[Federal Register Volume 69, Number 206 (Tuesday, October 26, 2004)]
[Notices]
[Pages 62467-62485]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-23664]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as 
amended

[[Page 62468]]

(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 October 1, 2004 through October 14, 2004. 
The last biweekly notice was published on October 12, 2004 (69 FR 
60677).

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 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/ 
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.

[[Page 62469]]

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

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

    Date of amendment request: April 30, 2004.
    Description of amendment request: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
technical specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of title 10 of the Code of Federal 
Regulations (10 CFR), part 50, Section 50.65(a)(4). Limiting Condition 
for Operation (LCO) 3.0.4 exceptions in individual TS would be 
eliminated, several notes or specific exceptions are revised to reflect 
the related changes to LCO 3.0.4, and Surveillance Requirement (SR) 
3.0.4 is revised to reflect the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The Nuclear Regulatory 
Commission (NRC) staff issued a notice of opportunity for comment in 
the Federal Register on August 2, 2002 (67 FR 50475), on possible 
amendments concerning TSTF-359, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability 
of the following NSHC determination in its application dated April 30, 
2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:

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

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The 
addition of a requirement to assess and manage the risk introduced 
by this change will further minimize possible concerns. Therefore, 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

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

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Thus, this change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.

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

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS LCO. The 
risk associated with this allowance is managed by the imposition of 
required actions that must be performed within the prescribed 
completion times. The net effect of being in a TS condition on the

[[Page 62470]]

margin of safety is not considered significant. The proposed change 
does not alter the required actions or completion times of the TS. 
The proposed change allows TS conditions to be entered, and the 
associated required actions and completion times to be used in new 
circumstances. This use is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk. 
The change also eliminates current allowances for utilizing required 
actions and completion times in similar circumstances, without 
assessing and managing risk. The net change to the margin of safety 
is insignificant. Therefore, this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60666.
    NRC Section Chief: Gene Y. Suh.

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

    Date of amendment request: December 9, 2003.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.7.1, ``Main Steam Safety Valves 
(MSSVs),'' to increase the maximum allowable lift setting on two MSSVs 
on each unit. In addition, the proposed amendment would increase the 
completion time for reducing the Power Level-High Trip setpoint.
    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. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    This license amendment request proposes to increase the upper 
range of the relief setting of the first two Main Steam Safety 
Valves (MSSVs) by 10 psi [pounds per square-inch]. The MSSVs are not 
accident initiators. They are credited with relieving secondary 
system pressure and act as a heat sink for the Reactor Coolant 
System (RCS) when the preferred heat sink is not available. 
Increasing the upper end of the setpoint for the first two MSSVs to 
lift does not affect the steam relieving capacity of the total or 
any combination of MSSVs that lift. This proposed amendment does not 
install any new components or change the physical characteristics of 
the MSSVs. Therefore, the change does not involve a significant 
increase in the probability of an evaluated accident.
    The Updated Final Safety Analysis Report Chapter 14 safety 
analyses were reviewed considering the change to the upper end of 
the lift settings range of the first two MSSVs. The analyses show 
that increasing the upper end of the lift setting range does not 
exceed the pressure limits of the reactor coolant or main steam 
systems, nor the radiological consequences anticipated by the safety 
analyses. Therefore, the change will not involve a significant 
increase in the consequences of an evaluated accident.
    This proposed amendment will also increase the Technical 
Specification Completion Time to reset the Power Level-High Trip 
from 12 hours to 36 hours. The purpose of the Power Level-High Trip 
is to trip the reactor if reactor power exceeds a set value, and is 
required by Technical Specifications to be reset according to the 
number of MSSVs remaining operable. The trip is not an accident 
initiator but is a signal that responds to an accident condition. 
Therefore, the change does not involve a significant increase in the 
probability of an evaluated accident.
    Reducing the setpoint of the Power Level-High Trip within the 
time allotted by Technical Specifications provides additional 
assurance that the MSSVs will be able to perform their design 
function by keeping the reactor power within the ability of the 
MSSVs to relieve steam volume. There is low probability of a 
transient that could result in steam generator overpressure during 
the proposed 36 hours to reset the Power Level-High Trip. Therefore, 
this change does not involve a significant increase in the 
consequences of an evaluated accident.
    Therefore, this proposed license amendment does not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    2. Would not create the possibility of a new or different [kind] 
of accident from any accident previously evaluated.
    The proposed amendment will increase the upper end of the lift 
pressure for the first two MSSVs and increase the Technical 
Specification Completion Time to reset the Power Level-High Trip 
setpoint.
    The proposed amendment does not involve a physical alteration of 
the plant or change the plant configuration. It does not require any 
new or unusual operator actions. The amendment does not alter the 
way any structure, system, or component functions and does not alter 
the manner in which the plant is operated. It does not introduce any 
new failure modes.
    Therefore, this proposed license amendment does not create the 
possibility of a new or different [kind] of accident from any 
accident previously evaluated.
    3. Would not involve a significant reduction in [a] margin of 
safety.
    The margin of safety in this case is that the MSSVs release 
sufficient steam to relieve pressure in the secondary system and to 
act as a heat sink to prevent over-pressurization of the RCS when 
the preferred heat sink is not available. Increasing the upper end 
of the setpoint for the first two MSSVs to lift does not affect the 
steam relieving capacity of the total or any combination of MSSVs 
that lift. Potential delay in the opening of the first two MSSVs 
does not result in exceeding the pressure limits of the reactor 
coolant or main steam systems.
    Reducing the Power Level-High Trip setpoint within the specified 
time limit provides additional assurance that the MSSVs will be able 
to perform their design function by keeping the reactor power within 
the ability of the MSSVs to relieve steam volume. A completion time 
of 36 hours to lower the Power Level-High Trip setpoint is based on 
a reasonable time to correct the MSSV inoperability, operating 
experience in resetting all channels of a protective function, and 
on the low probability of the occurrence of a transient that could 
result in steam generator overpressure during this period.
    Therefore, this proposed license amendment 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: James M. Petro, Jr., Esquire, Counsel, 
Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor, 
Baltimore, MD 21202.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendments request: August 16, 2004.
    Description of amendments request: The proposed change adds topical 
report NEDE-32906P-A, ``TRACG Application for Anticipated Operational 
Occurrences (AOO) Transient Analyses,'' to the documents listed in 
Technical Specification (TS) 5.6.5 describing the approved 
methodologies used to determine the core operating limits. Unit 2 will 
be unable to resume power operation following Refueling Outage 16 
without NRC approval for inclusion of the TRACG methodology in TS 
5.6.5.b.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?

[[Page 62471]]

