[Federal Register Volume 69, Number 22 (Tuesday, February 3, 2004)]
[Notices]
[Pages 5200-5216]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 04-2017]


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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 Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from, January 9, 2004, through January 22, 2004. 
The last biweekly notice was published on January 20, 2004 (69 FR 
2735).

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.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By March 4, 2004, 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 2.714, 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

[[Page 5201]]

leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, 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 factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide 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 to relief. A 
petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, 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 with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, Public File Area 01F21, 11555 
Rockville Pike (first floor), Rockville, Maryland, by the above date. 
Because of continuing disruptions in delivery of mail to United States 
Government offices, it is requested that petitions for leave to 
intervene and requests for hearing be transmitted to the Secretary of 
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to hearingdocket@nrc.gov. 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 because of continuing disruptions in 
delivery of mail to United States Government offices, 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 filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    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, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: December 3, 2003, as supplemented by 
letter dated January 14, 2004.
    Description of amendment request: Appendix B, Additional 
Conditions, to Operating License No. DPR-23 for H. B. Robinson Steam 
Electric Plant (HBRSEP), Unit No. 2, contains the following condition: 
``Operation of H. B. Robinson Steam Electric Plant, Unit No. 2, is 
limited to 504 effective full power days. This additional condition 
shall remain in effect until approval of a license amendment that 
removes this limitation.'' The proposed change will delete the 
condition described above.
    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:

    An evaluation of the proposed change has been performed in 
accordance with 10 CFR 50.91(a)(1) regarding no significant hazards 
considerations using the standards in 10 CFR 50.92(c). A discussion 
of these standards as they relate to this amendment request follows:
    1. The Proposed Change Does Not Involve a Significant Increase 
in the Probability or Consequences of an Accident Previously 
Evaluated.
    The proposed change to Appendix B of the HBRSEP, Unit No. 2, 
Operating License

[[Page 5202]]

deletes a restriction on effective full-power days (EFPD) that was 
incorporated to ensure the source term used for radiological dose 
analyses remains bounded by the analyses of record for operation at 
the approved, uprated power level. The restriction was imposed 
solely for the post-accident radiological analyses assumption. Since 
this restriction is only related to post-accident analytical 
assumptions, it is unrelated to the probability of an accident 
occurring. Therefore, the proposed Operating License change does not 
involve a significant increase in the probability of an accident 
previously evaluated.
    The proposed change can impact the consequences of previously 
evaluated accidents by impacting the core inventory of radionuclides 
for operating periods exceeding the existing 504 EFPD restriction. 
An evaluation of the potential impact of removing the EFPD 
restriction on the accident consequences has determined that any 
increase in consequences would be less than 10% of the difference 
between the existing dose analysis results and the acceptable dose 
limits. The proposed change therefore results in less than a minimal 
increase in accident consequences. Therefore, the proposed 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 Change Does Not Create the Possibility of a New 
or Different Kind of Accident From Any Previously Evaluated.
    The proposed change to Appendix B of the HBRSEP, Unit No. 2, 
Operating License deletes a restriction on effective full-power days 
(EFPD) that was incorporated to ensure the source term used for 
radiological dose analyses remain bounded by the dose analyses of 
record for operation at the approved, uprated power level. The 
restriction was imposed solely for post-accident radiological 
analyses assumptions. Since this restriction is only related to 
post-accident analytical assumptions, it is unrelated to the 
possibility of an accident occurring. Therefore, this 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 the Margin of Safety.
    The applicable margin of safety is that related to the dose 
consequences of analyzed accidents. The proposed change results in 
potential increased consequences that are less than 10% of the 
difference between the existing dose analyses results and acceptable 
dose limits. This is less than a minimal increase in accident 
consequences, as defined by NEI [Nuclear Energy Institute] 96-07, 
Revision 1, which is endorsed by Regulatory Guide 1.187. Therefore, 
this change does not involve a significant reduction in a margin of 
safety.
    Based on the above discussion, Progress Energy Carolinas, Inc. 
[Carolina Power & Light Company] has determined that the requested 
change does not involve a significant hazards consideration.

    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: Allen G. Howe.

