[Federal Register Volume 69, Number 119 (Tuesday, June 22, 2004)]
[Notices]
[Pages 34696-34712]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 04-13753]


-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from, May 28, 2004, through June 10, 2004. The 
last biweekly notice was published on June 8, 2004 (69 FR 32070).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be

[[Page 34697]]

affected by this proceeding and who wishes to participate as a party in 
the proceeding must file a written request for a hearing and a petition 
for leave to intervene. Requests for a hearing and a petition for leave 
to intervene shall be filed in accordance with the Commission's ``Rules 
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. 
Interested persons should consult a current copy of 10 CFR 2.309, which 
is available at the Commission's PDR, located at One White Flint North, 
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System's (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://
www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a 
hearing or petition for leave to intervene is filed within 60 days, the 
Commission or a presiding officer designated by the Commission or by 
the Chief Administrative Judge of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) e-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, hearingdocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to 301-415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/
reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to 
pdr@nrc.gov.

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 amendment request: May 5, 2004.
    Description of amendment request: The proposed change will revise 
Technical Specification Surveillance Requirement (SR) 4.0.5.a for 
inservice inspection (ISI) and testing of American Society of 
Mechanical Engineers (ASME) Code Class 1, 2, and 3 components, to 
include a reference to the ASME Code for Operation and Maintenance of 
Nuclear Power Plants (OM Code) in addition to Section XI of the ASME 
Boiler and Pressure Vessel Code and applicable Addenda as required by 
Title 10 of the Code of Federal Regulations (10 CFR), Section 
50.55a(g).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 34698]]

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 changes to the Technical Specification SR 4.0.5.a 
and the associated Bases are requested to add a reference to the 
ASME OM Code and applicable Addenda for inservice inspection of ASME 
Code Class 1, 2, and 3 components.
    The existing Technical [Specification] requires inservice 
inspection of ASME Code Class 1, 2, and 3, components and inservice 
testing of ASME Code Class 1, 2 and 3 pumps and valves as required 
by 10 CFR 50.55a. The purposes of the inservice inspection and 
inservice testing programs are to assess the operational readiness 
of pumps and valves, to detect degradation that might affect 
component operability, and to maintain safety margins with 
provisions for increased surveillance and corrective action. 10 CFR 
50.55a defines the requirements for applying industry codes and 
standards to each licensed nuclear power facility. The initial HNP 
[Shearon Harris Nuclear Power Plant, Unit 1] ISI program was 
developed in accordance with NRC regulations (10 CFR 
50.55a(g)(4)(i)) to comply with the 1983 Edition of the ASME Boiler 
and Pressure Vessel Code, including Addenda through the Summer of 
1983 and is reflected in the existing Technical Specifications and 
associated Bases sections.
    The current, second ten-year interval HNP ISI program was 
developed in accordance with the 1989 Edition (no Addenda) of ASME 
Boiler and Pressure Vessel Code, Section XI. Subarticles IWF-1200 
and IWF-5300 require the examination and testing of snubbers per the 
first Addenda of ASME/ANSI [American National Standards Institute] 
OM-1987, Part 4 (published in 1988), generally referred to as ``OM-
4.'' HNP Relief Request 2RG-008, Revision 1, grants HNP the ability 
to retain the snubber testing and examination program in Technical 
Specification 3/4.7.8.
    The 1995 Edition with 1996 Addenda of the ASME OM Code, 
Subsection ISTD, is the applicable Code per Code Case OMN-13. HNP 
plans to utilize the 1995 Edition with 1996 Addenda of the ASME OM 
Code for snubber visual examinations as an approved alternative to 
the snubber visual examination requirements of the 1989 Edition of 
ASME Section XI and as modified by HNP Relief Request 2RG-008, 
Revision 1. Code Case OMN-13 has been evaluated and approved by the 
NRC in Reg Guide 1.192.
    The proposed change to Technical Specification SR 4.0.5.a is 
also administrative in nature. The proposed changes comply with 
approved codes and standards. As a result, there will be no affect 
on plant safety.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The changes to Technical Specification SR 4.0.5.a and Bases 
section 4.0.5 and are being proposed to reference the ASME OM Code 
in addition to Section XI of the ASME Boiler and Pressure Vessel 
Code. The proposed changes are administrative in nature and do not 
adversely affect accident initiators or precursors nor alter the 
design assumptions, conditions, or configuration of the facility.
    The use of the ASME OM Code 1995 Edition with 1996 Addenda, 
Subsection ISTD, with incorporation of the snubber visual 
examination frequency of Code Case OMN-13 will result in an 
improvement in personnel safety and dose reduction.
    This change will have no operational impact, therefore, the 
proposed change will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The changes to Technical Specification SR 4.0.5.a and Bases 
section 4.0.5 do not involve a reduction in the margin of safety. As 
previously identified, the subject changes are administrative in 
nature and will add a reference to the ASME OM Code in Technical 
Specification SR 4.0.5.a. Therefore, the proposed changes to the 
Technical Specifications and Bases will not result in a reduction in 
the margin of safety.
    Based on the above, HNP concludes that the proposed amendment 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and accordingly, a finding of ``no 
significant hazards consideration'' is justified.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: William Burton (Acting).

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

    Date of amendment request: January 30, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.3.6.2, ``Secondary Containment 
Isolation Instrumentation,'' Condition C, to add the words, ``not 
met,'' to the end of the sentence, ``Required Action and associated 
Completion Time.'' The omission of the words, ``not met,'' was an 
oversight during the change to the Improved Standard Technical 
Specifications (ISTS), NUREG 1433.
    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 proposed change corrects the sentence in Condition C of TS 
3.3.6.2 by indicating that when this condition is not met, certain 
actions are required. This terminology is prevalent throughout the 
ISTS and is implied in this section as well. No changes in operating 
practices or physical plant equipment are created as a result of 
this terminology addition. 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 type of accident from any accident previously evaluated?
    Response: No.
    This proposed change is a correction of an action statement in 
TS 3.3.6.2. No physical change in plant equipment will result from 
this proposed change. Therefore, the proposed change does not create 
the possibility of a new or different type of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change is editorial in nature and only provides a 
correction to an action statement in the Secondary Containment 
Isolation Instrumentation involving inoperable channels and 
automatic functions to agree with NUREG 1433. 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: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: March 19, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.3.6.1, ``Primary Containment 
Isolation Instrumentation,'' to correct a formatting error introduced 
during conversion to Improved Technical Specifications (ITS)

