Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 59082-59096 [05-20168]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 195 (Tuesday, October 11, 2005)]
[Notices]
[Pages 59082-59096]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 05-20168]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from September 6, 2005, to September 29, 2005. 
The last biweekly notice was published on September 27, 2005 (70 FR 
56499).

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.

[[Page 59083]]

    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, https://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 ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

[[Page 59084]]

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: April 6, 2005, as supplemented by letter 
dated August 8, 2005.
    Description of amendment request: The proposed amendment will 
modify Technical Specification (TS) 6.8.4.k, ``Containment Leakage Rate 
Testing Program,'' and TS Surveillance Requirement (SR) 4.6.1.6.1, 
``Containment Vessel Surfaces.'' The proposed amendment would modify 
the TS to allow for a one-time extension of the containment Type A test 
interval from once in 10 years to once in 15 years.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    This change does not involve a significant hazards consideration 
for the following reasons:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change to HNP [Harris Nuclear Plant] TS 6.8.4.k and 
TS SR 4.6.1.6.1 provide a one-time extension of the containment Type 
A test interval from 10 years to 15 years and specifies that 
additional visual inspections are done in accordance with 
Subsections IWE and IWL of the ASME [American Society of Mechanical 
Engineers] Section XI Code. The existing 10-year test interval is 
based on past test performance. The proposed TS change does not 
involve a physical change to the plant or a change in the manner in 
which the plant is operated or controlled. The containment vessel is 
designed to provide a leak-tight barrier against the uncontrolled 
release of radioactivity to the environment in the unlikely event of 
postulated accidents. As such, the containment vessel is not 
considered as the initiator of an accident. Therefore, the proposed 
TS change does not involve a significant increase in the probability 
of an accident previously evaluated.
    The proposed change involves only a one-time change to the 
interval between containment Type A tests. Type B and C leakage 
testing will continue to be performed at the intervals specified in 
10 CFR Part 50, Appendix J, Option A, as required by the HNP TS. As 
documented in NUREG-1493, ``Performance-Based Containment Leakage-
Test Program,'' industry experience has shown that Type B and C 
containment leak rate tests have identified a very large percentage 
of containment leak paths, and that the percentage of containment 
leak paths that are detected only by Type A testing is very small. 
In fact, an analysis of 144 integrated leak rate tests, including 23 
failures, found that none of the failures involved a containment 
liner breach. NUREG-1493 also concluded, in part, that reducing the 
frequency of containment Type A testing to once per 20 years results 
in an imperceptible increase in risk. The HNP test history and risk-
based evaluation of the proposed extension to the Type A test 
interval supports this conclusion. The design and construction 
requirements of the containment vessel, combined with the 
containment inspections performed in accordance with the American 
Society of Mechanical Engineers (ASME) Code, Section XI, and the 
Maintenance Rule (10 CFR 50.65) provide a high degree of assurance 
that the containment vessel will not degrade in a manner that is 
detectable only by Type A testing. Therefore, the proposed TS change 
does not involve a significant increase in the consequences of an 
accident previously evaluated.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change to HNP TS 6.8.4.k and TS SR 4.6.1.6.1 
provide a one-time extension of the containment Type A test interval 
to 15 years and specifies that additional visual inspections are 
done in accordance with Subsections IWE and IWL of the ASME Section 
XI Code. The existing 10-year test interval is based on past test 
performance. The proposed change to the Type A test interval does 
not result in any physical changes to HNP. In addition, the proposed 
test interval extension does not change the operation of HNP such 
that a failure mode involving the possibility of a new or different 
kind of accident from any accident previously evaluated is created. 
Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change to HNP TS 6.8.4.k and TS SR 4.6.1.6.1 
provide a one-time extension of the containment Type A test interval 
from 10 years to 15 years and specifies that additional visual 
inspections are done in accordance with Subsections IWE and IWL of 
the ASME Section XI Code. The existing 10-year test interval is 
based on past test performance. The NUREG-1493 study of the effects 
of extending containment leak rate testing found that a 20 year 
extension for Type A testing resulted in an imperceptible increase 
in risk to the public. NUREG-1493 found that, generically, the 
design containment leak rate contributes a very small amount to the 
individual risk and that the decrease in Type A testing frequency 
would have a minimal affect on this risk since most potential leak 
paths are detected by Type B and C testing. The proposed change 
involves only a one-time extension of the interval for containment 
Type A testing; the overall containment leak rate specified by the 
HNP TS is being maintained. Type B and C testing will continue to be 
performed at the frequency required by the HNP TS. The regular 
containment inspections being performed in accordance with the ASME 
Code, Section XI, and the Maintenance Rule (10 CFR 50.65) provide a 
high degree of assurance that the containment will not degrade in a 
manner that is only detectable by Type A testing. In addition, a 
plant-specific risk evaluation has demonstrated that the one-time 
extension of the Type A test interval from 10 years to 15 years 
results in a very small increase in risk for those accident 
sequences influenced by Type A testing.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: June 20, 2005.
    Description of amendment request: The amendment would revise 
Technical Specifications (TS) 3/4.4.7, ``Reactor Coolant System 
Chemistry.'' Specifically, the proposed amendment would revise the 
footnotes in Tables 3.4-2 and 4.4-3 of the TS to increase the 
temperature limit from 180 [deg]F to 250 [deg]F above which reactor 
coolant sampling and analysis for dissolved oxygen is required and 
dissolved oxygen limits apply.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    This amendment does not involve a significant hazards consideration 
for the following reasons:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Operation of HNP in accordance with the proposed amendment does 
not increase the