    Response: No.
    The proposed change to TS 5.6.5.b will add General Electric 
Nuclear Energy topical report NEDE-32906P-A, ``TRACG Application for 
Anticipated Operational Occurrences (AOO) Transient Analyses,'' to 
the list of documents describing approved methodologies for 
determining core operating limits. NRC review and acceptance of the 
TRACG methodology is documented in an October 22, 2001, letter and 
associated safety evaluation issued to General Electric Nuclear 
Energy (i.e., refer to ADAMS Accession Numbers ML012740390 and 
ML012740161). Analyzed events are assumed to be initiated by the 
failure of plant structures, systems, or components. The core 
operating limits, which are developed using the topical report being 
added, ensure that the integrity of the fuel will be maintained 
during normal operations and that design requirements will continue 
to be met. The proposed change does not involve physical changes to 
any plant structure, system, or component. Therefore, the 
probability of occurrence for a previously analyzed accident is not 
significantly increased.
    The consequences of a previously analyzed accident are dependent 
on the initial conditions assumed for the analysis, the behavior of 
the fuel during the analyzed accident, the availability and 
successful functioning of the equipment assumed to operate in 
response to the analyzed event, and the setpoints at which these 
actions are initiated. Use of the analytical methodologies described 
in the topical report being added to TS 5.6.5.b will ensure that 
applicable design and safety analyses acceptance criteria are met. 
Use of these NRC-approved methodologies does not affect the 
performance of any equipment used to mitigate the consequences of an 
analyzed accident. As a result, no analysis assumptions are violated 
and there are no adverse effects on the factors that contribute to 
offsite or onsite dose as the result of an accident. Use of the 
approved methodologies described in the topical report being added 
to TS 5.6.5.b ensures that plant structures, systems, or components 
are maintained consistent with the safety analysis and licensing 
bases. Based on this evaluation, there is no significant increase in 
the consequences of a previously analyzed event.
    Therefore, the proposed change adding General Electric Nuclear 
Energy licensing topical report NEDE-32906P-A to the TS 5.6.5.b list 
of documents describing approved methodologies for determining core 
operating limits does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change adding licensing topical report NEDE-32906P-
A to TS 5.6.5.b, and the use of the analytical methods described 
therein, does not involve any physical alteration of plant systems, 
structures, or components, other than allowing for fuel and core 
designs in accordance with NRC approved methodologies. The proposed 
methodology continues to meet applicable criteria for core operating 
limit analysis. No new or different equipment is being installed. No 
installed equipment is being operated in a different manner. There 
is no alteration to the parameters within which the plant is 
normally operated or in the setpoints that initiate protective or 
mitigative actions. As a result no new failure modes are being 
introduced.
    Therefore, the proposed change adding General Electric Nuclear 
Energy licensing topical report NEDE-32906P-A to the TS 5.6.5.b list 
of documents describing approved methodologies for determining core 
operating limits does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established through the design of the 
plant structures, systems, and components, through the parameters 
within which the plant is operated, through the establishment of the 
setpoints for the actuation of equipment relied upon to respond to 
an event, and through margins contained within the safety analyses. 
The proposed change adding General Electric Nuclear Energy licensing 
topical report NEDE-32906P-A to the TS 5.6.5.b list of documents 
describing approved methodologies for determining core operating 
limits does not impact the condition or performance of structures, 
systems, setpoints, and components relied upon for accident 
mitigation. The proposed change does not significantly impact any 
safety analysis assumptions or results. Therefore, the proposed 
change adding topical report NEDE-32906P-A to the TS 5.6.5.b list of 
documents describing approved methodologies for determining core 
operating limits does not result in a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief (Acting): Michael L. Marshall.

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

    Date of amendment request: September 22, 2004.
    Description of amendment request: The proposed change will revise 
Columbia Generating Station's licensing basis by replacing the current 
plant-specific reactor pressure vessel (RPV) material surveillance 
program with the Boiling Water Reactor Vessels and Internals Project 
(BWRVIP) Integrated Surveillance Program (ISP). Specifically, the 
proposed amendment would revise Columbia's Final Safety Analysis Report 
(FSAR) to include participation in the ISP as described in the program 
document BWRVIP-86-A, ``BWR Vessel and Internals Project Updated BWR 
Integrated Surveillance Program (ISP) Implementation Plan,'' dated 
October 2002.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated
    The proposed change implements an ISP program that meets the 
requirements of 10 CFR 50, Appendix H, Paragraph III.C, 
``Requirements for an Integrated Surveillance Program.'' The 
proposed ISP program ensures the same level of RPV integrity as 
Columbia's current material surveillance program. Implementation of 
the proposed ISP is not a precursor or initiator of any previously 
evaluated accident. No physical changes to Columbia Generating 
Station are involved with the proposed change. The proposed change 
will not cause the RPV or interfacing systems to be operated outside 
of any design limit or testing limit, and will not alter any 
assumptions or initial conditions previously used in evaluating the 
radiological consequences of an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change revises the licensing basis for Columbia 
Generating Station to reflect participation in the BWRVIP ISP. The 
NRC has approved the ISP as an acceptable material surveillance 
program pursuant to 10 CFR 50, Appendix H, paragraph III.C. No 
physical changes to the plant are associated with the proposed 
change. No changes in design or operation of any system, structure, 
or component will be made as a result of the proposed change. The 
ISP is an alternative monitoring program and cannot create a new 
failure mode or a new or different kind of accident from any 
previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Compliance with RPV material surveillance program requirements 
specified in 10 CFR 50, Appendix H and the fracture toughness 
requirements contained in 10 CFR

[[Page 62472]]

50, Appendix G ensure an adequate margin of safety exists in the 
fracture toughness of RPV beltline ferritic materials during any 
condition of normal operation, anticipated operational occurrence, 
and system hydrostatic tests. Implementation of the proposed ISP has 
been evaluated to meet the requirements of 10 CFR 50, Appendix H and 
this margin of safety is not impacted. Compliance with the 
requirements of 10 CFR 50, Appendix G will not be affected by this 
proposed change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: September 27, 2004.
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) 5.6.1, ``Occupational Radiation 
Exposure Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated September 27, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change eliminates the TS reporting requirements to 
provide a monthly operating report of shutdown experience and 
operating statistics if the equivalent data is submitted using an 
industry electronic database. It also eliminates the TS reporting 
requirement for an annual occupational radiation exposure report, 
which provides information beyond that specified in NRC regulations. 
The proposed change involves no changes to plant systems or accident 
analyses. As such, the change is administrative in nature and does 
not affect initiators of analyzed events or assumed mitigation of 
accidents or transients. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.

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

    The proposed change does not involve a physical alteration of 
the plant, add any new equipment, or require any existing equipment 
to be operated in a manner different from the present design. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.

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

    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in the margin of safety.

    Based upon the reasoning presented above, the requested change does 
not involve significant hazards consideration.
    Attorney for licensee: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: September 30, 2004.
    Description of amendment request: The proposed license amendment 
request would change the technical specifications and the Final Safety 
Analysis Report to revise the Columbia Generating Station's licensing 
and design bases to reflect the application of the alternative source 
term (AST) methodology with an exception. That exception is TID-14844, 
``Calculation of Distance Factors for Power and Test Reactor Sites,'' 
which will continue to be used as the radiation dose basis for 
equipment qualification, and radiation zone maps/shielding 
calculations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The alternative source term does not affect the design or 
operation of the facility in a manner that would impact the 
probability of an accident previously evaluated. Assumed performance 
requirements of the system structures and components are within 
existing design capability. The manner in which the systems are 
required to operate has not changed.
    Once the occurrence of an accident has been postulated, the new 
source term is an input to evaluate the consequences. The 
implementation of the alternative source term methodology has been 
evaluated in revisions to the analyses of the following limiting 
design basis accidents at Columbia Generating Station:

 Control Rod Drop Accident
 Fuel Handling Accident
 Main Steam Line Break Accident
 Loss of Coolant Accident

    This amendment request includes changes to the Technical 
Specifications based on assumptions in the accident analyses. The 
results of these analyses demonstrate that, with the requested 
changes, the dose consequences of these limiting events are within 
the regulatory limits provided by the NRC for use with the 
alternative source term.
    A new license and design basis analysis on secondary containment 
drawdown is provided to resolve a Justification for Continued 
Operation. The consequences, based on alternative source term 
methodology, remain within regulatory limits. This change to the 
licensing and design basis does not result in a significant increase 
in consequences.
    Alternative source term methodology has been applied to resolve 
the Unresolved Safety Question on control room unfiltered air 
inleakage. The accident analyses results show, with the increased 
unfiltered air inleakage, the control room operator doses remain 
within regulatory limits.
    Therefore, approval of the proposed amendment request does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The requested changes are based on accident analyses. System[,] 
structure and component performance assumptions included in the 
accident analyses result in doses within regulatory limits. Use of 
these performance assumptions does not:
     Require the installation of any new equipment,
     Require the modification of any existing equipment,
     Change the manner in which the equipment is required to 
be operated,
     Assume equipment performance outside existing design 
capabilities, or
     Require new operator actions.
    Therefore Energy Northwest application of the alternative source 
term methodology does not create any new accident initiators or 
precursors of a new or different kind of accident.