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

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

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

    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen and oxygen monitors are no longer required to mitigate 
design-basis accidents and, therefore, the hydrogen monitors do not 
meet the definition of a safety-related component as defined in 10 
CFR 50.2. RG [Regulatory Guide] 1.97 Category 1, is intended for key 
variables that most directly indicate the accomplishment of a safety 
function for design-basis accident events. The hydrogen and oxygen 
monitors no longer meet the definition of Category 1 in RG 1.97. As 
part of the rulemaking to revise 10 CFR 50.44 the Commission found 
that Category 3, as defined in RG 1.97, is an appropriate 
categorization for the hydrogen monitors because the monitors are 
required to diagnose the course of beyond design-basis accidents. 
Also, as part of the rulemaking to revise 10 CFR 50.44, the 
Commission found that Category 2, as defined in RG 1.97, is an 
appropriate categorization for the oxygen monitors, because the 
monitors are required to verify the status of the inert containment.
    The regulatory requirements for the hydrogen and oxygen monitors 
can be relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, 
[classification of the oxygen monitors as Category 2,] and removal 
of the hydrogen and oxygen monitors from TS will not prevent an 
accident management strategy through the use of the severe accident 
management

[[Page 5203]]

guidelines (SAMGs), the emergency plan (EP), the emergency operating 
procedures (EOPs), and site survey monitoring that support 
modification of emergency plan protective action recommendations 
(PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen and oxygen monitor 
requirements, including removal of these requirements from TS, does 
not involve a significant increase in the probability or the 
consequences of any accident previously evaluated.

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

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

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

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

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: L. Raghavan.

Exelon Generation Company, LLC, Docket No. 50-352, Limerick Generating 
Station, Unit 1, Montgomery County, Pennsylvania

    Date of amendment request: December 22, 2003.
    Description of amendment request: Exelon Generation Company, LLC, 
the licensee, is proposing a change to the Limerick Generating Station 
(LGS), Unit 1, Technical Specifications (TSs) contained in Appendix A 
to the Operating License. This proposed change will revise the TS 
section on safety limits to incorporate revised safety limit minimum 
critical power ratios (SLMCPRs) based on cycle-specific analysis 
performed by Global Nuclear Fuel for LGS, Unit 1, Cycle 11.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Changing the SLMCPRs does not require any physical plant 
modifications, physically affect any plant components, or involve 
changes in plant operation. Therefore, the probability of an 
accident previously evaluated remains unchanged. The operability of 
plant systems designed to mitigate any consequences of accidents has 
not changed, therefore, the consequences of an accident previously 
evaluated are not expected to increase.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    The proposed change does not involve any modifications of the 
plant configuration for allowable modes of operation. The SLMCPRs 
are not accident initiators, and their revision will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The proposed SLMCPRs provide a margin of safety by ensuring that 
no more than 0.1% of the fuel rods are in a boiling transition if 
the operating limit minimum critical power ratios are exceeded 
during any mode of operation. Although the SLMCPRS are being reduced 
from 1.10 to 1.07 for two loop operation, and from 1.11 to 1.08 for 
single loop operation, the SLMCPRs continue to ensure that during 
normal operation and abnormal operational transients at least 99.9% 
of all fuel rods in the core do not experience transition boiling if 
the limit is not violated when all uncertainties are considered, 
thereby preserving the fuel cladding integrity. Therefore, the 
proposed TS change will not involve a significant reduction in a 
margin of safety.

    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Mr. Edward Cullen, Vice President & General 
Counsel, Exelon Generation Company, LLC, 2301 Market Street, 
Philadelphia, PA 19101.
    NRC Section Chief: Darrell Roberts, Acting.

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

    Date of amendment request: November 25, 2003.
    Description of amendment request: Exelon Generation Company, LLC, 
the licensee, is proposing a change to the Limerick Generating Station 
(LGS), Units 1 and 2, Technical Specifications (TSs) contained in 
Appendix A to Operating Licenses NPF-39 and NPF-85, respectively. This 
proposed change will add a footnote to TS 3.4.3.2.e to indicate that 
reactor coolant system (RCS) pressure isolation valve (PIV) leakage is 
excluded from any other allowable RCS operational leakage specified in 
TS Section 3.4.3.2.
    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. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No. PIV leakage is not operational leakage. PIV 
leakage limits are used in conjunction with the system 
specifications for the PIVs to ensure that plant operation is 
appropriately limited. The

[[Page 5204]]