[[Page 59085]]

probability or consequences of accidents previously evaluated. The 
Final Safety Analysis Report (FSAR) documents the analyses of design 
basis accidents (DBA) at HNP. Any scenario or previously analyzed 
accident that results in offsite dose were evaluated as part of this 
analysis. The proposed amendment does not change or affect any 
accident previously evaluated in the FSAR. The proposed amendment 
does not modify any plant equipment. In addition, the proposed 
amendment does not result in a change to a structure, system, or 
component (SSC), or adversely affect its design function.
    The purpose of the temperature limit for RCS [Reactor Coolant 
System] oxygen control is to minimize corrosion at high temperatures 
on RCS components. Increasing the temperature at which oxygen levels 
are required to be maintained within specified limits from 180 
[deg]F to 250 [deg]F is supported by industry and vendor data which 
indicates that the influence of dissolved oxygen at or below 250 
[deg]F is not significant with regard to stress corrosion cracking 
and general corrosion of RCS components. The proposed amendment is 
consistent with the Electric Power Research Institute's (EPRI's) 
guidelines for Pressurized Water Reactor (PWR) Primary Water 
Chemistry. This amendment places HNP in line with standard industry 
specifications for reactors of similar size and vintage. HNP's 
proposed amendment to increase the temperature limit for 
applicability to 250 [deg]F would decrease the time needed to 
achieve compliance with the dissolved oxygen limit and decrease the 
overall time to restart the plant from cold shutdown. Removing 
oxygen in a more expeditious fashion enhances RCS chemistry. Based 
on the above, RCS integrity is maintained by this amendment.
    Therefore, this amendment 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.
    Operation of HNP in accordance with the proposed amendment does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated. The FSAR documents the 
analyses of design basis accidents (DBA) at HNP. Any scenario or 
previously analyzed accident that results in offsite dose were 
evaluated as part of this analysis. The proposed amendment does not 
change or affect any accident previously evaluated in the FSAR, and 
no new or different scenarios are created by the proposed amendment 
to the TS. The proposed amendment does not modify any plant 
equipment. In addition, the proposed amendment does not result in a 
change to an SSC [structure, system, or component] or adversely 
affect its design function.
    The purpose of the temperature limit for RCS oxygen control is 
to minimize corrosion at high temperatures on RCS components. 
Increasing the temperature at which oxygen levels are required to be 
maintained within specified limits from 180 [deg]F to 250 [deg]F is 
supported by industry and vendor data which indicates that the 
influence of dissolved oxygen at or below 250 [deg]F is not 
significant with regard to stress corrosion cracking and general 
corrosion of RCS components. The proposed amendment is consistent 
with EPRI's guidelines for PWR Primary Water Chemistry. This 
amendment places HNP in line with standard industry specifications 
for reactors of similar size and vintage. HNP's proposed amendment 
to increase the temperature limit for applicability to 250 [deg]F 
would decrease the time needed to achieve compliance with the 
dissolved oxygen limit and decrease the overall time to restart the 
plant from cold shutdown. Removing oxygen in a more expeditious 
fashion enhances RCS chemistry. Based on the above, RCS integrity is 
maintained by this amendment.
    Therefore, this amendment 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.
    Operation of HNP in accordance with the proposed amendment does 
not involve a significant reduction in a margin of safety. Existing 
TS operability and surveillance requirements are not reduced by the 
proposed amendment. The proposed amendment does not modify any plant 
equipment. In addition, the proposed amendment does not result in a 
change to a structure, system, or component (SSC), or its design 
function. The proposed amendment does not adversely affect existing 
plant safety margins or the reliability of equipment assumed to 
mitigate accidents in the FSAR.
    The purpose of the temperature limit for RCS oxygen control is 
to minimize corrosion at high temperatures on RCS components. 
Increasing the temperature at which oxygen levels are required to be 