[[Page 62473]]

    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The changes proposed are associated with the implementation of a 
new licensing basis for Columbia Generating Station. Approval of a 
basis change from the original source term developed in accordance 
with TID-14844 to a new alternative source term as described in RG 
[Regulatory Guide] 1.183 is requested. The results of the accident 
analyses revised in support of this submittal, and the requested 
Technical Specification changes, are subject to revised acceptance 
criteria. These analyses have been performed using conservative 
methodologies.
    Safety margins and analytical conservatisms have been evaluated 
and are satisfied. The analyzed accidents have been carefully 
selected and margin has been retained to ensure that the analyses 
adequately bound postulated event scenarios. The dose consequences 
of these limiting design basis accidents are within the acceptance 
criteria found in the applicable regulatory requirements and 
guidance. These requirements and guidance are presented in 10 CFR 
50, App. A, 10 CFR 50.67, GDC [General Design Criterion] 19, and RG 
1.183.
    The proposed changes can be made while still satisfying 
regulatory requirements and review criteria, with margin. The 
changes continue to ensure that the doses at the exclusion area and 
low population zone boundaries, as well as the control room, are 
within the corresponding regulatory limits. Therefore, operation of 
Columbia Generating Station in accordance with the requested 
amendment does not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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 
revise Technical Specifications (TSs) to allow a one-time interval 
extension of no more than five years for the Type A, Integrated Leakage 
Rate Test (ILRT) of the primary containment. The proposed amendment 
would also correct the TSs to remove a reference to an obsolete 
alphanumeric identifier in TS 4.7.A.2.a, and reformat existing text on 
TS Pages 3/4.7-4 and 3/4.7-5 to improve consistency in its 
presentation.
    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 changes do not modify the design or operation of the 
containment. Therefore, the proposed changes, therefore, will not 
increase the probability of accidents previously evaluated.
    The proposed extension to Type A, ILRT testing does not involve a 
significant increase in the consequences of an accident. Research 
documented in NUREG-1493 has found that Type A tests identify only a 
few potential containment leakage paths that also cannot be identified 
by Type B and C tests. The leaks that have been found by Type A tests 
have only been marginally above existing requirements. The NUREG then 
concluded that reducing the Type A testing frequency to once every 20 
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 Pilgrim individual plant examination (IPE) 
internal events model that concluded the radiological consequences are 
low to negligible, and remain below regulatory limits. Therefore, any 
potential change in the radiological consequences is not considered 
significant.
    The proposed correction to remove the alphanumeric identifier 
(i.e., definition 1.U), which is no longer used in the TSs, from the 
statements regarding the applicability of surveillance frequency to 
leak rate tests is editorial in nature. Likewise, the proposed 
formatting changes to existing information to improve its presentation 
are also editorial in nature. Since these changes are administrative in 
nature, they cannot increase the probability or consequences of 
previously analyzed accidents.
    Therefore, since the radiological consequences are below the 
regulatory limits and the probability of an accident is unchanged, the 
proposed changes do 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?
    There are no new plant operation modes or physical modifications 
being proposed. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any previously 
analyzed.
    3. Does the change involve a significant reduction in a margin of 
safety?
    The proposed revisions to the TSs add a one-time, 5-year extension 
to the current interval of 10 years from the last Type A test. 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. 
The NUREG also 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 percent of the potential leakage paths are detected 
by Type C testing. This was further confirmed by a plant-specific risk 
assessment using the current Pilgrim IPE internal events model. 
Therefore, by meeting applicatory regulatory limits, any potential 
decrease in margin of safety would not be considered significant.
    The proposed correction to remove the alphanumeric identifier 
(i.e., definition 1.U), which is no longer used in Pilgrim TSs, from 
the statements regarding the applicability of surveillance frequency to 
leak rate tests is editorial in nature. Likewise, the proposed 
formatting changes to existing information to improve its presentation 
are also editorial in nature. As these changes are administrative in 
nature, the proposed changes do not involve a significant reduction in 
the 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 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: Daniel Collins, Acting.

[[Page 62474]]

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

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

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

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

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

    Date of amendment request: April 30, 2004.
    Description of amendment request: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
technical specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of Title 10 of the Code of Federal 
Regulations (10 CFR), Part 50, Section 50.65(a)(4). Limiting Condition 
for Operation (LCO) 3.0.4 exceptions in individual TS would be 
eliminated, several notes or specific exceptions are revised to reflect 
the related changes to LCO 3.0.4, and Surveillance Requirement (SR) 
3.0.4 is revised to reflect the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The Nuclear Regulatory 
Commission (NRC) staff issued a notice of opportunity for comment in 
the Federal Register on August 2, 2002 (67 FR 50475), on possible 
amendments concerning TSTF-359, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability 
of the following NSHC determination in its application dated April 30, 
2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:

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

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The 
addition of a requirement to assess and manage the risk introduced 
by this change will further minimize possible concerns. Therefore, 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

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

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Thus, this change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.

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

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS LCO. The 
risk associated with this allowance is managed by the imposition of 
required actions that must be performed within the prescribed 
completion times. The net effect of being in a TS condition on the 
margin of safety is not considered significant. The proposed change 
does not alter the required actions or completion times of the TS. 
The proposed change allows TS conditions to be entered, and the 
associated required actions and completion times to be used in new 
circumstances. This use is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk. 
The change also eliminates current allowances for utilizing required 
actions and completion times in similar circumstances, without 
assessing and managing risk. The net change to the margin of safety 
is insignificant. Therefore, this change does not involve a 
significant reduction in a margin of safety.

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

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

    Date of amendment request: May 20, 2004.
    Description of amendment request: The proposed changes would modify 
the Limerick Generating Station (LGS), Units 1 and 2, Technical 
Specifications (TSs) to support activation of the trip outputs of the 
previously-installed Oscillation Power Range Monitor (OPRM) portion of 
the Power Range Neutron Monitoring (PRNM) system. Specifically, the 
proposed changes would revise LGS TS 2.2.1, ``Reactor Protection System 
Instrumentation Setpoints,'' TS 3/4.3.1, ``Reactor Protection System 
Instrumentation,'' TS 3/4.3.6, ``Control Rod Block Instrumentation,'' 
TS 3/4.4.1, ``Recirculation System'' and their associated Bases.
    The proposed changes would also revise TS 6.9.1, ``Routine 
Reports,'' and delete interim corrective action requirements from the 
Recirculation System TS.
    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. This modification has no impact on any of the 
previously installed PRNM functions. Plant operation in portions of 
the former restricted zone may potentially cause a marginal increase 
in the probability of occurrence of an instability event. This 
potential increase in probability is acceptable