[[Page 34699]]

by replacing ``1 per room'' with ``2'' for the Required Channels Per 
Trip System for the Reactor Water Cleanup (RWCU) Area Ventilation 
Differential Temperature--High primary containment isolation 
instrumentation.
    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 proposed change restores the number of Required Channels Per 
Trip System of the RWCU Area Ventilation Differential Temperature--
High isolation, Function 5.c of Table 3.3.6.1-1 of TS 3.3.6.1, 
Primary Containment Isolation Instrumentation, to its pre-ITS value 
and adds an explanatory note. No changes in operating practices or 
physical plant equipment are created as a result of this change. 
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 type of accident from any accident previously evaluated?
    Response: No.
    The proposed change restores the number of Required Channels Per 
Trip System of the RWCU Area Ventilation Differential Temperature--
High isolation, Function 5.c of Table 3.3.6.1-1 of TS 3.3.6.1, 
Primary Containment Isolation Instrumentation, to its pre-ITS value 
and adds an explanatory note. No physical change in plant equipment 
will result from this proposed change. Therefore, the proposed 
change does not create the possibility of a new or different type of 
accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change is administrative in nature and only 
provides a correction to Table 3.3.6.1-1 of TS 3.3.6.1, Primary 
Containment Isolation Instrumentation, as well as an explanatory 
note. 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: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: May 19, 2004.
    Description of amendment request: The proposed change revises 
Technical Specification (TS) 3.8.1, ``AC Sources--Operating,'' to 
permit a longer completion time for the Division 1 and Division 2 
diesel generators (DGs). This is a risk-informed TS change that would 
extend the DG completion time from 72 hours (the current limit) to 14 
days.
    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 proposed change does not adversely affect the design of the 
DGs, the operational characteristics or function of the DGs, the 
interfaces between the DGs and other plant systems, or the 
reliability of the DGs. Required Actions and the associated 
Completion Times are not initiating conditions for any accident 
previously evaluated, and the DGs are not initiators of any 
previously evaluated accidents.
    The DGs support the mitigation of the consequences of previously 
evaluated accidents that involve a loss of offsite power. The 
consequences of a previously analyzed accident will not be 
significantly affected by the extended DG Completion Time since the 
remaining DGs will continue to be capable of performing their 
accident mitigation function as assumed in the accident analysis. 
Thus, the consequences of accidents previously analyzed are 
unchanged between the existing TS requirements and the proposed 
changes. The consequences of an accident are independent of the time 
the DGs are out of service as long as there are adequate DGs 
available.
    Based on the above, the proposed change to extend the DG allowed 
Completion Time during plant operation will not involve a 
significant increase in accident probabilities or consequences.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    No new accidents would be created since no changes are being 
made to the plant that would introduce any new accident causal 
mechanisms. This amendment request does not impact any plant systems 
that are accident initiators; neither does it adversely impact any 
accident mitigating systems. The addition of an independent AACSBC 
[alternate AC source to the Division 1 and Division 2 battery 
chargers] will provide added time for responding to a loss of all AC 
power assumed in the accident analyses. The design of the AACSBC 
will contain features and administrative controls to maintain the 
separation and protection of emergency AC distribution systems and 
does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    Based on the above, implementation of the proposed changes will 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident. These barriers include the fuel 
cladding, the reactor coolant system, and the containment system. 
Throughout the period of the current TS Completion Time, when one DG 
is out-of-service during power operation, the margin of safety is 
managed by limiting the allowed outage time and other concurrent 
power source outages within the TS. This time period is a temporary 
relaxation of the single failure criteria, which, consistent with 
overall system reliability considerations, provides a limited time 
to repair the equipment and conduct testing. The extension of the 
current TS Completion Time to 14 days has been determined not to be 
a significant reduction in the margin of safety. The proposed 
changes will not result in a significant decrease in DG availability 
so that the assumptions regarding DG availability are not impacted. 
Probabilistic Risk Assessment (PRA) methods, and a deterministic 
analysis were utilized to fully evaluate the effect of the proposed 
DG Completion Time extension. The results of the analysis show no 
significant increase in Core Damage Frequency (CDF) and Large Early 
Release Frequency (LERF). Energy Northwest has proposed a number of 
risk management actions to reduce the possibility of a plant 
transient; a loss of high-pressure injection and cooling systems, a 
loss of other on-site power sources, or a loss of offsite power 
during the period the DG is out-of-service.
    Based on the above, the change to the TS Completion Time does 
not result in a significant reduction in the margin of safety. This 
is based on our management of plant risk, the reliability of the 
other diesel generators, and the inclusion of risk management 
actions.

    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: Stephen Dembek.

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

    Date of amendment request: May 12, 2004.

[[Page 34700]]

    Description of amendment request: The proposed amendment would 
change the reactor core analytical methods used to determine the core 
operating limits, reflect the changes allowed by Technical 
Specification Task Force (TSTF) Traveler No. 363, ``Revised Topical 
Report References in ITS [Improved Standard Technical Specifications] 
5.6.5, COLR [Core Operating Limits Report],'' and delete the Index from 
the Technical Specifications (TSs).
    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.

TS 6.9.5.1, Core Operating Limits Report (COLR)

    The proposed amendment, in part, identifies a change in the 
nuclear physics codes used to confirm the values of selected cycle-
specific reactor physics parameter limits and includes minor 
editorial changes which do not alter the intent of stated 
requirements. The proposed change also allows the use of methods 
required for the implementation of ZIRLO clad fuel rods. Inasmuch as 
the proposed change includes codes that have been previously 
approved by the NRC [Nuclear Regulatory Commission] for CE 
[Combustion Engineering] cores, the amendment is administrative in 
nature and has no impact on any plant configuration or system 
performance relied upon to mitigate the consequences of an accident. 
Parameter limits specified in the COLR for this amendment are not 
changed from the values presently required by TSs. Future changes to 
the calculated values of such limits may only be made using NRC 
approved methodologies, must be consistent with all applicable 
safety analysis limits, and are controlled by the 10 CFR 50.59 
process. Assumptions used for accident initiators and/or safety 
analysis acceptance criteria are not altered by this change.
    The proposed change also implements NRC approved TSTF Traveler 
No. 363. This is an administrative change that will allow specific 
details, such as the revision number, revision date, and supplement 
number of topical reports that are referenced in the TSs, to be 
deleted and relocated in the cycle specific COLR. This proposed 
change does not result in any changes to the assumptions used to 
evaluated accident initiators and/or safety analysis acceptance 
criteria.

Index

    The proposed deletion of the Index is purely administrative and 
does not impact the accident analysis.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.

TS 6.9.5.1, Core Operating Limits Report (COLR)

    The proposed change, in part, identifies a change in the nuclear 
physics codes used to confirm the values of selected cycle-specific 
reactor physics parameter limits. The proposed change also allows 
the use of methods required for the implementation of ZIRLO clad 
fuel rods. Neither of these changes results in a change to the 
physical plant or to the modes of operation defined in the facility 
license.
    The proposed change also implements TSTF Traveler No. 363. The 
proposed change does not result in changes to the physical plant or 
to the modes of operation defined in the facility license nor does 
it involve the addition of new equipment or the modification of 
existing equipment.

Index

    The proposed deletion of the Index is purely administrative has 
no affect on existing equipment.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.

TS 6.9.5.1, Core Operating Limits Report (COLR)

    The proposed changes to change the nuclear physics code package 
and to add a topical report to support the use of ZIRLO do not amend 
the cycle specific parameter limits located in the COLR from the 
values presently required by the TS. The individual specifications 
continue to require operation of the plant within the bounds of the 
limits specified in COLR. Benchmarking has shown that uncertainties 
for the Westinghouse Physics code system yields are essentially the 
same or less than those obtained for the current ROCS/DIT 
methodology. Future changes to the values of these limits by the 
licensee may only be developed using NRC approved methodologies, 
must remain consistent with all applicable plant safety analysis 
limits addressed in the Safety Analysis Report, and are further 
controlled by the 10 CFR 50.59 process. The relocation of the 
supplement numbers, revision numbers, and approval dates of the 
analytical methods listed in the COLR does not affect the margin of 
safety. The analysis will continue to be performed using NRC 
approved methodology. Safety analysis acceptance criteria are not 
being altered by this amendment.