maintained within specified limits from 180 [deg]F to 250 [deg]F is 
supported by industry and vendor data which indicates that the 
influence of dissolved oxygen at or below 250 [deg]F is not 
significant with regard to stress corrosion cracking and general 
corrosion of RCS components. The proposed amendment is consistent 
with EPRI's guidelines for PWR Primary Water Chemistry. This 
amendment places HNP in line with standard industry specifications 
for reactors of similar size and vintage. HNP's proposed amendment 
to increase the temperature limit for applicability to 250 [deg]F 
would decrease the time needed to achieve compliance with the 
dissolved oxygen limit and decrease the overall time to restart the 
plant from cold shutdown. Removing oxygen in a more expeditious 
fashion enhances RCS chemistry. Based on the above, RCS integrity is 
maintained by this amendment.
    Therefore, this amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: June 20, 2005.
    Description of amendment request: The proposed amendment would 
revise Cooper Nuclear Station (CNS) Technical Specification (TS) 5.3, 
``Unit Staff Qualifications,'' to upgrade the qualification standard 
for the Shift Manager, Senior Operator, Licensed Operator, and Shift 
Technical Engineer from Regulatory Guide (RG) 1.8, Revision 2 
``Qualification and Training of Personnel for Nuclear Power Plants,'' 
to RG 1.8, Revision 3. It also clarifies qualification requirements 
applicable to the Operations Manager position.
    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.
    These changes are administrative in nature and do not require 
any physical modifications, affect any plant components, or result 
in any changes in plant operation. They provide clarity and 
consistency to the CNS licensing basis.
    Upgrading the unit staff qualifications for the Shift Manager, 
Senior Operator, Licensed Operator, and Shift Technical Engineer 
from Regulatory Guide 1.8, Revision 2, to Regulatory Guide 1.8, 
Revision 3, is an administrative change that will clarify the 
current requirements for qualification and training of operations 
personnel. The changes are consistent with the application of a 
systems approach to training in an accredited training program. By 
promulgation of the 10 CFR Part 55 rule change, the NRC determined 
that an accredited licensed operator training program based on a 
systems approach to training provides an acceptable means of 
qualifying licensed operating personnel.
    The addition of qualification requirements for the Operations 
Manager position clarifies SRO [Senior Reactor Operator] license 
requirements for Operations management personnel by specifying that 
the Operations Supervisor is the member of Operations management 
required to have a current SRO license at CNS. The Operations 
Manager is required to hold or have previously held a

[[Page 59086]]

SRO license. This will ensure an acceptable level of operations 
knowledge to perform in a managerial oversight role. This approach 
is consistent with current guidance in ANSI/ANS [American Nuclear 
Standards Institute/American Nuclear Society] 3.1-1993. This change 
is administrative in nature and has no impact on previously 
evaluated accidents.
    Therefore, these changes do 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.
    These changes are administrative in nature and do not involve a 
physical alteration of the plant or a change to plant operations. No 
new failure mechanisms, malfunctions, or accident initiators are 
introduced. The proposed changes provide clarity and consistency to 
the CNS licensing basis in regard to training and qualification of 
operations personnel and SRO license requirements for Operations 
management personnel.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    Response: No.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    These changes are administrative in nature and do not affect any 
Technical Specification safety limit or limiting condition for 
operation. No safety margins are affected by these changes. The 
proposed changes do not involve a change in plant design or 
operation for the mitigation of postulated accidents. The proposed 
changes provide clarity and consistency to the CNS licensing basis 
in regard to training and qualification of operations personnel and 
SRO license requirements for Operations management personnel.
    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. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: David Terao.