[[Page 62475]]

because the OPRM Upscale Function will automatically detect the 
condition and initiate a reactor scram before the Minimum Critical 
Power Ratio (MCPR) Safety Limit is reached. Consequences of the 
potential instability event are reduced because of the more reliable 
automatic detection and suppression of an instability event, and 
elimination of dependence on the manual operator actions.
    The change to align the operability requirements for the 
Intermediate Range Monitor (IRM) rod block function with those for 
the corresponding IRM Reactor Protection System (RPS) functions 
affects only the rod block function. The justification for the 
change to IRM RPS function (done with the original PRNM 
modification) concluded that the RPS change would not increase the 
probability of occurrence of an accident previously evaluated; 
therefore, changing the associated rod block to align with those 
requirements would not do so either.
    Therefore, the proposed changes do not involve a significant 
increase in the probability of 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 modification replaces procedural actions that 
were established to avoid operating conditions where reactor 
instabilities might occur with an NRC approved automatic detect and 
suppress function.
    Potential failures in the OPRM Upscale Function could result in 
either failure to take the required mitigating action or an 
unintended reactor scram. These are the same potential effects of 
failure of the operator to take the appropriate action under the 
current procedural directions. The net effect of the modification 
changes the method by which an instability event is detected and by 
which mitigating action is initiated, but does not change the type 
of stability event that could occur. The effects of failure of the 
OPRM equipment are limited to reduced or failed mitigation, but such 
failure cannot cause an instability event or other type of accident.
    The change to align the operability requirements for the IRM rod 
block function with those for the corresponding IRM RPS functions 
affects only the rod block function. The justification for the 
change to IRM RPS function (done with the original PRNM 
modification) concluded that the RPS change could not create the 
possibility of a new type of accident; therefore, changing the 
associated rod block to align with those requirements would not do 
so either.
    Therefore, since no radiological barrier will be challenged as a 
result of activating the OPRM Upscale Function, it is concluded that 
the proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The current safety analysis assumes that the 
existing procedural actions are adequate to prevent an instability 
event. As a result, there is currently no quantitative or 
qualitative assessment of an instability event with respect to its 
impact on the MCPR Safety Limit.
    The OPRM Upscale function is being implemented to automate the 
detection (via direct measurement of neutron flux) and subsequent 
suppression (via scram) of an instability event prior to exceeding 
the MCPR Safety Limit. The OPRM Upscale function provides a trip 
output of the same type as currently used for the Average Power 
Range Monitor (APRM). Its failure modes and types are identical to 
those for the present APRM output. Currently, the MCPR Safety Limit 
is not challenged by an instability event since the event is 
``mitigated'' by manual means via the procedural actions, which 
prevent plant operating conditions where an instability event is 
possible. In both methods of mitigation (manual and automated), the 
margin of safety associated with the MCPR Safety Limit is still 
maintained.
    Therefore, based on the fact that the MCPR Safety Limit will 
still be enforced, implementation of the OPRM Upscale function in 
place of the existing manual actions does not reduce the margin of 
safety.
    The IRM rod block function is not considered in any safety 
analysis. As a result, its failure will not affect the margin of 
safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: June 1, 2004.
    Description of amendment request: The proposed changes would 
relocate the operability and surveillance requirements for the reactor 
coolant system safety/relief valve position instrumentation from the 
Limerick Generating Station (LGS) Technical Specifications (TSs) to the 
LGS Technical Requirements Manual (TRM) and plant procedures. 
Specifically, the changes would relocate TSs 3.4.2.c, 4.4.2.1, and the 
associated footnotes to the TRM. Additionally, the ``Safety/Relief 
Valve Position Indicators'' instrumentation would be relocated from 
Tables 3.3.7.5-1 and 4.3.7.5-1 of TSs 3.3.7.5 and 4.3.7.5, respectively 
to the TRM.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No. The failure of the safety/relief valve (SRV) 
position instrumentation is not assumed to be an initiator of any 
analyzed event in the [Updated Final Safety Analysis Report] UFSAR. 
The proposed changes do not alter the physical design of the SRVs or 
any other plant structure, system, or component. The changes would 
remove the [SRV] position indicator operability and surveillance 
requirements from the LGS [TSs], and incorporate requirements 
verbatim for this instrumentation into a licensee-controlled 
document under the control of 10 CFR 50.59.
    The proposed changes conform to NRC regulatory guidance 
regarding the content of plant [TSs] as identified in regulation 10 
CFR 50.36, and NRC publication NUREG-1433, ``Standard Technical 
Specifications--General Electric Plants, BWR/4.''
    Therefore, this proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No. The proposed changes do not alter the physical 
design, safety limits, or safety analysis assumptions, associated 
with the operation of the plant. Accordingly, the proposed changes 
do not introduce any new accident initiators, nor do they reduce or 
adversely affect the capabilities of any plant structure or system 
in the performance of their safety function.
    Therefore, the proposed amendment does 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?
    Response: No. This instrumentation is not needed for manual 
operator actions necessary for safety systems to accomplish their 
safety function for the design basis accident events. The 
instrumentation provides only alarm and SRV position indication, and 
does not provide an input to any automatic trip function. Several 
diverse means are available to monitor SRV position, and operability 
and surveillance requirements will be established in a licensee-
controlled document to assure the reliability of SRV position 
monitoring capability. Changes to these requirements will be subject 
to the controls of regulation 10 CFR 50.59, providing the 
appropriate level of regulatory control.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.


[[Page 62476]]


    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. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Daniel S. Collins, Acting.

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

    Date of application for amendments: June 24, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.1.8, ``Scram Discharge Volume (SDV) 
Vent and Drain Valves,'' to allow a vent or drain line with one 
inoperable valve to be isolated instead of requiring the valve to be 
restored to Operable status within 7 days. Other changes included in 
the application are addressed in a separate notice.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on February 24, 2003 (68 FR 8637), on possible 
amendments to revise the action for one or more SDV vent or drain lines 
with an inoperable valve, including a model safety evaluation and model 
no significant hazards consideration (NSHC) determination, using the 
consolidated line-item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on April 15, 
2003 (68 FR 18294). The licensee affirmed the applicability of the 
model NSHC determination in its application dated June 24, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

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

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any previously evaluated.

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

    The proposed change ensures that the safety functions of the SDV 
vent and drain valves are fulfilled. The isolation function is 
maintained by redundant valves and by the required action to isolate 
the affected line. The ability to vent and drain the SDVs is 
maintained through administrative controls. In addition, the reactor 
protection system will prevent filling of an SDV to the point that 
it has insufficient volume to accept a full scram. Maintaining the 
safety functions related to isolation of the SDV and insertion of 
control rods ensures that the proposed change does not involve a 
significant reduction in the margin of safety.

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

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment requests: September 21, 2004.
    Description of amendment requests: The proposed amendments would 
extend the allowed outage times from 72 hours to 14 days for an 
inoperable emergency diesel generator, an inoperable component cooling 
water system loop, an inoperable essential service water system loop, 
or an inoperable alternate offsite power circuit (69 kilovolt circuit). 
The proposed amendments would also change formats of the affected 
technical specification pages to improve their appearance but not alter 
any requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed changes to the Technical Specifications (TS) will 
extend the allowed outage time (AOT) for a single inoperable 
emergency diesel generator (EDG), one inoperable component cooling 
water (CCW) or essential service water (ESW) loop, or an inoperable 
69 kilovolt (kV) offsite circuit from the current limit of 72 hours 
to 14 days. An independent alternating current (AC) power source 
consisting of two supplemental diesel generators (SDGs) will be 
installed to support the extended AOTs for the EDGs and the CCW and 
ESW systems. The SDGs will supply power to required safe shutdown 
loads in the affected unit.
    The EDGs are backup AC power sources designed to power safe 
shutdown systems in the event of a loss of offsite power. As such, 
the EDGs are not initiators for any accident previously evaluated. 
The CCW and ESW systems provide cooling water to safety-related 
components. This is a support function, and malfunctions of the CCW 
and ESW systems are not initiators of any accidents previously 
analyzed. The 69 kV circuit is an alternate offsite power supply 
that must be manually connected by the control room operators to 
provide power to safety-related buses upon loss of the preferred 
34.5 kV offsite power source. As such, the 69 kV circuit is not an 
initiator for any accident previously evaluated. The AOT extension 
for an inoperable EDG, a CCW or ESW loop, or 69 kV circuit does not 
introduce any failure mechanisms that would initiate a previously 
analyzed accident. Therefore, the proposed change permitting 
extension of the AOTs for the EDG, ESW, CCW, and 69 kV systems do 
not result in a significant increase in the probability of a 
previously evaluated accident.
    The potential effect of the proposed change on the consequences 
of a previously evaluated accident has been considered. There are 
two EDGs per unit, and only one EDG per unit is required to fulfill 
the onsite AC power system safety function. During the extended AOT, 
the redundant EDG will be available to provide AC power to safety-
related components. There are two CCW loops per unit, and only one 
CCW loop per unit is required to fulfill the CCW system safety 
function. During the extended AOT, the redundant CCW loop will be 
available to provide cooling water to safety-related components. 
There are two ESW loops per unit, and only one ESW loop per unit is