PIV leakage limit provides for monitoring the condition of the RCPB 
[Reactor Coolant Pressure Boundary] to detect PIV degradation. 
Although the proposed change will result in a change to the current 
method of calculating the RCS operational leakage, the proposed 
change does not affect the actual PIV leakage limit itself, and 
therefore, does not affect the ability to detect PIV degradation. 
The proposed change does not affect the basis for the safety 
analysis used to determine the probability or consequences of an 
accident since PIV leakage is not considered in any design basis 
accident.
    Although the effect of the proposed change will allow for the 
potential increase in identified leakage, the total RCS operational 
leakage is still limited by the Technical Specifications (TS) 
Limiting Condition for Operation (LCO) which itself is not being 
changed. In addition, current TS Applicability, Action and 
Surveillance requirements for detection, monitoring, and 
appropriately limiting operational leakage are not being changed.
    The proposed change does not alter the leakage detection system 
monitors, design features, operation, or accident analysis 
assumptions which could affect the ability of the reactor coolant 
pressure boundary (RCPB) to mitigate the consequences of a 
previously evaluated accident. The proposed change will not increase 
the likelihood of the malfunction of another system, structure or 
component which has been assumed as an accident initiator or 
credited in the mitigation 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 operational leakage requirements for the RCPB 
leakage, unidentified leakage, and total leakage ensures corrective 
action can be taken to protect the RCPB from degradation. The PIV 
leakage provides an indication that the PIVs, between the RCS and 
the connecting systems, are degraded or degrading.
    No change in the ability to perform the design function of the 
leak detection system, the protection afforded by the operational 
leakage requirements, or PIV leakage requirements is involved. No 
change in the operation of the leak detection system or PIVs is 
required. Instrumentation setpoints, monitoring frequencies and 
leakage limitations associated with RCS operational leakage and PIV 
leakage are not affected by the proposed change. No modifications to 
the PlVs or RCS leak detection system or associated components are 
required to implement the proposed change. Therefore, no new failure 
mechanism, malfunction, or accident initiator is considered 
credible.
    Additionally, the proposed change does not affect other plant 
design, hardware, or system 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 change does not involve a relaxation 
of the criteria used to establish safety limits, a relaxation of the 
bases for the limiting safety system settings, or a relaxation of 
the bases for the limiting conditions for operation, other than 
excluding PIV leakage from the other RCS operational leakage.
    Controlling values for the RCS operational leakage and PIV 
leakage are included in current TS testing measurements, monitors, 
detection methods and procedures. The proposed change will not 
modify these requirements or the accident analysis assumptions 
regarding the performance of the RCS operational leakage and PIV 
leakage monitoring which could potentially challenge safety margins 
established to ensure fuel cladding integrity, as well as reactor 
coolant and containment system integrity.
    The safety analyses of the RCPB integrity and the ability to 
mitigate accidents do not require revision in order to implement the 
proposed change. Modification of the existing margins is not 
required.
    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. Edward Cullen, Vice President & General 
Counsel, Exelon Generation Company, LLC, 2301 Market Street, 
Philadelphia, PA 19101.
    NRC Section Chief: Darrell Roberts, Acting.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of amendment request: October 14, 2003.
    Description of amendment request: The proposed changes would 
eliminate License information that no longer applies to a license that 
has permanently ceased operation. The proposed changes would also 
simplify the Technical Specifications. Maine Yankee proposes to remove 
certain design and administrative requirements, relocate them to the 
Defueled Safety Analysis Report (DSAR) (i.e., Updated Final Safety 
Analysis Report for Maine Yankee), or the Quality Assurance Program and 
make other minor administrative changes. The DSAR is controlled by 10 
CFR 50.59, and the Quality Assurance Program is controlled by 10 CFR 
50.54(a). The Technical Specification relocation is being proposed 
pursuant to the criteria contained in 10 CFR 50.36, and is consistent 
with NRC Administrative Letter 95-06. Additionally, Maine Yankee 
proposes to eliminate technical specifications which will no longer be 
applicable following the transfer of the last fuel assembly from the 
spent fuel pool to spent fuel storage cask.
    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 License changes delete License information that 
does not apply to a plant that has permanently ceased operation. 
These changes are in compliance with 10 CFR Part 50 regulations and 
are not associated with the probability or consequences of accidents 
previously evaluated.
    The proposed Technical Specification changes reflect the 
complete transfer of all spent nuclear fuel from the Spent Fuel Pool 
(SFP) to the Independent Spent Fuel Storage Installation (ISFSI). 
Design basis accidents related to the Spent Fuel Pool are discussed 
in the MY Defueled Safety Analysis Report (DSAR). These postulated 
accidents are predicated on spent nuclear fuel being stored in the 
Spent Fuel Pool. With the removal of the spent fuel from the Spent 
Fuel Pool, there are no remaining safety related systems required to 
be monitored and there are no remaining credible design basis 
accidents related to the SFP.
    The proposed relocation of the specified minimum distance to the 
Exclusion Area Boundary from the Technical Specification to the DSAR 
has no impact on the probability or consequences of the remaining 
applicable design basis accidents.
    The proposed changes do not affect design functions of 
structures, systems or components (SSC's) associated with the safe 
storage of fuel or radioactive material. Nor do any of these changes 
increase the likelihood of the malfunction of an SSC. The proposed 
changes do not affect operating procedures or administrative 
controls that have the function of preventing or mitigating any 
design basis accidents.
    The MY DSAR provides a discussion of radiological events 
postulated to occur as a result of decommissioning with the bounding 
consequence resulting from a materials handling event. The proposed 
changes do not have an adverse impact on decommissioning activities 
or any of their postulated consequences.
    In addition, the proposed Technical Specification changes are 
consistent with the guidance provided in NRC Administrative Letter 
95-06. Therefore, these proposed changes do not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.