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

    Date of amendment request: August 25, 2005.
    Description of amendment request: The proposed amendment would 
revise the definitions of Channel Calibration, Channel Function Test, 
and Logic System Functional Test in accordance with the Technical 
Specification Task Force Traveler 205-A.
    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?
    Response: No.
    The definitions of Channel Calibration, Channel Functional Test, 
and Logic System Functional Test specified in Technical 
Specifications (TS) provide basic information regarding what the 
test involves, the components involved in the test, and general 
information regarding how the test is to be performed. These 
definitions and their specific wording are not precursors to any 
accident. As a result these revised definitions result in no 
increase in the probability of an accident previously evaluated.
    The proposed revisions of these definitions involve no changes 
to plant design, equipment, or operation related to mitigation of 
accidents. The proposed revisions of these definitions do not change 
their meaning or intent. The proposed revisions clarify the 
definitions and do not result in a reduction of required testing of 
instrumentation used to mitigate 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?
    Response: No.
    The proposed revisions of the definitions do not involve a 
change to the design or operation of any plant structure, system, or 
component (SSC). As a result the plant will continue to be operated 
in the same manner. The proposed revisions will not result in a 
change to how the instrumentation used to monitor plant operation 
and to mitigate accidents is tested. Operating the plant and testing 
the plant's instrumentation in the same manner as is currently done 
will not create the possibility of a new or different kind of 
accident.
    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?
    Response: No.
    The affected definitions involve testing of instrumentation used 
in the mitigation of accidents to ensure that the instrumentation 
will perform as assumed in safety analyses. The proposed revisions 
of these definitions will not change their meaning or intent. As a 
result, the instrumentation will continue to be tested in the same 
manner as is currently done. Revising these definitions as proposed 
will not result in a change to the design or operation of any plant 
SSC used to shutdown the plant, initiate the Emergency Core Cooling 
Systems, or isolate primary or secondary containment. As a result 
the ability of the plant to respond to and mitigate accidents is 
unchanged by the revised definitions.
    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 C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: David Terao.

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: July 29, 2005.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification 3.7.5, ``Auxiliary Feedwater (AFW) 
System,'' to change the frequency of Surveillance Requirement 3.7.5.6 
from 92 days to 24 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to increase [the] frequency interval for 
Surveillance Requirement (SR) 3.7.5.6 from 92 days to 24 months has 
no impact on the probability of accidents previously evaluated. The 
valves controlled by SR 3.7.5.6 are used to provide an alternate 
supply of water to the auxiliary feedwater (AFW) system from the 
fire water storage tank (FWST) and are only operated after an 
accident has occurred. They are not accident initiators.
    Misoperation, or failure of a[n] FWST supply to be correctly 
positioned following an accident, could result in an inadequate 
supply of water to the AFW system. Failure to provide adequate core 
cooling could increase the radiological consequences of an accident. 
However, operating and maintenance histories of the FWST supply 
valves show that these valves have been

[[Page 59087]]

capable of full stroke cycling each time they have been tested. 
There is no evidence of any time-related degradation mechanism that 
would prevent the valves from performing their design function. 
Thus[,] the proposed change has no impact on the consequences of an 
accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different [kind of] accident from any accident previously evaluated?
    Response: No.
    The proposed change to increase frequency interval for SR 
3.7.5.6 from 92 days to 24 months has no impact on the probability 
of accidents of the type evaluated in the Final Safety Analysis 
Report, as updated. The valves are used to provide an alternate 
supply of water to the AFW system from the FWST, and are only 
operated after an accident has occurred. They are not accident 
initiators. Review of the operating and maintenance histories of the 
FWST supply valves show that they are highly reliable in maintaining 
their capability to perform their design function.
    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 proposed change to SR 3.7.5.6 involves only an increase in 
the frequency interval. No physical changes are required to the 
facility or to the plant operating or emergency procedures as a 
result of the change. Based on review of the operating and 
maintenance histories of the FWST supply valves, they have been 
capable of full stroke cycling each time they have been tested. 
There is no evidence of any time-related degradation mechanism that 
would prevent the valves from performing their design function. This 
evidence supports the conclusion that there will be no impact in the 
operation of these valves following an accident.
    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 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: Daniel S. Collins (Acting).