Index

    The proposed deletion of the Index, which is an administrative 
document, does not impact any TS values or safety limits.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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: February 27, 2004.
    Description of amendment request: This amendment request 
incorporates a revision to the Technical Specifications and licensing 
and design bases that supports a full-scope application of an 
Alternative Source Term (AST) methodology.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Adoption of the AST and those plant systems affected by 
implementation of the AST do not initiate design-basis accidents 
(DBAs). The proposed changes do not affect the design or manner in 
which the facility is operated; rather, once the occurrence of an 
accident has been postulated, the new AST is an input to analyses 
that evaluate the radiological consequences. Therefore, the proposed 
changes do not involve an increase in the probability of an accident 
previously evaluated.
    The structures, systems and components (SSCs) affected by the 
proposed change act as mitigators to the consequences of accidents. 
Based on the revised analyses, the proposed changes do revise 
certain performance requirements; however, the proposed changes 
involve different acceptance criteria. There cannot, therefore, be a 
direct comparison to determine if the proposed change would result 
in an increase in consequences over the current design. However, the 
licensee's analysis proposes that, with implementation of AST, all 
regulatory acceptance criteria continue to be met. Therefore, any 
potential increase in consequences would not be considered 
significant.

[[Page 34701]]

    Therefore, the proposed changes do not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Implementation of AST does not affect the design function or 
mode of operations of SSCs in the facility prior to a postulated 
accident. Since SSCs are operated essentially the same after the AST 
implementation, no new failure modes are created by this proposed 
change.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The changes proposed are associated with a revision to the 
licensing basis. These changes would modify the input to DBA 
analyses from the original source term to the AST. Based on the 
revised analyses, the proposed changes involve different acceptance 
criteria. There cannot, therefore, be a direct comparison to 
determine if the proposed change would result in a reduction in a 
margin of safety. However, the licensee's analysis proposes that, 
with implementation of AST, all regulatory acceptance criteria 
continue to be met. The dose consequences of the accident analyses 
revised in support of the proposed changes are subject to the 
acceptance criteria in 10 CFR 50.67, ``Accident source term,'' 
Regulatory Guide 1.183, ``Alternative Radiological Source Terms for 
Evaluating Design Basis Accidents at Nuclear Power Reactors,'' and 
Standard Review Plan 15.0.1, ``Radiological Consequence Analyses 
Using Alternative Source Terms.'' Thus, by meeting the applicable 
regulatory limits for AST, any potential decrease in a margin of 
safety would not be considered significant.
    Therefore, because the proposed changes continue to result in 
dose consequences within the applicable regulatory limits, the 
changes are considered to not result in a significant reduction in a 
margin of safety.

    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the 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: James W. Clifford.

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: April 8, 2004.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS), Section 6, Administrative 
Controls, to relocate (1) the Plant Operations Review Committee and 
Nuclear Review Board requirements, (2) the program/procedure review and 
approval requirements, and (3) the record retention requirements to the 
Quality Assurance Topical Report, the document controlling the 
licensee's quality assurance program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No. The proposed changes involve the relocation of 
several administrative requirements from the Technical 
Specifications (TS) to a document subject to the control of 10 CFR 
50.54(a), and is therefore, administrative in nature. The relocated 
requirements involve the onsite and offsite organization's review 
and audit, the review and approval of procedures, and the retention 
of records. The change will not alter the physical design or 
operational procedures associated with any plant structure, system, 
or component. The change does not reduce the duties and 
responsibilities of the organizations performing the review, audit, 
and approval functions essential to ensuring the safe operation of 
the plant.
    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 are administrative in nature. 
The changes do not alter the physical design, safety limits, or 
safety analysis assumptions, associated with the operation of the 
plant. Accordingly, the changes do not introduce any new accident 
initiators, nor do they reduce or adversely affect the capabilities 
of any plant structure, system, or component to perform 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. The proposed changes conform to NRC regulatory 
guidance regarding the content of plant Technical Specifications. 
The guidance is presented in Administrative Letter 95-06, and NUREG-
1433, Rev. 2. The relocation of these administrative requirements 
will not reduce the quality assurance commitments as accepted by the 
NRC, nor reduce administrative controls essential to the safe 
operation of the plant. Future changes to these administrative 
requirements will be performed in accordance with NRC regulation 10 
CFR 50.54(a), consistent with the guidance identified above. 
Accordingly, the relocation results in an equivalent level of 
regulatory control.
    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: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: James W. Clifford.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: March 17, 2004.
    Description of amendment request: The proposed amendment would 
revise the operating license and Technical Specifications (TSs) to 
support an increase in the licensed power from 3411 megawatts thermal 
(MWt) to 3587 MWt. This represents an increase of approximately 5.2 
percent above the current rated licensed thermal power.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Plant structures, systems and components (SSCs) have been 
verified to be capable of performing their intended design functions 
at uprated power conditions. Where necessary, some components will 
be modified prior to implementation of uprated power operations to 
accommodate the revised operating conditions. The analysis indicated 
that operation at uprated power conditions will not adversely affect 
the capability of plant equipment. Current TS surveillance 
requirements ensure frequent and adequate monitoring of system and 
component operability. All systems will continue to be operated in 
accordance with current design requirements under uprated 
conditions; therefore, no new components or system interactions have 
been identified that could lead to an increase in the probability of 
any accident previously evaluated in the Updated Final Safety 
Analysis Report (UFSAR).
    The radiological consequences were reviewed for design basis 
accidents (DBAs) previously analyzed in the UFSAR. The analysis 
showed that the resultant radiological consequences for both loss-
of-coolant accidents (LOCAs) and non-LOCAs remain either unchanged 
or have increased due to operation at uprated power conditions. Any 
increase in the radiological

[[Page 34702]]