[[Page 62477]]

required to fulfill the ESW system safety function for the affected 
unit. During the extended AOT, the redundant ESW loop will be 
available to provide cooling water to the safety-related components. 
The 69 kV offsite circuit is the alternate offsite power source. 
Only one offsite power source is required to fulfill the offsite 
power system safety function. During an extended AOT, the preferred 
offsite source will be available. Thus, the systems affected by the 
proposed amendment will still be capable of performing the safety 
functions needed to mitigate the consequences of an accident as 
previously evaluated.
    The format changes improve appearance, but do not affect any 
requirements.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change consists of increasing the AOTs allowed by 
TS for the EDG, CCW, ESW, and 69 kV systems. Extending existing 
AOTs, does not result in operation of the plant in any new or 
different manner, nor does it create any new accident precursors. 
The format changes improve appearance, but do not affect any 
requirements.
    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?
    Response: No.
    The margins of safety are established through design parameters, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed change does not adversely affect any 
design or operating parameter or any setpoint used in the 
deterministic accident analyses to establish the margin of safety. 
Probabilistic risk assessment methods were used to evaluate the 
risked-based margins of safety for the proposed change. The results 
of these evaluations indicated the proposed AOT extensions combined 
with installation of additional on-site electrical power supplies 
results in a net risk reduction. The format changes improve 
appearance, but do not affect any requirements.
    Therefore, the proposed change will not create a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: L. Raghavan.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: September 7, 2004.
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) 5.9.1b, ``Annual Occupational 
Exposure Report,'' and TS 5.9.1c, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated September 7, 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 proposed change eliminates the TS reporting requirements to 
provide a monthly operating report of shutdown experience and 
operating statistics if the equivalent data is submitted using an 
industry electronic database. It also eliminates the TS reporting 
requirement for an annual occupational radiation exposure report, 
which provides information beyond that specified in NRC regulations. 
The proposed change involves no changes to plant systems or accident 
analyses. As such, the change is administrative in nature and does 
not affect initiators of analyzed events or assumed mitigation of 
accidents or transients. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.

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

    The proposed change does not involve a physical alteration of 
the plant, add any new equipment, or require any existing equipment 
to be operated in a manner different from the present design. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.

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

    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in the margin of safety.

    Based upon the reasoning presented above, the requested change does 
not involve significant hazards consideration.
    Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: September 30, 2004.
    Description of amendment request: The amendment would change 
Technical Specification 4.4.4.2 to expand the range of conditions under 
which quarterly testing of block valves for the pressurizer power 
operated relief valves would be unnecessary.
    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 block valve for the pressurizer power operated relief valve 
is not a potential accident initiator. Therefore, not requiring a 
surveillance of the block valve while it is being used to isolate 
its associated power operated relief valve will not increase the 
probability of an accident previously evaluated. Not requiring the 
surveillance of the block valve may slightly reduce the probability 
of a loss of coolant accident from a stuck open power operated 
relief valve since it will eliminate the challenge to the power 
operated relief valve from the pressure transient that results from 
cycling the block valve.
    If pressurizer spray is not available or is not effective, 
either one of the two pressurizer power operated relief valves may 
be manually actuated to depressurize the reactor coolant system to 
mitigate the consequences of a steam generator tube rupture. Not 
performing the surveillance on the block valve is not relevant to 
the primary system for depressurizing the reactor coolant system 
(pressurizer spray). The block valves have been demonstrated by 
operating experience to be reliable and are also subject to the 
motor-operated valve testing program. Consequently, the proposed 
change does not significantly reduce the confidence that the block 
valve can be opened to permit manual actuation of the power operated 
relief valve to depressurize the reactor coolant system to mitigate 
an accident. Therefore, the proposed change does not involve a 
significant

[[Page 62478]]

increase in the 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 change only affects the performance of the 
surveillance test for the block valve and does not introduce any 
operating configurations not previously evaluated.
    Therefore, STPNOC concludes the proposed change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to the surveillance requirement for the 
block valve for the pressurizer power operated relief valve does not 
affect the assumptions in any accident analyses. There are no 
changes in plant performance parameters associated with the proposed 
change to the surveillance requirement for the block valve. 
Therefore, STPNOC concludes the proposed changes do not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendment involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Michael K. Webb, Acting.

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

    Date of amendment request: September 27, 2004.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TS) to provide consistency between 
Surveillance Requirement (SR) 4.7.1.6 and TS 3.3.5.1 regarding 
atmospheric steam relief valve instrumentation controls. The proposed 
amendment would also correct editorial errors in TS 3.7.1.6.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not involve a significant increase in 
the probability or consequences of a previously evaluated accident. 
The first proposed change only clarifies when SR 4.7.1.6 for the 
automatic controls of the atmospheric steam relief valve is 
applicable. The applicability is already established in TS 3.3.5.1 
and meets the safety analysis. The second proposed change is 
editorial.
    2. Does the proposed change create the possibility of a new or 
different accident from any accident previously evaluated?
    Response: No.
    The first proposed change does not create the possibility of a 
new or different accident from any previously evaluated. The 
proposed change only clarifies when SR 4.7.1.6 for the automatic 
controls of the atmospheric steam relief valve is applicable. The 
applicability is already established in TS 3.3.5.1 and meets the 
safety analysis. The second proposed change is editorial.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The first proposed change does not involve a significant 
reduction in the margin of safety. The proposed change only 
clarifies when SR 4.7.1.6 for the automatic controls of the 
atmospheric steam relief valve is applicable. The applicability is 
already established in TS 3.3.5.1 and meets the safety analysis. The 
second proposed change is editorial.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Michael K. Webb, Acting.

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

    Date of amendment request: September 30, 2004.
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) 6.9.1.2, ``Occupational Radiation 
Exposure Report,'' and TS 6.9.1.5, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated September 30, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--Does the Proposed Change Involve a Significant Increase in 
the Probability or Consequences of an Accident Previously Evaluated?

    The proposed change eliminates the TS reporting requirements to 
provide a monthly operating report of shutdown experience and 
operating statistics if the equivalent data is submitted using an 
industry electronic database. It also eliminates the TS reporting 
requirement for an annual occupational radiation exposure report, 
which provides information beyond that specified in NRC regulations. 
The proposed change involves no changes to plant systems or accident 
analyses. As such, the change is administrative in nature and does 
not affect initiators of analyzed events or assumed mitigation of 
accidents or transients. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.

Criterion 2--Does the Proposed Change Create the Possibility of a New 
or Different Kind of Accident From Any Accident Previously Evaluated?

    The proposed change does not involve a physical alteration of 
the plant, add any new equipment, or require any existing equipment 
to be operated in a manner different from the present design. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.

Criterion 3--Does the Proposed Change Involve a Significant Reduction 
in a Margin of Safety?

    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in the margin of safety.

    Based upon the reasoning presented above, the requested change does 
not involve significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Michael K. Webb, Acting.