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: August 23, 2005.
    Description of amendment requests: The proposed amendments would 
revise the expiration dates of the Units 1 and 2 facility-operating 
licenses to recapture low-power testing time, and to reflect a 40-year 
term measured from the date of issuance of each unit's full-power 
operating license.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed additional operating license periods do not affect 
the probability or consequences of an accident previously evaluated 
since they require no physical change in the plant equipment or 
operating procedures and the Final Safety Analysis Report (FSAR) 
Update safety analyses are based on [a] 40-year full[-]power 
operation. Surveillance and maintenance practices, as well as other 
programs such as environmental qualification of equipment, ensure 
timely identification and correction of any degradation of safety-
related plant equipment. The long-term integrity of the reactor 
vessels has been evaluated using currently acceptable NRC 
calculational methods and best available Diablo Canyon Power Plant 
(DCPP) specific data. The evaluation results demonstrate that both 
reactor vessels are safe for normal operations in excess of 40 
years.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different [kind of] accident from any accident previously evaluated?
    Response: No.
    The possibility of a new or different kind of accident is not 
created by the proposed additional operating periods since at least 
40 years of full[-]power operation was assumed in the design and 
construction of DCPP Units 1 and 2. The plant maintenance programs 
are also designed to both maintain and determine the need to replace 
safety-related components. These programs will continue to be 
applied as they are presently to assure safe operation.
    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 proposed additional operating periods do not involve a 
significant reduction in a margin of safety since, as is the case 
with present operation, degradation of safety-related equipment will 
be identified and corrected by ongoing surveillance and maintenance 
practices. Existing programs, routine maintenance, and compliance 
with Technical Specifications assure that an adequate margin of 
safety is maintained. These activities will remain in effect for the 
duration of the proposed additional operating periods.
    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 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: Daniel S. Collins (Acting).

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of amendment request: June 30, 2005.
    Description of amendment request: The proposed changes would revise 
the Administrative Control section of the Technical Specifications 
(TSs) to permit the Westinghouse best estimate methodology for loss-of-
coolant-accident (LOCA) analysis methodology to be utilized for 
analyses as required by Title 10 of the Code of Federal Regulations, 
Part 50, Section 46, ``Acceptance criteria for emergency core cooling 
systems [ECCS] for light water nuclear power reactors' (10 CFR 50.46).
    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.
    Implementation of the best-estimate large break LOCA methodology 
and associated TS changes is proposed to increase margin to the peak 
clad temperature limits defined in 10 CFR 50.46. There are no 
physical plant changes or changes in manner in which the plant will 
be operated as a result of this

[[Page 59088]]

change. Since the plant conditions and ECCS performance assumed in 
the analysis are consistent with the plant's current design, the 
proposed change in methodology will thus have no impact on the 
probability of a LOCA. When applied, the best estimate methodology 
shows that the ECCS is more effective than previously evaluated in 
mitigating the consequences of a LOCA, as lower peak clad 
temperatures are predicted relative to current 10 CFR 50.46 Appendix 
K results. Since the proposed best-estimate methodology is only 
applicable to a large break LOCA and since the application of the 
proposed methodology shows there is a high probability that all of 
the acceptance criteria contained in 10 CFR 50.46, Paragraph b are 
met, the proposed change does not increase the consequences of an 
accident previously evaluated.
    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.
    There are no physical changes being made to the plant. No new 
modes of plant operation are being introduced. The parameters 
assumed in the analysis remain within the design limits of the 
existing plant equipment. All plant systems will perform as designed 
during the response to a potential accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously analyzed.
    3. Does this change involve a significant reduction in a margin 
of safety?
    Response: No.
    It has been shown that the methodology used in the analysis 
would more realistically describe the expected behavior of V. C. 
Summer Nuclear Station systems during a postulated loss of coolant 
accident. Uncertainties have been accounted for as required by 10 
CFR 50.46. A sufficient number of loss of coolant accidents with 
different break sizes, different locations and other variations in 
properties are analyzed to provide assurance that the most severe 
postulated loss of coolant accidents are calculated. It has been 
shown by analysis that there is a high level of probability that all 
criteria contained in 10 CFR 50.46, Paragraph b are met.
    Pursuant to 10 CFR 50.91, the preceding analyses provide a 
determination that the proposed Technical Specifications change 
poses no significant hazard as delineated by 10 CFR 50.92.