consequences of DBAs is not considered significant because plant 
operation at uprated power conditions continue to meet established 
regulatory limits.
    Therefore, the proposed changes do not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The configuration, operation, and accident response of the SSCs 
are unchanged by operation at uprated power conditions or by the 
associated proposed TS changes. Analyses of transient events have 
confirmed that no transient event results in a new sequence of 
events that could lead to a new accident scenario.
    The effect of operation at uprated power conditions on plant 
equipment has been evaluated. No new operating mode, safety-related 
equipment lineup, accident scenario, or equipment failure mode was 
identified as a result of operating at uprated conditions. In 
addition, operation at uprated power conditions does not create any 
new failure modes that could lead to a different kind of accident. 
Minor plant modifications, to support implementation of uprated 
power conditions, will be made as required to existing systems and 
components. The basic design function of all SSCs remains unchanged 
and no new safety-related equipment or systems will be installed 
which could potentially introduce new failure modes or accident 
sequences.
    Based on this analysis, it is concluded that no new accident 
scenarios, failure mechanisms or limiting single failures are 
introduced as a result of the proposed changes. The proposed TS 
changes do not have an adverse effect on any safety. Therefore, the 
proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    A comprehensive analysis was performed to support the power 
uprate program at the Seabrook Station. This analysis identified and 
defined the major input parameters to the Nuclear Steam Supply 
System (NSSS), reviewed NSSS design transients, and reviewed the 
capabilities of the NSSS fluid systems, NSSS/BOP (balance-of-plant) 
interfaces, and NSSS and BOP components. The nuclear and thermal 
hydraulic performance of nuclear fuel was also reviewed to confirm 
acceptable results. Only minor plant modifications, to support 
implementation of uprated power conditions, will be made as required 
to existing systems and components. Changes in setpoints for 
actuation of equipment do not adversely affect the outcome of any 
postulated accident. The analysis indicated that all NSSS and BOP 
systems and components will continue to operate within existing 
design and safety limits at uprated power conditions.
    The margin of safety of the reactor coolant pressure boundary is 
maintained under uprated power conditions. The design pressure of 
the reactor pressure vessel and reactor coolant system will not be 
challenged as the pressure mitigating systems were confirmed to be 
sufficiently sized to adequately control pressure under uprated 
power conditions.
    The radiological consequences were reviewed for DBAs previously 
analyzed in the UFSAR. The analysis showed that the radiological 
consequences of DBAs continue to meet established regulatory limits 
at uprated power conditions.
    The analyses supporting the power uprate program have 
demonstrated that all systems and components are capable of safely 
operating at uprated power conditions. All DBA acceptance criteria 
will continue to be met. Therefore, it is concluded that the 
proposed changes do not result in 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: M.S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Section Chief: James W. Clifford.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: May 27, 2004.
    Description of amendment request: The proposed amendment would 
revise the Cooper Nuclear Station (CNS) Technical Specifications (TS). 
The proposed amendment would lower the reactor vessel water level at 
which the reactor water cleanup (RWCU) system isolates, secondary 
containment isolates, and the control room emergency filter system 
(CREFS) starts. General Electric (GE) Service Information Letter (SIL) 
No. 131 discussed problems that result from isolation of the RWCU and 
start of the standby gas treatment (SGT) system, in conjunction with 
isolation of secondary containment. The SIL recommended that the vessel 
water level at which these actions occur be lowered, thereby 
eliminating these problems and the resulting unnecessary complications 
with scram recovery. The proposed changes to the CNS TS are in 
accordance with SIL 131 Recommendations 1 and 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The values of various plant parameters at which piping 
connected to the reactor vessel and containment isolates and air-
filtering systems start are not accident precursors. Thus, lowering 
the reactor vessel water level at which RWCU and secondary 
containment isolate and SGT and CREFS initiate has no impact on the 
probability of a design basis accident evaluated in the CNS Station 
Safety Analysis. Therefore the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The proposed logic changes involve no changes to the logic of 
the Reactor Protection System that initiates automatic reactor 
shutdown in response to an accident. The proposed logic changes 
involve no changes to the logic of the Emergency Core Cooling System 
(ECCS) that initiates automatic actions to ensure adequate core 
cooling and containment integrity in response to an accident. The 
CNS response to the design basis accidents (DBAs) addressed in the 
Station Safety Analysis with the proposed changes to the logic was 
evaluated. This evaluation has demonstrated that there is no 
increase in the offsite radiological doses to the public resulting 
from these accidents.
    Based on the above NPPD [Nebraska Public Power District] 
concludes that the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed lowering of the level of water in the reactor 
vessel at which certain automatic actions would occur changes the 
operation of various systems at CNS. However, the change in system 
operation is not significant. Currently automatic actions occur in 
the RWCU System, SGT System, CREFS, and secondary containment in 
response to reactor vessel water level. Changing the level at which 
these automatic actions occur is not a significant change in the 
systems operation. Hardware changes needed to implement the modified 
logic are minor. Lowering the reactor vessel water level for these 
actions does not introduce a new mode of plant operation and does 
not create a potential for any new failure mechanisms, malfunctions, 
or accident initiators. Making this change does not involve adding 
new systems to the CNS design.
    Based on the above NPPD concludes that the proposed changes do 
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?
    The safety margin associated with dose consequences to the 
public following DBAs is based on, in part, automatic operation of 
systems that shut down the reactor, automatic initiation of ECCS, 
and automatic isolation of primary and secondary containment. The 
proposed changes to the CNS TS make no changes that affect the 
automatic shutdown of the reactor or the

[[Page 34703]]

automatic initiation and operation of ECCS. The plant response to 
DBAs with the proposed revisions to the RWCU isolation (primary 
containment) and the SGT and the CREFS initiation (secondary 
containment) have been evaluated and shown to not result in any 
increase in dose to the public. The safety margin associated with 
dose consequences to the control room operators is based on 
automatic isolation of secondary containment, and initiation of 
CREFS. The plant response to DBAs with the proposed revisions to the 
RWCU isolation (primary containment) and SGT and CREFS initiation 
(secondary containment) have been evaluated and shown to not result 
in any increase in dose to the control room operators.
    Based on the above NPPD concludes that 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. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: May 14, 2004.
    Description of amendment request: The proposed amendment will 
relocate the requirements of Technical Specification (TS) 3.3(1)a, 
``Reactor Coolant System and Other Components Subject to ASME XI Boiler 
& Pressure Vessel Code Inspection and Testing Surveillance,'' 
concerning inservice inspection of ASME Class 1, 2, and 3 components 
and TS 3.4, ``Reactor Coolant System Integrity Testing,'' concerning 
reactor coolant system integrity testing to the Fort Calhoun Station 
(FCS) Updated Safety Analysis Report (USAR). These TSs do not meet the 
criterion in 10 CFR 50.36(c)(2)(ii) for inclusion in the TSs.
    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 amendment relocates the requirements of TS 3.3(1)a 
concerning inservice inspection of ASME Class 1, 2, and 3 components 
and TS 3.4 concerning reactor coolant system integrity testing to 
the FCS USAR. These TSs are directed toward prevention of component 
degradation and continued long term maintenance of acceptable 
structural conditions. It is not necessary to retain these TSs to 
ensure immediate operability of safety systems. Therefore these TSs 
do not meet the criteria set forth in 10 CFR 50.36(c)(2)(ii) for 
inclusion in the TS. The requirements are being relocated from TS to 
the FCS USAR, which will be maintained pursuant to 10 CFR 50.59, 
thereby reducing the level of regulatory control. [This reduction in 
the] level of regulatory control has no impact on the probability or 
consequences of an accident previously evaluated. Therefore, the 
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 relocates requirements of TS 3.3(1)a 
concerning inservice inspection of ASME Class 1, 2, and 3 components 
and TS 3.4 concerning reactor coolant system integrity testing that 
do not meet the criteria for inclusion in TS set forth in 10 CFR 
50.36(c)(2)(ii). The change does not involve a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or make changes in the methods governing normal plant 
operation. The change will not impose different requirements, and 
adequate control of information will be maintained. This change will 
not alter assumptions made in the safety analysis and licensing 
basis. Therefore, the change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change relocates requirements of TS 3.3(1)a 
concerning inservice inspection of ASME Class 1, 2, and 3 components 
and TS 3.4 concerning reactor coolant system integrity testing that 
do not meet the criteria for inclusion in TS set forth in 10 CFR 
50.36(c)(2)(ii). The change will not reduce a margin of safety since 
the location of a requirement has no impact on any safety analysis 
assumptions. In addition, the relocated requirements of TS 3.3(1)a 
and TS 3.4 concerning inservice inspection and testing of ASME Class 
1, 2, and 3 components remain the same as the existing TS. Since any 
future changes to these requirements will be evaluated per the 
requirements of 10 CFR 50.59, there will be no 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 R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: May 21, 2004.
    Description of amendment request: The proposed amendment would add 
information to the Technical Specification (TS) Basis for TS 2.4, 
``Containment Cooling,'' to allow containment spray pumps to be secured 
during a loss-of-coolant accident (LOCA) to minimize the potential for 
containment sump clogging when certain conditions are met. NRC Bulletin 
2003-01, ``Potential Impact of Debris Blockage on Emergency Sump 
Recirculation at Pressurized Water Reactors,'' required that operators 
of pressurized water reactor (PWR) plants state that the emergency core 
cooling systems (ECCS) and the containment spray (CS) recirculation 
functions meet applicable regulatory requirements with respect to 
adverse post-accident debris blockage or describe interim compensatory 
measures to reduce the risk associated with the potentially degraded or 
non-conforming ECCS and CS recirculation functions.
    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 changes will not [significantly] increase the 
probability or consequences of any accident based on the following:
    The proposed compensatory action is only taken following a LOCA 
if all safeguards have functioned and if an excess of CS flow exists 
above that required to control containment pressure, temperature, 
and remove the accident source term. The proposed action is only 
taken if the worst-case single failure has not occurred indicating 
maximum containment cooling and SI [safety injection] flow 
delivered, and minimum source term due to no severe core damage. The 
proposed action occurs following the peak containment pressure 
transient, therefore, the action has no impact on the peak 
containment pressure analysis. A quantitative analysis of the change 
in LOCA consequences due to suspension of CS flow for 10 minutes has 
not been performed. However, the prerequisite conditions for taking 
this action provide reasonable assurance that the loss of the 
remaining CS train for ten minutes will not result in a significant 
increase in the LOCA consequences. Therefore, the proposed changes 
will not [significantly] increase the probability or consequence of 
any accident.