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: September 9, 2004.
    Brief description of amendments: The proposed change will revise 
the surveillance requirement (SR) 3.6.6.8

[[Page 62479]]

frequency of every 10 years. Instead, the proposed change to SR 3.6.6.8 
will require verification that spray nozzles are unobstructed 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 reviewed the Licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review 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 Containment Spray System is not considered an initiator of any 
analyzed event. The proposed change does not have a detrimental impact 
on the integrity of any plant structure, system, or component that may 
initiate an analyzed event. The proposed change will not alter the 
operation or otherwise increase the failure probability of any plant 
equipment that can initiate an analyzed accident. This change does not 
affect the plant design. There is no increase in the likelihood of 
formation of significant corrosion products. Due to their location at 
the top of the containment, introduction of foreign material into the 
spray headers is unlikely. Foreign material introduced during 
maintenance activities would be the most likely source for obstruction, 
and verification following such maintenance would confirm the nozzles 
remain unobstructed.
    Consequently, there is no significant increase in the probability 
of an accident previously evaluated.
    The Containment Spray System is designed to address the 
consequences of a LOCA [loss-of-coolant accident]. The Containment 
Spray System is capable of performing its function effectively with the 
single failure of any active component in the system, any of its 
subsystems, or any of its support systems. A plugged nozzle would have 
negligible impact on the capability of the Containment Spray System to 
respond to a Loss of Coolant Accident.
    Therefore, the consequences of an accident previously evaluated are 
not significantly affected by the proposed 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 will not physically alter the plant (no new or 
different type of equipment will be installed) or change the methods 
governing normal plant operation.
    Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The system is not susceptible to corrosion-induced obstruction or 
obstruction from sources external to the system. Maintenance activities 
that could introduce foreign material into the system would require 
subsequent verification to ensure there is no nozzle blockage. The 
spray header nozzles are expected to remain unblocked and available in 
the event that the safety function is required. Therefore, the capacity 
of the system would remain unaffected.
    Therefore the proposed change does not involve a 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 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: Mohan Thadani, 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: September 10, 2004.
    Brief description of amendments: The proposed amendment would 
delete Technical Specification (TS) 5.6.1, ``Occupational Radiation 
Exposure Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated September 10, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    Based upon the reasoning presented above, the requested change does 
not involve significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Michael Webb, Acting.

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 62480]]

page cited. This notice does not extend the notice period of the 
original notice.

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

    Date of amendment request: April 1, 2004.
    Brief description of amendment request: The proposed amendment 
would allow entry into a mode or other specified condition in the 
applicability of a technical specification (TS), while in a condition 
statement and the associated required actions of the TS, provided the 
licensee performs a risk assessment and manages risk consistent with 
the program in place for complying with the requirements of Title 10 of 
the Code of Federal Regulations, Part 50, Section 50.65(a)(4). Limiting 
Condition for Operation (LCO) 3.0.4 exceptions in individual TSs would 
be eliminated, and Surveillance Requirement 3.0.4 revised to reflect 
the LCO 3.0.4 allowance.
    Date of publication of individual notice in Federal Register: 
August 24, 2004 (69 FR 52037).
    Expiration date of individual notice: October 23, 2004.

Florida Power and Light Company, et al., Docket Nos. 50-335, and 50-
389, St. Lucie Plant, Unit No. 1, and Unit No. 2, St. Lucie County, 
Florida

    Date of amendment request: November 21, 2003.
    Description of amendment request: Revise Technical Specifications 
to eliminate certain pressure sensor response time testing requirements 
as discussed in the Combustion Engineering Owners Group Topical Report 
NPSD-1167, Revision 2, Elimination of Pressure Sensor Response Time 
Testing Requirements.''
    Date of publication of individual notice in the Federal Register: 
September 28, 2004 (69 FR 57975).
    Expiration date of individual notice: November 29, 2004.

STP Nuclear Operating Company, Docket No. 50-499, South Texas Project, 
Unit 2, Matagorda County, Texas

    Date of amendment request: September 30, 2004.
    Description of amendment request: The amendment changes TS 4.4.4.2 
to expand the range of conditions under which quarterly testing of 
block valves for the pressurizer power operated relief valves would be 
unnecessary.
    Date of publication of individual notice in Federal Register: 
October 6, 2004 (69 FR 59969).
    Expiration date of individual notice: October 20, 2004 (Comment); 
December 6, 2004 (Hearing).

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, (301) 415-4737 or by e-mail to [email protected].

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

    Date of application for amendment: February 27, 2004, as 
supplemented by letter dated August 11, 2004.
    Brief description of amendment: The amendment revised the Technical 
Specifications, relocating the average power range monitor flux scram 
setting and rod block setting from the to the Core Operating Limits 
Report.
    Date of Issuance: October 4, 2004.
    Effective date: October 4, 2004 and shall be implemented within 60 
days of issuance.
    Amendment No.: 248.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19563).
    The supplement dated August 11, 2004, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the Nuclear 
Regulatory Commission (NRC) staff's original proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated October 4, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 19, 2004.
    Brief description of amendment: The amendment revised the Technical 
Specifications, changing the surveillance requirements associated with 
control rod scram time testing. Specifically, the amendment modified 
the conditions under which scram time testing of control rods is 
required, and added a requirement to perform such testing on a defined 
portion of control rods at a specified frequency during the operating 
cycle.
    Date of Issuance: October 4, 2004.
    Effective date: October 4, 2004 and shall be implemented within 60 
days of issuance.
    Amendment No.: 249.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 27, 2004 (69 FR 
22878).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated October 4, 2004.

[[Page 62481]]

    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 5, 2004, as supplemented 
August 6, 2004.
    Brief description of amendment: This amendment adds a reference to 
the American Society of Mechanical Engineers Code for Operation and 
Maintenance of Nuclear Power Plants in Technical Specification 
Surveillance Requirement 4.0.5.a for the snubbers.
    Date of issuance: October 1, 2004.
    Effective date: October 1, 2004.
    Amendment No. 117.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 22, 2004 (69 FR 
34697). The August 6, 2004, supplement contained clarifying information 
only and did not change the initial proposed no significant hazards 
consideration determination or expand the scope of the initial 
application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 1, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 13, 2003, as 
supplemented July 8, 2003, December 12, 2003, June 4, 2004, July 30, 
2004, and September 16, 2004.
    Brief description of amendment: The amendment approves the use of 
an alternative source term methodology in accordance with Title 10 of 
the Code of Federal Regulations, Part 50, Section 50.67, based on a 
reevaluation of the design-basis loss-of-coolant and fuel handling 
accidents. In addition to related design-basis changes, the amendment 
revises the Technical Specifications to (1) permit an increase in the 
allowable leak rate for the main steam isolation valves (MSIVs), (2) 
increase the allowable secondary containment bypass leakage, (3) delete 
the MSIV leakage control system, and (4) increase the allowed secondary 
containment draw-down time.
    Date of issuance: September 28, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 160.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications and authorizes changes to the Updated Final 
Safety Analysis Report.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28847).
    The July 8, 2003, December 12, 2003, June 4, 2004, July 30, 2004, 
and September 16, 2004, supplemental letters provided additional 
clarifying information that was within the scope of the original 
application and did not change the Nuclear Regulatory Commission 
staff's initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 28, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 30, 2004.
    Brief description of amendment: The amendment revises technical 
specification 3.3.6.2, ``Secondary Containment Isolation 
Instrumentation, Condition C, to add the words ``not met'' to the end 
of the phrase, ``Required Action and associated Completion Time.'' The 
omission of the words ``not met'' was an oversight during the change to 
Improved Standard Technical Specifications, NUREG 1433.
    Date of issuance: January 30, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 161.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 22, 2004, (69 FR 
34698).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 7, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 4, 2004, as supplemented by letter 
dated June 16, 2004.
    Brief description of amendment: The amendment revises the Technical 
Specification requirements by eliminating the requirements associated 
with hydrogen recombiners and hydrogen monitors. These changes support 
implementation of the revisions to Title 10 of the Code of Federal 
Regulations (10 CFR), Section 50.44, ``Standards for Combustible Gas 
Control System In Light-Water-Cooled Power Reactors.'' A notice of 
availability of this TS improvement was published in the Federal 
Register on September 25, 2003 (68 FR 55416).
    Date of issuance: October 4, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of Issuance.
    Amendment No.: 142.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 11, 2004 (69 FR 
26187). The supplement dated June 16, 2004, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 4, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 28, 2003, as supplemented 
on May 20, 2004.
    Brief description of amendment: The amendment revised Technical 
Specification 5.5.6, ``Primary Containment Leakage Rate Testing 
Program,'' to allow a one-time extension of the interval between the 
Type A, integrated leakage rate tests, from 10 years to no more than 15 
years. Therefore, the first Type A test performed after the March 7, 
1995, test shall be performed no later than March 7, 2010.
    Date of issuance: September 28, 2004.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 279.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 27, 2004 (69 FR 
44696).
    The supplement dated May 20, 2004, provided additional information 
that clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a