    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 G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: Evangelos C. Marinos.

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

    Date of amendment request: August 30, 2005.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) to reflect the use of the 
Westinghouse Best Estimate Analyzer for Core Operations--Nuclear 
(BEACON) to augment the functional capability of the flux mapping 
system for the purpose of power distribution surveillances. In 
addition, editorial changes to the TSs are proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The PDMS [power distribution monitoring system] performs 
continuous core power distribution monitoring. This system utilizes 
the NRC-approved Westinghouse proprietary computer code BEACON to 
provide data reduction for incore flux maps, core parameter 
analysis, load follow operation simulation, and core prediction. It 
in no way provides any protection or control system function. 
Fission product barriers are not impacted by these proposed changes. 
The proposed changes occurring with PDMS will not result in any 
additional challenges to plant equipment that could increase the 
probability of any previously evaluated accident. The changes 
associated with the PDMS do not affect plant systems such that their 
function in the control of radiological consequences is adversely 
affected. These proposed changes will therefore not affect the 
mitigation of the radiological consequences of any accident 
described in the Updated Final Safety Analysis Report Update 
(UFSAR).
    Continuous on-line monitoring through the use of PDMS provides 
significantly more information about the power distributions present 
in the core than is currently available. This results in more time 
(i.e., earlier determination of an adverse condition developing) for 
operator action prior to having an adverse condition develop that 
could lead to an accident condition or to unfavorable initial 
conditions for an accident.
    Each accident analysis addressed in the UFSAR is examined with 
respect to changes in cycle-dependent parameters, which are obtained 
from application of the NRC-approved reload design methodologies, to 
ensure that the transient evaluations of reload cores are bounded by 
previously accepted analyses. This examination, which is performed 
in accordance with the requirements set forth in 10 CFR [Title 10 of 
the Code of Federal Regulations] 50.59, ensures that future reloads 
will not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    The three editorial changes only correct typographical errors 
made in previously approved TS changes. They do not affect plant 
operation or structures, systems, and components important to 
safety.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequence 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 implementation of the PDMS has no influence or impact on 
plant operations or safety, nor does it contribute in any way to the 
probability or consequences of an accident. No safety-related 
equipment, safety function, or plant operation will be altered as a 
result of this proposed change. The possibility for a new or 
different type of accident from any accident previously evaluated is 
not created since the changes associated with implementation of the 
PDMS do not result in a change to the design basis of any plant 
component or system. The evaluation of the effects of using the PDMS 
to monitor core power distribution parameters shows that all design 
standards and applicable safety criteria limits are met.
    The proposed changes do not result in any event previously 
deemed incredible being made credible. Implementation of the PDMS 
will not result in more adverse conditions and will not result in 
any increase in the challenges to safety systems. The cycle-specific 
variables required by the PDMS are calculated using NRC-approved 
methods. The TS will continue to require operation within the 
required core operating limits and appropriate actions will be taken 
if limits are exceeded.
    The three editorial changes only correct typographical errors 
made in previously approved TS changes. They do not affect plant 
operation or structures, systems, and components important to 
safety.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is not affected by implementation of the 
PDMS. The margin of safety provided by current TS is unchanged. The 
proposed changes continue to require operation within the core 
limits that are based on NRC-approved reload design methodologies. 
Appropriate measures exist to control the values of these cycle-
specific limits. The proposed changes continue to ensure that 
appropriate actions will be taken

[[Page 59089]]

if limits are violated. These actions remain unchanged.
    The three editorial changes only correct typographical errors 
made in previously approved TS changes. They do not affect plant 
operation or structures, systems, and components important to 
safety.

    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: David Terao.