[[Page 34704]]

    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed revision does not involve physical changes to any 
equipment required to mitigate the consequences of an accident, nor 
alter how design basis accident events are postulated. The proposed 
change alters the method of controlling an Engineered Safety Feature 
following a design basis event so that manual actions are 
substituted for automatic actions. Reasonable assurance exists that 
these manual actions can be taken in a timely manner to allow 
continued CS system operation to provide containment cooling and 
source term reduction with no significant increases in the 
radiological consequences or approaching of design containment 
limits. 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.
    The proposed change alters the method of controlling an 
Engineered Safety Feature following a design basis event so that 
manual actions are substituted for automatic actions. The proposed 
actions are only taken following a LOCA if all safeguards have 
functioned and if an excess of CS flow exists above that required to 
control containment pressure, temperature, and remove the accident 
source term. The prerequisite conditions for taking this action 
provide reasonable assurance that the loss of the remaining CS train 
will not result in a reduction in the margin of safety for 
radiological consequences or containment design parameters. 
Therefore, the proposed changes do not involve a significant 
reduction to 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: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment requests: March 18, 2004.
    Description of amendment requests: The proposed amendments would 
authorize updates of the Diablo Canyon Power Plant (DCPP) Final Safety 
Analysis Report (FSAR) Update to use on a permanent basis, a revised 
steam generator (SG) voltage-based repair criteria probability of 
detection (POD) method using plant specific SG tube inspection results, 
referred to as the probability of prior cycle detection (POPCD) method.
    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 use of a revised steam generator (SG) voltage-based repair 
criteria probability of detection (POD) method, the probability of 
prior cycle detection (POPCD) method, to determine the beginning of 
cycle (BOC) indication voltage distribution for the Diablo Canyon 
Power Plant (DCPP) Units 1 and 2 operational assessments does not 
increase the probability of an accident. Based on industry and plant 
specific bobbin detection data for outside diameter stress corrosion 
cracks (ODSCC) within the SG tube support plate (TSP) region, large 
voltage bobbin indications which individually can challenge 
structural or leakage integrity can be detected with near 100 
percent certainty. Since large voltage ODSCC bobbin indications 
within the SG TSP can be detected, they will not be left in service, 
and therefore these indications should not be included in the 
voltage distribution for the purpose of operational assessments. The 
POPCD method improves the estimate of potentially undetected 
indications for operational assessments, but does not directly 
affect the inspection results. Since large voltage indications are 
detected, they will not result in an increase in the probability of 
a steam generator tube rupture (SGTR) accident or an increase in the 
consequences of a SGTR or main steam line break (MSLB) 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 use of the POPCD method to determine the BOC voltage 
distribution for the DCPP Units 1 and 2 operational assessments 
concerns the SG tubes and can only affect numerical predictions of 
probabilities for the SGTR accident. Since the SGTR accident is 
already considered in the Final Safety Analysis Report Update, there 
[is] no possibility to create a design basis accident that has not 
been previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The use of the POPCD method to determine the BOC voltage 
distribution for the DCPP Units 1 and 2 operational assessments does 
not involve a significant reduction in a margin of safety. The 
applicable margin of safety potentially impacted is the Technical 
Specification 5.6.10, ``Steam Generator (SG) Tube Inspection 
Report,'' projected end-of-cycle leakage for a MSLB [main steam line 
break] accident and the projected end-of-cycle probability of burst. 
Based on industry and plant specific bobbin detection data for ODSCC 
within the SG TSP region, large voltage bobbin indications that can 
individually challenge structural or leakage integrity can be 
detected with near 100 percent certainty and will not be left in 
service. Therefore these indications should not be included in the 
voltage distribution for the purpose of operational assessments. 
Since these large voltage indications are detected, they will not 
result in a significant increase in the actual end-of-cycle leakage 
for a MSLB accident or the actual end-of-cycle probability of burst. 
The POPCD method approach to POD considers the potential for missing 
indications that might challenge structural or leakage integrity by 
applying the POPCD data from successive inspections. If a large 
indication was missed in one inspection, it would continue to grow 
until finally detected in a later inspection.

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

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

    Date of amendment request: April 27, 2004.
    Description of amendment request: The proposed change will revise 
the Safety Limit Minimum Critical Power Ratio (SLMCPR) values for two 
recirculation loop and one recirculation loop operation. Each safety 
limit value will be applicable for all fuel types in the Hope Creek 
Generating Station core. In the amendment request, PSEG Nuclear LLC 
requested changes to the Technical Specifications to support the use of 
GE14 fuel and General Electric Company (GE) reload analysis methods 
beginning with the upcoming Cycle 13.
    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?

[[Page 34705]]

    Response: No.
    The SLMCPR ensures that no mechanistic fuel damage occurs in the 
core if the limit is not violated. The revised SLMCPR values 
maintain the appropriate conservative margin to boiling transition 
and the probability of fuel damage is not increased. The derivation 
of the revised SLMCPR values specified in the Technical 
Specifications has been performed using NRC approved methods and 
uncertainties. The analysis methodology incorporates appropriate 
cycle-specific parameters and uncertainties in determining the 
revised SLMCPR values. The analyses do not change the method of 
operating the plant and have no effect on the probability of an 
accident initiating event or transient. The revised SLMCPR values do 
not affect the performance of systems or components used to mitigate 
the consequences of accidents previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or radiological consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The revised SLMCPR values specified in the Technical 
Specifications have been calculated in accordance with NRC approved 
methods and uncertainties. The changes do not involve any new method 
for operating the facility and do not involve any facility 
modifications. No new initiating events or anticipated operational 
occurrences result from these changes.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The revised SLMCPR values are calculated using NRC approved 
methods and uncertainties. The revised SLMCPR values continue to 
ensure that greater than 99.9% of all fuel rods in the core are 
expected to avoid boiling transition if the safety limits are not 
violated, thereby maintaining the fuel cladding integrity during 
normal plant operation and anticipated operational occurrences.
    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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of amendment request: April 26, 2004.
    Description of amendment request: The proposed change will revise 
the Salem Unit Nos. 1 and 2 source term used for design basis 
radiological analysis, in accordance with the provisions of 10 CFR 
50.67, ``Accident Source Term''. The proposed change will also revise 
certain requirements in the Technical Specifications (TSs) and the 
Updated Final Safety Analysis Report (UFSAR) based on the radiological 
dose analysis margins obtained in the Alternate Source Term 
application.
    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 three standards of 10 CFR 50.92(c). The NRC staff's 
analysis is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The alternative source term analysis does not change the design 
of the plant or affect the performance of the systems or components 
used to mitigate the consequences of accidents previously evaluated. 
The analyses do not change the method of operating the plant and has 
no effect on the probability of an accident initiating event or a 
transient. The alternative source term calculations demonstrate the 
radiological consequences to the design basis accidents specified in 
the plant's UFSAR will still remain well below the radiological 
limits specified in 10 CFR 100.11. Therefore, since the radiological 
consequences are well below the specified 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?
    The proposed amendment is not the result of a hardware design 
change, nor does it lead to the need for a hardware design change. 
There is no change in the methods or procedures by which the unit is 
operated. As a result, all structures, systems, and components will 
continue to perform as previously analyzed by the licensee, and 
previously evaluated and accepted by the NRC staff. Therefore, the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed changes result in operation in accordance with 
regulatory guidelines and support the revisions to the radiological 
analysis of the limiting design basis accidents. The radiological 
consequences of these accidents are all within the regulatory 
acceptance criteria associated with the use of the alternative 
source term methodology. Therefore, the proposed changes do not 
involve a significant reduction in a margin of safety.