[[Page 62482]]

Safety Evaluation dated September 28, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 21, 2004.
    Brief description of amendments: The amendments revise the 
technical specifications to add an additional reference as an 
acceptable method for determining the reactor pressure vessel pressure-
temperature limits.
    Date of issuance: October 4, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 139/139, 132/132.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 3, 2004.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 4, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: January 15, 2004, and 
supplemented on June 22, 2004.
    Brief description of amendments: The amendments allow for a one-
time deferral of the Dresden, Units 2 and 3, Appendix J, Type A, 
Integrated Leakage Rate Test (ILRT) to no later than February 27, 2011, 
and July 13, 2009, respectively.
    Date of issuance: October 13, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 210/202.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 16, 2004.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 13, 2004.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois; Docket Nos. 
50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy 
County, Illinois; Docket Nos. 50-254 and 50-265, Quad Cities Nuclear 
Power Station, Units 1 and 2, Rock Island County, Illinois

    Date of application for amendments: November 3, 2003.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.4.1, ``Recirculation Loops Operating,'' by adding 
a limiting condition for operation requirement that the linear heat 
generation rate (LHGR) limits shall be modified for single 
recirculation loop operation as specified in the Core Operating Limits 
Report. The associated TS Bases are also revised to reflect the new 
LHGR limit requirement.
    Date of issuance: October 4, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 167, 153, 209, 201, 221, 216.
    Facility Operating License Nos. NPF-11, NPF-18, DPR-19, DPR-25, 
DPR-29 and DPR-30: The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 6, 2004 (69 FR 
694).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 4, 2004.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, 
Beaver Valley Power Station, Unit No. 1, Beaver County, Pennsylvania

    Date of application for amendment: January 27, 2004, as 
supplemented May 27, 2004.
    Brief description of amendment: The amendment revised TS 3.4.5 to 
allow a one-cycle use of Westinghouse leak-limiting Alloy 800 SG tube 
sleeves as an acceptable SG tube repair. Specifically, surveillance 
requirements 4.4.5.4.a.6 and 4.4.5.4.a.9 are revised to list the 
Westinghouse leak-limiting Alloy 800 sleeves as an acceptable SG tube 
sleeving method in addition to the currently approved Westinghouse 
laser welded sleeves and the former ABB Combustion Engineering tungsten 
inert gas welded sleeves.
    Date of issuance: October 5, 2004.
    Effective date: Within 60 days of the date of issuance and shall 
include the licensee commitments contained in the licensee letters of 
January 27 and May 27, 2004. The commitments shall remain in effect for 
the authorized period of sleeving with Westinghouse Alloy 800 tubes, 
i.e., Cycle 17.
    Amendment No: 260.
    Facility Operating License No. DPR-66: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 16, 2004 (69 FR 
12369).
    The supplement dated May 27, 2004, provided additional information 
that clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the Nuclear 
Regulatory Commission (NRC) staff's original proposed no significant 
hazards consideration determination as published in the Federal 
Register on March 16, 2004 (69 FR 12369).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 5, 2004.
    No significant hazards consideration comments received: No.

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: October 17, 2003.
    Brief description of amendments: These amendments revised the 
action requirements in TS 3/4 6.3 to more clearly define the action 
requirements for inoperable containment isolation valves (CIVs). The 
amendments also allowed under administrative control, the intermittent 
unisolating of penetration flow paths which have previously been 
isolated per the action requirements. The amendments also allowed the 
use of check valves as an isolation device, and an increase in the 
allowed outage time to 72 hours for CIVs associated with closed systems 
inside containment. The amendments also deleted existing surveillance 
requirements (SRs) and provided new SRs similar to those in the 
Improved Standard Technical Specifications.
    Date of issuance: October 5, 2004.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 261 and 143.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 25, 2003 (68 
FR 66136).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 5, 2004.
    No significant hazards consideration comments received: No.

[[Page 62483]]

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: October 29, 2003.
    Brief description of amendments: These amendments relocate specific 
pressure and flow values associated with the high pressure safety 
injection, low pressure safety injection, boric acid makeup, and 
containment spray pumps from the Technical Specification to the St. 
Lucie Units 1 and 2 Updated Final Safety Analysis Reports.
    Date of Issuance: October 6, 2004.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 194, 136.
    Renewed Facility Operating License Nos. DPR-67 and NPF-16: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 6, 2004 (69 FR 
697).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 6, 2004.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: July 6, 2004.
    Brief description of amendment: The amendment revises technical 
specification 3.3.a.2.B, by extending the completion time from 1 hour 
to 24 hours for an accumulator that is inoperable for any reason other 
than failure to meet minimum boron concentration requirements.
    Date of issuance: October 5, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 178.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 2004 (69 FR 
53111).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 5, 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: November 1, 2002, and its 
supplement dated April 2, 2004.
    Brief description of amendments: The amendments (1) change the 
allowances for bypassing and tripping tested channels, (2) remove a 
surveillance requirement for reactor trip system (RTS) turbine trip-
turbine stop valve closure, (3) revise the nominal trip setpoint for 
RTS turbine trip-turbine stop valve closure, (4) revise the allowable 
value and nominal trip setpoint for RTS interlock, (5) and remove and 
relocate the turbine trip function from engineered safety feature 
actuation system turbine trip and feedwater isolation to other 
licensee-controlled documents.
    Date of issuance: September 24, 2004.
    Effective date: September 24, 2004, and shall be implemented within 
120 days from the date of issuance.
    Amendment Nos.: Unit 1-173; Unit 2-175.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 7, 2003 (68 FR 
810).
    The April 2, 2004, supplemental letter 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 September 24, 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: May 29, 2003, as supplemented 
by letter dated December 23, 2003, and May 7, 2004.
    Brief description of amendments: The amendments revise several 
surveillance requirements (SRs) in Technical Specification (TS) 3.8.1 
on alternating current sources for plant operation. The revised SRs 
have notes deleted or modified to adopt in part Staff-approved TSTF-
283, Revision 3, which will allow these revised SRs to be performed, or 
partially performed, in reactor modes that previously were not allowed 
by the TSs. The proposed changes to SRs 3.8.4.7 and 3.8.4.8 for direct 
current sources were withdrawn in the licensee's letter dated May 7, 
2004.
    Date of issuance: September 28, 2004.
    Effective date: September 28, 2004, and shall be implemented within 
60 days of the date of issuance including the incorporation of the 
changes to the Technical Specification Bases for Technical 
Specification 3.8.1 as described in the licensee's letters dated May 29 
and December 23, 2003, and May 7, 2004, and the NRC safety evaluation 
attached to the amendments. This includes the revision of procedures to 
instruct operator action to be taken to manually reset the emergency 
diesel generator, as discussed in Section 4.3 of the licensee's May 29, 
2003, letter.
    Amendment Nos.: Unit 1-174; Unit 2-176.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2003 (68 FR 
40715).
    The December 23, 2003, and May 7, 2004, supplemental letters 
provided additional clarifying information, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 28, 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: October 22, 2003.
    Brief description of amendments: The amendments revise Section 
3.6.3 of the Diablo Canyon Power Plant Technical Specifications to 
extend the local leakage rate testing intervals for the containment 
purge and vent valves with resilient seals from 184 days to 24 months.
    Date of issuance: October 6, 2004.
    Effective date: October 6, 2004, and shall be implemented within 60 
days from the date of issuance.
    Amendment Nos.: Unit 1-175; Unit 2-177.
    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 66139).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 6, 2004.
    No significant hazards consideration comments received: No.