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

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

    Date of application for amendment: April 3, 2003, as supplemented 
December 23, 2003, December 9 and 17, 2004, and March 30 and August 19, 
2005.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to support the application of an alternative 
source term methodology in accordance with Title 10 of the Code of 
Federal Regulations, Section 50.67, ``Accident Source Term,'' with the 
exception that Technical Information Document 14844, ``Calculation of 
Distance Factors for Power and Test Reactor Sites,'' was used as the 
radiation dose basis for equipment qualification.
    Date of issuance: September 19, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 167.
    Facility Operating License No. NPF-62: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: September 2, 2003 (68 
FR 52234).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 19, 2005.
    The supplements dated December 23, 2003, December 9 and 17, 2004, 
and March 30 and August 19, 2005 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.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 11, 2003, as 
supplemented April 16 and September 10, 2004, and March 30 and 
September 21, 2005.
    Brief description of amendment: The amendment revised the 
instrument channel trip setpoint allowable values for thirteen 
Technical Specification (TS) functions at Clinton Power Station, Unit 
1.
    Date of issuance: September 27, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 168.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 16, 2004 (69 FR 
12363).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 21, 2005. The supplements dated 
April 16 and September 10, 2004, and March 30 and September 21, 2005, 
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. No significant hazards consideration 
comments received: No.

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

    Date of application for amendments: August 3, 2004, as supplemented 
on July 8 and August 26, 2005.
    Brief description of amendments: The amendments extend the 
surveillance frequency interval from monthly to quarterly for Technical 
Specification surveillance requirement (SR) 3.3.3.1, which involves a 
channel functional test of each reactor trip circuit breaker (RTCB). 
SRs 3.3.3.1 and 3.3.3.2 will be scheduled such that the RTCBs testing 
is performed every 6 weeks, which meets the vendor-recommended interval 
for cycling each RTCB.
    Date of issuance: September 26, 2005.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: 275 and 252.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 4, 2005 (70 FR 
400).

[[Page 59090]]

    The July 8 and August 26, 2005, supplemental 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 these amendments is 
contained in a Safety Evaluation dated September 26, 2005.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-245, Millstone Power 
Station Unit No. 1, New London County, Connecticut

    Date of application for amendment: September 8, 2004, as 
supplemented by letters dated May 5 and July 27, 2005.
    Brief description of amendment: The amendment revised the Millstone 
Power Station, Unit No. 1 Technical Specifications (TSs) to support the 
implementation of the proposed Dominion Nuclear Facility Quality 
Assurance Program (Topical Report DOM-QA-1). Implementation of this 
Topical Report would create a common quality assurance program for all 
sites owned by Dominion Nuclear Connecticut, Inc. Review of this 
proposed amendment was requested in concert with the review of the 
Topical Report.
    Date of issuance: September 15, 2005.
    Effective date: As of the date of issuance, and shall be 
implemented by February 28, 2006.
    Amendment No.: 115.
    Facility Operating License No. DPR-21: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2888).
    The additional information provided in the supplemental letters 
dated May 5 and July 27, 2005, did not expand the scope of the 
application as noticed and did not change the NRC staff's original 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 15, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 15, 2004, as supplemented 
by letter dated August 23, 2004.
    Brief description of amendment: The amendment revised the Facility 
Operating License DPR-65 to address the resolution of a non-
conservative Technical Specifications (TSs) associated with control 
room isolation radiation monitoring instrumentation. Specifically, the 
amendment would revise the TSs to require two operable channels of 
control room isolation radiation monitoring instrumentation.
    Date of issuance: September 23, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 289.
    Facility Operating License No. DPR-65: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2887).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 23, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 15, 2004, as supplemented 
on June 23, 2005.
    Brief description of amendment: The amendment approves 
modifications to the Fire Protection Program. Specifically, the 
modifications involve converting the existing automatic carbon dioxide 
fire suppression systems installed in the cable spreading room to 
manual actuation.
    Date of issuance: September 22, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 227.
    Facility Operating License No. NPF-49: The amendment allows for 
conversion from an automatic to a manual carbon dioxide suppression 
system in the cable spreading area.
    Date of initial notice in Federal Register: July 6, 2004 (69 FR 
40672). The supplement dated June 23, 2005, provided clarifying 
information and did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 22, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 8, 2004, as 
supplemented by letters dated May 5 and July 27, 2005.
    Brief description of amendments: The amendments revised the 
Millstone Power Station, Unit N