    Based on the NRC staff's analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

Rochester Gas and Electric Corporation, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: March 1, 2004.
    Description of amendment request: The proposed amendment would 
extend the completion time (CT) from 1 hour to 24 hours for Condition B 
of Technical Specification (TS) 3.5.1, ``Accumulators.'' The 
accumulators are part of the emergency core cooling system and consist 
of tanks partially filled with borated water and pressurized with 
nitrogen gas. The contents of the tank are discharged to the reactor 
coolant system (RCS) if, as during a loss-of-coolant accident, the 
coolant pressure decreases to below the accumulator pressure. 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. This change was proposed by the Westinghouse Owners 
Group participants in the TS Task Force (TSTF) and is designated TSTF-
370. TSTF-370 is supported by NRC-approved Topical Report WCAP-15049-A, 
``Risk-Informed Evaluation of an Extension to Accumulator Completion 
Times,'' submitted on May 18, 1999. The NRC staff issued a notice of 
opportunity for comment in the Federal Register on July 15, 2002 (67 FR 
46542), on possible amendments concerning TSTF-370, 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

[[Page 34706]]

referencing in license amendment applications in the Federal Register 
on March 12, 2003 (68 FR 11880). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated March 1, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The basis for the accumulator limiting condition for operation 
(LCO), as discussed in Bases Section 3.5.1, is to ensure that a 
sufficient volume of borated water will be immediately forced into 
the core through each of the cold legs in the event the RCS pressure 
falls below the pressure of the accumulators, thereby providing the 
initial cooling mechanism during large RCS pipe ruptures. As 
described in Section 9.2 of the WCAP-15049, ``Risk-Informed 
Evaluation of an Extension to Accumulator Completion Times,'' 
evaluation, the proposed change will allow plant operation with an 
inoperable accumulator for up to 24 hours, instead of 1 hour, before 
being required to begin shutdown. The impact of the increase in the 
accumulator CT on core damage frequency for all the cases evaluated 
in WCAP-15049 is within the acceptance limit of 1.0E-06/yr for a 
total plant core damage frequency (CDF) less than 1.0E-03/yr. The 
incremental conditional core damage probabilities calculated in 
WCAP-15049 for the accumulator CT increase meet the criterion of 5E-
07 in Regulatory Guides (RG) 1.174 [``An Approach for Using 
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis''] and 1.177 [``An Approach 
for Plant-Specific, Risk-Informed Decisionmaking: Technical 
Specifications''] for all cases except those that are based on 
design basis success criteria. As indicated in WCAP-15049, design 
basis accumulator success criteria are not considered necessary to 
mitigate large break loss-of-coolant accident (LOCA) events, and 
were only included in the WCAP-15049 evaluation as a worst case data 
point. In addition, WCAP-15049 states that the NRC has indicated 
that an incremental conditional core damage frequency (ICCDP) 
greater than 5E-07 does not necessarily mean the change is 
unacceptable.
    The proposed technical specification change does not involve any 
hardware changes nor does it affect the probability of any event 
initiators. There will be no change to normal plant operating 
parameters, engineered safety feature (ESF) actuation setpoints, 
accident mitigation capabilities, accident analysis assumptions or 
inputs.
    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

    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed change. As described in Section 9.1 of the WCAP-
15049 evaluation, the plant design will not be changed with this 
proposed technical specification CT increase. All safety systems 
still function in the same manner and there is no additional 
reliance on additional systems or procedures. The proposed 
accumulator CT increase has a very small impact on core damage 
frequency. The WCAP-15049 evaluation demonstrates that the small 
increase in risk due to increasing the accumulator allowed outage 
time (AOT) is within the acceptance criteria provided in RGs 1.174 
and 1.177. No new accidents or transients can be introduced with the 
requested change and the likelihood of an accident or transient is 
not impacted.
    The malfunction of safety related equipment, assumed to be 
operable in the accident analyses, would not be caused as a result 
of the proposed technical specification change. No new failure mode 
has been created and no new equipment performance burdens are 
imposed.
    Therefore, this 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 the Margin of Safety

    The proposed change does not involve a significant reduction in 
a margin of safety. There will be no change to the departure from 
nucleate boiling ratio (DNBR) correlation limit, the design DNBR 
limits, or the safety analysis DNBR limits.
    The basis for the accumulator LCO, as discussed in Bases Section 
3.5.1, is to ensure that a sufficient volume of borated water will 
be immediately forced into the core through each of the cold legs in 
the event the RCS pressure falls below the pressure of the 
accumulators, thereby providing the initial cooling mechanism during 
large RCS pipe ruptures. As described in Section 9.2 of the WCAP-
15049 evaluation, the proposed change will allow plant operation 
with an inoperable accumulator for up to 24 hours, instead of 1 
hour, before being required to begin shutdown. The impact of this on 
plant risk was evaluated and found to be very small. That is, 
increasing the time the accumulators will be unavailable to respond 
to a large LOCA event, assuming accumulators are needed to mitigate 
the design basis event, has a very small impact on plant risk. Since 
the frequency of a design basis large LOCA (a large LOCA with loss 
of offsite power) would be significantly lower than the large LOCA 
frequency of the WCAP-15049 evaluation, the impact of increasing the 
accumulator CT from 1 hour to 24 hours on plant risk due to a design 
basis large LOCA would be significantly less than the plant risk 
increase presented in the WCAP-15049 evaluation.

    Therefore, this change does not involve a significant reduction in 
a margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.
    Attorney for licensee: Daniel F. Stenger, Ballard Spahr Andrews & 
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 
20005.
    NRC Section Chief: Richard J. Laufer.

