 
provided an opportunity to request a hearing by January 10, 2005, but 
indicated that if the Commission makes a final NSHC determination, any 
such hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a safety evaluation dated November 30, 2004.
    Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: L. Raghavan.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), 
Beaver County, Pennsylvania

    Date of application for amendments: January 26, 2004, as 
supplemented September 13, 2004.
    Brief description of amendments: These amendments authorized 
changes to the BVPS-1 and 2 Updated Final Safety Analysis Reports 
(UFSARs) to revise the level of the Ohio River that is assumed at the 
onset of an accident during power operation to be 654.0' mean sea level 
(msl) instead of 649.0' msl for BVPS-1 and 2. The proposed change is 
consistent with current Technical Specification 3.7.5.1, which requires 
the plant to shut down when the Ohio River reaches a level below 654.0' 
msl.
    Date of issuance: November 29, 2004.
    Effective date: As of the date of issuance and shall submit the 
changes authorized by these amendments with the next update of the 
UFSARs in accordance with 10 CFR 50.71(e).
    Amendment Nos.: 264 and 145.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
authorize changes to the UFSARs.
    Date of initial notice in Federal Register: March 16, 2004 (69 FR 
12369).
    The supplement dated September 13, 2004, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 29, 2004 .
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 21, 2003, as 
supplemented by letters dated May 18, and August 23, 2004.
    Brief description of amendments: These amendments revised the St. 
Lucie Unit 1 and 2 Technical Specifications (TSs) to eliminate certain 
pressure sensor response time testing requirements. Elimination of 
these tests is discussed in the Combustion Engineering Owners Group 
Topical Report CE NPSD-1167, Revision 2, ``Elimination of Pressure 
Sensor

[[Page 76497]]

Response Time Testing Requirements,'' which was approved by the NRC 
staff in letters dated July 24, 2000, and December 5, 2000. 
Specifically, these amendments revise the St. Lucie Units 1 and 2 TS 
Definitions 1.12, ``Engineered Safety Features Response Time,'' and 
1.26, ``Reactor Protection System Response Time.''
    Date of Issuance: November 30, 2004.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 195 and 137
    Renewed Facility Operating License Nos. DPR-67 and NPF-16: 
Amendments revised the TSs.
    Date of initial notice in Federal Register: September 28, 2004 (69 
FR 57675).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 30, 2004.
    No significant hazards consideration comments received: No.

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: November 26, 2002, as 
supplemented by letters dated September 8, 2003, October 30, 2003, June 
21, 2004, and October 8, 2004.
    Brief description of amendments: The amendments increased the total 
spent fuel wet storage capacity for each unit, by adding a spent fuel 
storage rack in the cask area in each unit's spent fuel pool. Each rack 
increased both units' storage capacity by 131 fuel assemblies. The 
amendments also included the addition of the design of the racks in 
Section 5.6.1.1.c of the Technical Specifications (TSs), and revised 
the stated spent fuel capacity in TS Section 5.6.3 and the location 
called out in the Design Features Sections 5.6.1.1a and b of the TSs 
referring to Updated Final Safety Analysis Report Appendix 14D rather 
the Westinghouse Report WCAP-14416-P.
    Date of issuance: November 24, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos: 226 and 222.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the TSs.
    Date of initial notice in Federal Register: January 28, 2003 (69 FR 
4246). The supplemental letters provided clarifying information that 
did not expand the scope of the original application or change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in an Environmental Assessment dated October 17, 2003, and in a Safety 
Evaluation dated November 24, 2004.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station, Unit No. 1, Oswego County, New York

    Date of application for amendment: August 17, 2004.
    Brief description of amendment: The amendment revised Section 
3.3.1, ``Oxygen Concentration,'' of the Technical Specifications to add 
a new action, allowing 24 hours to restore the oxygen concentration 
within the limit of <4% by volume if the limit is exceeded when the 
reactor is operating in the power operating condition.
    Date of Issuance: November 29, 2004.
    Effective date: November 29, 2004 and shall be implemented within 
15 days of issuance.
    Amendment No.: 185.
    Facility Operating License No. DPR-63: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 2004 (69 FR 
53110).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated November 29, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 12, 2003, and its 
supplements dated April 23, June 4, and August 30, 2004.
    Brief description of amendments: The amendments increase the 
current steam generator narrow range water level-low low setpoints from 
greater or equal to 7.0 percent allowable value and 7.2 percent nominal 
trip setpoint to greater than or equal to 14.8 percent allowable value 
and 15.0 percent nominal trip setpoint. The reactor trip setpoint is 
specified in TS Table 3.3.1-1, ``Reactor Trip System Instrumentation,'' 
and the actuation setpoint to start the auxiliary feedwater pumps is 
specified in TS Table 3.3.2-1, ``Engineered Safety Feature Actuation 
System Instrumentation.''
    Date of issuance: December 2, 2004.
    Effective date: December 2, 2004, and shall be implemented within 
90 days from the date of issuance.
    Amendment Nos.: Unit 1-178; Unit 2-180.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 25, 2003 (68 
FR 66138) The April 23, June 4, and August 30, 2004, supplemental 
letters provided additional clarifying information, did not expand the 
scope of the application as originally noticed, and did not change the 
staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 2, 2004.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, 
Pennsylvania

    Date of application for amendments: December 5, 2003, as 
supplemented by letter dated June 4, 2004.
    Brief description of amendments: The amendments revised SSES 1 and 
2 Technical Specifications (TSs) by adding a requirement to apply 
linear heat generation (LHGR) limits if the main turbine bypass system 
becomes inoperable. The proposed changes clarify TS 3.7.6 to state that 
both minimum critical power ratio and LHGR limits for an inoperable 
main turbine bypass system are required if the system becomes 
inoperable.
    Date of issuance: December 3, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 218 and 193.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 6, 2004 (69 FR 
698). The supplement dated June 6, 2004, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 3, 2004.
    No significant hazards consideration comments received: No.

[[Page 76498]]

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: August 12, 2004, as superseded by letter 
dated October 5, 2004, as supplemented by letter dated October 11, 
2004.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3/4.4.5, in conjunction with the new administrative 
control TS 6.8.3.o and reporting requirement TS 6.9.1.7, to establish a 
new programmatic, largely performance-based framework for ensuring SG 
tube integrity. The reactor coolant system leakage requirements of TS 
3.4.6.2 are also revised.
    Date of issuance: November 24, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 1--164; Unit 2--154.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 31, 2004 (69 FR 
53113). The October 5, 2004, letter which superseded the August 12, 
2004, letter and the supplement dated October 11, 2004, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not 
significantly 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 November 24, 2004.
    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: August 26, 2004.
    Brief description of amendments: The amendments eliminate the 
requirements in the TS associated with hydrogen recombiners and 
hydrogen 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).
    Date of issuance: November 30, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 1--165; Unit 2--155.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 28, 2004 (69 
FR 57996).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 30, 2004.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 13th day of December 2004.

    For the Nuclear Regulatory Commission.
James E. Lyons,
Deputy Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 04-27614 Filed 12-20-04; 8:45 am]
BILLING CODE 7590-01-P