[[Page 62484]]

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: December 19, 2003, and its 
supplement dated May 13, 2004.
    Brief description of amendments: The amendments change Technical 
Specification 5.5.9, ``Steam Generator (SG) Tube Surveillance 
Program,'' to revise the wedge region exclusion zones for outside 
diameter stress corrosion cracking alternate repair criteria (ARC) at 
tube support plate (TSP) intersections and for primary water stress 
corrosion cracking ARC at dented TSP intersections. The new wedge 
region exclusion zones are based on new analyses of loss-of-coolant 
accident plus safe shutdown earthquake loads completed in 2003 using 
plant-specific accident loads.
    Date of issuance: October 6, 2004.
    Effective date: October 6, 2004, and shall be implemented within 60 
days of issuance.
    Amendment Nos.: Unit 1-176; Unit 2-178.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 3, 2004 (69 FR 
5205).
    The May 13, 2004, supplemental letter 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 October 6, 2004.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of application for amendment: November 17, 2003, as 
supplemented July 15, and August 23, 2004.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to delete the primary containment isolation valves 
and instrumentation associated with the permanent removal of the 
reactor vessel head spray piping.
    Date of issuance: October 5, 2004.
    Effective date: As of the date of issuance, to be implemented prior 
to restart from the fall 2004 refueling outage.
    Amendment No.: 152.
    Facility Operating License No. NPF-57: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: January 20, 2004 (69 FR 
2746). The supplements dated July 15, and August 23, 2004, contained 
clarifying information and did not change the staff's proposed finding 
of no significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 5, 2004.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of application for amendment: October 24, 2003, as 
supplemented by letter dated June 29, 2004.
    Brief description of amendment: The amendment revised the 
Surveillance Requirements (SRs) associated with reactor protection 
system instrumentation. Specifically, the amendment revised the SRs 
associated with the control rod block instrumentation, source range 
monitors, and power distribution limits by removing unnecessary testing 
requirements.
    Date of issuance: October 13, 2004.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 153.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 9, 2003 (68 FR 
68672). The June 29, 2004 letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination or expand the application beyond the scope 
of the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 13, 2004.
    No significant hazards consideration comments received: No.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R. E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: March 1, 2004.
    Brief description of amendment: The amendment extends the 
completion time (CT) from 1 hour to 24 hours for Condition B of 
Technical Specification (TS) 3.5.1, which defines requirements for the 
emergency core cooling system accumulators. Condition B of TS 3.5.1 
specifies a CT to restore an accumulator to operable status when it has 
been declared inoperable for a reason other than the boron 
concentration of the water in the accumulator not being within the 
required range.
    Date of issuance: October 4, 2004.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 86.
    Renewed Facility Operating License No. DPR-18: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: June 22, 2004 (69 FR 
34706).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 4, 2004.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348, Joseph M. 
Farley Nuclear Plant, Unit 1, Houston County, Alabama

    Date of amendments request: September 19, 2003, as supplemented by 
letters dated March 31, June 18, and August 6, 2004.
    Brief Description of amendments: This amendment revised Technical 
Specifications (TS) Limiting Conditions for Operation 3.8.4, ``DC 
Sources--Operating,'' for the remainder of operating cycle 19. 
Specifically, the TS change increased the Completion Time for the 1B 
Auxiliary Building DC electrical power system inoperability due to an 
inoperable battery to allow for on-line replacement of individual 
cells.
    Date of issuance: September 30, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 164.
    Facility Operating License Nos. NPF-2: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2003 (68 
FR 64137).
    The supplements dated March 31, June 18 and August 6, 2004, 
provided clarifying information that did not change the scope of the 
September 19, 2003, application, nor the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 29, 2004.
    No significant hazards consideration comments received: No.

[[Page 62485]]

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendments request: August 29, 2003, as supplemented by 
letters dated November 11, 2003, and May 5, June 10, August 5, August 
25, and September 27, 2004.
    Brief Description of amendments: The amendments revised Technical 
Specifications Limiting Condition of Operation 3.9.3, ``Containment 
Penetrations.'' The changes allow the equipment hatch to be open during 
core alterations and/or during movement of irradiated fuel assemblies 
within containment.
    Date of issuance: September 30, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 165 and 157.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2003 (68 
FR 64137).
    The supplements dated November 11, 2003, and May 5, June 10, August 
5, August 25, and September 27, 2004, provided clarifying information 
that did not change the scope of the August 29, 2003, application nor 
the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 30, 2004.
    No significant hazards consideration comments received: No.
    Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama
    Date of amendments request: August 25, 2004, as supplemented by 
letter dated September 27, 2004.
    Brief Description of amendments: The amendments address the control 
room habitability guidance of Regulatory Guide 1.196, ``Control Room 
Habitability at Light-Water Nuclear Power Reactors,'' by revising 
Limiting Condition for Operation 3.7.10, ``Control Room Emergency 
Filtration/Pressurization System (CREFS)'' and TS 5.5.11, ``Ventilation 
Filter Testing Program. The amendments also add a new section, TS 
5.5.18, ``Control Room Integrity Program (CRIP).''
    Date of issuance: September 30, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 166, 158.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: August 31, 2004 (69 FR 
53095). The supplement dated September 27, 2004, provided clarifying 
information that did not change the scope of the August 25, 2004, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 30, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 8, 2004, as supplemented 
by letter dated September 24, 2004.
    Brief description of amendment: The amendment revises requirements 
in the technical specifications to adopt the provisions of Industry/
Technical Specification Task Force (TSTF) change TSTF-359, ``Increase 
Flexibility in Mode Restraints.'' The availability of TSTF-359 for 
adoption by licensees was announced in the Federal Register on April 4, 
2003 (68 FR 16579).
    Date of issuance: October 8, 2004.
    Effective date: October 8, 2004, and shall be implemented within 90 
days of the date of issuance.
    Amendment No.: 164.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 11, 2004 (69 FR 
26194).
    The additional information provided in the supplemental letter 
dated September 24, 2004, does not expand the scope of the application 
as noticed and does not change the NRC staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 8, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: February 9, 2004, as supplemented by the 
letter dated September 14, 2004.
    Brief description of amendment: The amendment revises requirements 
in the technical specifications to adopt the provisions of Industry/
Technical Specification Task Force (TSTF) change TSTF-359, ``Increase 
Flexibility in Mode Restraints.''
    Date of issuance: October 7, 2004.
    Effective date: October 7, 2004, and shall be implemented within 90 
days of the date of issuance.
    Amendment No.: 155.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 16, 2004 (69 FR 
12373).
    The additional information provided in the supplemental letter 
dated September 14, 2004, does not expand the scope of the application 
as noticed and does not change the NRC staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 7, 2004.

    No significant hazards consideration comments received: No.

    Dated in Rockville, Maryland, this 18th day of October 2004.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 04-23664 Filed 10-25-04; 8:45 am]
BILLING CODE 7590-01-P