Rochester Gas and Electric Corporation, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: March 1, 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 TSs 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 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 March 1, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:


[[Page 34707]]



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 proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Daniel F. Stenger, Ballard Spahr Andrews & 
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 
20005.
    NRC Section Chief: Richard J. Laufer.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of amendment request: April 28, 2004.
    Description of amendment request: The proposed amendments would 
relocate requirements related to the Cold Over Pressure Protection 
System (COPS) arming temperature from the Technical Specifications 
(TSs) to the Pressure and Temperature Limits Report (PTLR) to 
facilitate future licensee-controlled changes to the COPS arming 
temperature. The licensee also proposed to change the COPS arming 
temperature from 350 [deg]F to 220 [deg]F.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed changes to the Technical Specifications do not 
affect any plant equipment, test methods, or plant operation, and 
are not initiators of any analyzed accident sequence. COPS will 
continue to perform its function as designed to provide cold over 
pressure protection, and the pressurizer safety valves will provide 
over pressure protection during operation when COPS is not in 
service. Operation in accordance with the proposed TS will ensure 
that all analyzed accidents will continue to be mitigated by the 
Structures, Systems, and Components (SSCs) as previously analyzed. 
Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    No. The proposed changes do not introduce any new equipment, 
create new failure modes for existing equipment, or create any new 
limiting single failures. COPS will continue to ensure that 
appropriate fracture toughness margins are maintained to protect 
against reactor vessel failure during low temperature operation. The 
proposed changes are consistent with [technical specification task 
force] TSTF-233, Revision 0, which was approved by the NRC. Plant 
operation will not be altered, and all safety functions will 
continue to perform as previously assumed in accident analyses. 
Therefore, the proposed changes do 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?
    No. The proposed changes will not adversely affect the operation 
of plant equipment or the function of any equipment assumed in the 
accident analysis. The COPS arming temperature has been established 
in accordance with an NRC-approved methodology. No changes are being 
made to the cold Over pressure protection analysis and the function 
of COPS as assumed in the analysis. Therefore, the proposed changes 
do not involve a significant reduction in any margin to 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. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: Stephanie M. Coffin, Acting Section Chief.

Yankee Atomic Electric Co., Docket No. 50-29, Yankee Nuclear Power 
Station (YNPS) Franklin County, Massachusetts

    Date of amendment request: November 24, 2003, and supplemented 
December 10, 2003, December 16, 2003, January 19, 2004, January 20, 
2004, February 2, 2004, February 10, 2004, and March 4, 2004.
    Description of amendment request: The licensee has proposed to 
amend its license to incorporate a new license condition addressing the 
license termination plan (LTP). The new license condition would 
document the date of NRC approval of the LTP and provide criteria to 
determine the need for NRC approval of changes to the approved LTP.
    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:


[[Page 34708]]


    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Currently, the bounding airborne radioactivity event given in 
the YNPS [Yankee Nuclear Power Station] FSAR [Final Safety Analysis 
Report] is the materials handling event (FSAR Section 403.5). This 
event considered the non-mechanistic release of the contents of the 
dominant plant component that could have caused the highest offsite 
dose as a result of the release of airborne radioactivity during 
handling. The dominant component was the feed and bleed heat 
exchanger which has since been removed from the site. The bounding 
analysis resulted in an offsite dose at the Exclusion Area Boundary 
of about 0.320 rem, significantly less than the EPA Protective 
Action Guidelines. Other airborne particulate radwaste or 
radioactive materials accidents considered in the FSAR but bounded 
by the materials handling event are as follows:
     Fire in a sea-land container containing combustible 
radioactive material,
     Dismantlement activities (i.e., cutting , segmentation) 
during decommissioning,
     A gas bottle explosion inside containment,
     An explosion of a propane tank stored onsite.
    All spent fuel is located at the ISFSI [Independent Spent Fuel 
Storage Installation] and is stored within fifteen NAC Multi-Purpose 
Canisters and associated vertical concrete casks. A sixteenth cask 
contains Greater Than Class C material. The NAC-MPC FSAR addresses 
the various off-normal and accident events which were postulated in 
support of the licensing and certification of the system. In each 
case, there were no radiological consequences as a result of a 
postulated event.
    The requested license amendment is consistent with plant 
activities described in the PSDAR [Post Shutdown Decommissioning 
Activities Report] and the YNPS FSAR. Accordingly, no systems, 
structures, or components that could initiate the previously 
evaluated accident or are required to mitigate these accidents are 
adversely affected by this proposed change. Therefore, the proposed 
change does not involve an increase in the probability or 
consequences of any previously evaluated accident.
    2. The proposed change does not create the possibility of a new 
or different accident from any previously evaluated.
    Accident analyses related to decommissioning activities are 
addressed in the FSAR. The requested license amendment is consistent 
with the plant activities described in the YNPS FSAR and the PSDAR. 
The proposed change does not affect plant systems, structures, or 
components in a way not previously evaluated. The changes do not 
affect any of the parameters or condition that could contribute to 
the initiation of an accident. No new accident scenarios are created 
nor are any new failure mechanisms created by this activity. 
Therefore, the proposed activity does not create the possibility of 
a new or different kind of accident than those previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The LTP [License Termination Plan] is a plan for demonstrating 
compliance with the radiological criteria for license termination as 
provided in 10 CFR 20.1402. The margin of safety defined in the 
statements of consideration for the final rule on the Radiological 
Criteria for License Termination is described as the margin between 
the 100 mrem/yr public dose limit established in 10 CFR 20.1301 for 
licensed operation and the 25 mrem/yr dose limit to the average 
member of the critical group at a site considered acceptable for 
unrestricted use (one of the criteria of 10 CFR 20.1402). This 
margin of safety accounts for the potential effect of multiple 
sources of radiation exposure to the critical group. Since the 
License Termination Plan was designed to comply with the 
radiological criteria for license termination for unrestricted use, 
the LTP supports this margin of safety.
    In addition, the LTP provides the methodologies and criteria 
that will be used to perform remediation activities of residual 
radioactivity to demonstrate compliance with the ALARA [As Low As 
Reasonably Achievable] criterion of 10 CFR 20.1402.
    Also, as previously discussed, the bounding accident for 
decommissioning is the materials handling event. Since the bounding 
decommissioning accident results in more airborne radioactivity than 
can be released from other decommissioning events, the margin of 
safety associated with the consequences of decommissioning accidents 
is not reduced by this activity. Therefore, the proposed change 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: Gerald Garfield, Esq., Day, Berry & Howard, 
City Place 1, Hartford, CT 06103.
    NRC Section Chief: Claudia Craig.

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 pdr@nrc.gov.

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois, Docket No. 50-219, Oyster Creek 
Generating Station, Ocean County, New Jersey, Three Mile Island Nuclear 
Station, Unit 1, Dauphin County, Pennsylvania, Docket No. 50-289

    Date of application for amendments: January 30, 2004.
    Brief description of amendment: The amendments conformed the 
Operating Licenses to reflect the current ownership structure of 
AmerGen Energy Company, LLC. Exelon Generation Company currently owns 
100% of AmerGen both directly and indirectly as a result of its 
purchase on December 22, 2003, of the stock of British Energy U.S. 
Holdings, Inc. The amendments deleted the License Conditions that are 
no longer valid as a result of the change of the AmerGen ownership.
    Date of Issuance: May 27, 2004.

[[Page 34709]]

    Effective date: These license amendments are effective as of their 
date of issuance.
    Amendment Nos.: 160, 243, 249.
    Facility Operating License Nos. DPR-16, DPR-50, and NPF-62: 
Amendments revised the Operating Licenses.
    Date of initial notice in Federal Register: March 2, 2004 (69 FR 
9859).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated May 27, 2004.
    No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-317, Calvert 
Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

    Date of application for amendment: May 1, 2003, as supplemented 
September 25, 2003, November 3, 2003, and February 25, 2004.
    Brief description of amendment: The amendment adds Technical 
Specification (TS) 3.7.16, ``Spent Fuel Pool Boron Concentration,'' 
modifies TS 4.3.1, ``Criticality'' and adds an additional license 
condition that requires the licensee to develop a long-term coupon 
surveillance program for the Carborundum samples.
    Date of issuance: June 3, 2004.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 267.
    Renewed Facility Operating License No. DPR-53: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28846).
    The September 25, 2003, November 3, 2003, and February 25, 2004, 
letters 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 Safety Evaluation dated June 3, 2004.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina.

    Date of application of amendments: October 16, 2001; as 
supplemented by letters dated May 20, September 12, and November 21, 
2002; September 22 and November 20, 2003; and February 18 and April 14, 
2004.
    Brief description of amendments: The amendments revised the 
Technical Specifications to incorporate changes resulting from use of 
an alternate source term.
    Date of Issuance: June 1, 2004.
    Effective date: These license amendments are effective as of the 
date of issuance and shall be implemented in accordance with the 
schedule provided in the licensee's letter dated February 18, 2004.
    Amendment Nos.: 338, 339 & 339.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2922).
    The supplements dated May 20, September 12, and November 21, 2002; 
and February 18 and April 14, 2004, provided additional information 
that clarified the application, 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 as 
published in the Federal Register on January 22, 2002 (67 FR 2922). The 
supplements dated September 22, 2003, and November 20, 2003, did change 
the NRC staff's proposed no significant hazards consideration 
determination. The NRC staff's proposed no significant hazards 
consideration determination based on the submittals dated September 22, 
2003, and November 20, 2003, were published in the Federal Register on 
October 14, 2003 (68 FR 59215), and December 9, 2003 (68 FR 68660), 
respectively.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 1, 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: July 29, 2003, and as 
supplemented by submittal dated January 14, 2004.
    Brief description of amendments: Revise the technical 
specifications by adding required actions for inoperable 250 VDC or 125 
VDC battery charger, by relocating certain DC power surveillance 
requirements and criteria to a licensee controlled program, and by 
providing alternative criteria for battery charger testing and battery 
monitoring with required actions. Additionally, a new program for 
battery monitoring and maintenance is added to the technical 
specifications.
    Date of issuance: June 8, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 207/199.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 14, 2003 (68 FR 
59215).
    The supplemental submittal contained clarifying information that 
was within the scope of the original application and did not change the 
staff's initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 8, 2004.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit No. 2, Beaver County, Pennsylvania

    Date of application for amendment: February 4, 2003, as 
supplemented by letters dated October 24, 2003, and April 6, 2004.
    Brief description of amendment: The amendment allowed the 
engineered safeguards features actuation system slave relay test 
frequency in footnote (1) to Technical Specification (TS) 4.3.2.1.1 to 
be changed from once per 92 days to once per 12 months provided a 
satisfactory contact loading analysis has been completed, and a 
satisfactory slave relay service life has been established, for the 
slave relay being tested.
    Date of issuance: May 14, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No: 141.
    Facility Operating License No. NPF-73. Amendment revised the TSs.
    Date of initial notice in Federal Register: March 18, 2003 (68 FR 
12953).
    The supplements dated October 24, 2003, and April 6, 2004, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 14, 2004.
    No significant hazards consideration comments received: No.

[[Page 34710]]

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

    Date of application for amendment: January 30, 2004.
    Brief description of amendment: The amendment eliminates 
requirements for hydrogen recombiners and relocates the requirements 
for hydrogen and oxygen monitors to the licensee's Commitment Tracking 
Program.
    Date of issuance: May 21, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 138.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 2, 2004 (69 FR 
9862).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 21, 2004.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of application for amendments: January 30, 2004, supplemented 
by letter dated May 6, 2004.
    Brief description of amendments: The amendments eliminate 
requirements for hydrogen recombiners and relocate the requirements for 
hydrogen monitors to the Technical Requirements Manual.
    Date of issuance: June 8, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment Nos.: 163 and 154.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 2, 2004 (69 FR 
9862).
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 8, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 1, 2003, as 
supplemented on March 10 and 30, 2004.
    Brief description of amendments: The amendments revised the 
Technical Specifications to change the peak calculated post accident 
primary containment internal pressure values for the primary 
containment leakage rate testing program.
    Date of issuance: May 28, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 241 and 184.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 20, 2004 (69 FR 
2747).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 28, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 30, 2003.
    Brief description of amendments: The amendments revised the staff 
position titles in Section 5.0 ``Administrative Controls'' of the 
Technical Specifications.
    Date of issuance: June 3, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 242 and 185.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 2, 2004 (69 FR 
9865).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 3, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 16, 2003, as supplemented March 
3, 2004.
    Brief description of amendments: The amendment provides a one-time 
change to Technical Specification 4.4.5.3a to extend the steam 
generator inspection interval to 44 months for STP, Unit 1.
    Date of issuance: June 8, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: Unit 1--162.
    Facility Operating License No. NPF-76: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2003 (68 
FR 64139). The supplement dated March 4, 2003, provided additional 
information that clarified the application, 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 June 8, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 23, 2004.
    Brief description of amendment: This amendment revises Technical 
Specification Surveillance Requirements 3.5.1.4, 3.5.4.3, and 3.6.7.3 
in order to delete a note that differentiates between the boron 
concentrations at North Anna, Units 1 and 2, for the safety injection 
accumulators, the refueling water storage tank, and the casing cooling 
tank.
    Date of issuance: June 4, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No: 218.
    Renewed Facility Operating License No. NPF-7: Amendment changes the 
Technical Specifications.
    Date of initial notice in Federal Register: March 30, 2004 (69 FR 
16624).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 4, 2004.
    No significant hazards consideration comments received: No.

[[Page 34711]]

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, 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 Public 
Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or 
by e-mail to pdr@nrc.gov.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. Within 60 days after the date 
of publication of this notice, the licensee may file a request for a 
hearing with respect to issuance of the amendment to the subject 
facility operating license and any person whose interest may be 
affected by this proceeding and who wishes to participate as a party in 
the proceeding must file a written request for a hearing and a petition 
for leave to intervene. Requests for a hearing and a petition for leave 
to intervene shall be filed in accordance with the Commission's ``Rules 
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. 
Interested persons should consult a current copy of 10 CFR 2.309, which 
is available at the Commission's PDR, located at One White Flint North, 
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland, and electronically on the Internet at the NRC Web site, 
http://www@nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR Reference staff at 
1-800-397-4209, 301-415-4737, or by e-mail to pdr@nrc.gov. If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert

[[Page 34712]]

opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
intends to rely to establish those facts or expert opinion. The 
petition must include sufficient information to show that a genuine 
dispute exists with the applicant on a material issue of law or 
fact.\1\ Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner to relief. A petitioner/requestor 
who fails to satisfy these requirements with respect to at least one 
contention will not be permitted to participate as a party.
---------------------------------------------------------------------------

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

    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) e-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, hearingdocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to 301-415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer or the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).

Duke Energy Corporation, Docket No. 50-270, Oconee Nuclear Station, 
Unit 2, Oconee County, South Carolina

    Date of amendment request: June 4, 2004.
    Description of amendment request: The amendment revised Technical 
Specification 3.6.5, ``Reactor Building Spray and Cooling Systems,'' to 
add a note that states that Limiting Condition of Operation 3.0.4 is 
not applicable.
    Date of issuance: June 4, 2004.
    Effective date: June 4, 2004.
    Amendment No.: 340.
    Facility Operating License No. DPR-47: Amendment revises the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): No. The Commission's related evaluation of the 
amendment, finding of emergency circumstances, state consultation, and 
final NSHC determination are contained in a safety evaluation dated 
June 4, 2004.
    Attorney for licensee: Anne W. Cottingham, Winston and Strawn LPP, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Stephanie M. Coffin, Acting.

    Dated at Rockville, Maryland, this 14th day of June 2004.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 04-13753 Filed 6-21-04; 8:45 am]
BILLING CODE 7590-01-P