[[Page 5205]]

    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 License changes delete License information that 
does not apply to a plant that has permanently ceased operation. 
These changes are in compliance with 10 CFR Part 50 regulations and 
are not associated with any accidents previously evaluated.
    These proposed Technical Specification changes relocate 
requirements from the Technical Specifications to the Defueled 
Safety Analysis Report, eliminate Technical Specifications 
associated with the storage of spent fuel in the SFP, and relocate 
Technical Administrative Controls to the MY Quality Assurance 
Program. With the complete removal of spent fuel assemblies from the 
plant there are no safety related SSC's that remain at the plant. 
Thus, these proposed changes will not have any affect on the 
operation or design function of safety related SSC's. These changes 
do not create new component failure mechanisms, malfunctions or 
accident initiators. Therefore, these proposed changes would not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed License changes delete License information that 
does not apply to a plant that has permanently ceased operation. 
These changes are in compliance with 10 CFR Part 50 regulations and 
do not involve a reduction in a margin of safety.
    The design basis and accident assumptions within the MY DSAR and 
the Defueled Technical Specifications relating to spent fuel are no 
longer applicable. The proposed Technical Specification changes do 
not affect remaining plant operations, systems, or components 
supporting decommissioning activities. In addition, the proposed 
changes do not result in a change in initial conditions, system 
response time, or in any other parameter affecting the course of a 
decommissioning activity accident analysis.
    The relocation of the specified minimum distance to the 
Exclusion Area Boundary from the Technical Specifications to the 
Defueled Safety Analysis Report is consistent with the criterion set 
forth in 10 CFR 50.36 (c)(4). This criterion states that design 
features to be included in the Technical Specifications are those 
features of the facility such as materials of construction and 
geometric arrangement, which if altered or modified, would have a 
significant effect on safety and are not covered in other Technical 
Specification categories. The minimum distance to the Exclusion Area 
Boundary is established to maintain compliance within the limits 
specified in 10 CFR Part 100. The relocation of the specified 
minimum distance to the Exclusion Area Boundary to the DSAR 
continues to provide the safety analysis controls to assure 
compliance with 10 CFR Part 100 regulation.

Conclusion

    Based on the above, Maine Yankee 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 
requested amendment involves no significant hazards consideration.
    Attorney for licensee: Joe Fay, Esquire, Maine Yankee Atomic Power 
Company, 321 Old Ferry Road, Wiscasset, Maine 04578.
    NRC Section Chief: Claudia M. Craig.

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: December 19, 2003.
    Description of amendment requests: The proposed license amendment 
request (LAR) would revise Technical Specification (TS) 5.5.9, ``Steam 
Generator (SG) Tube Surveillance Program.'' The LAR proposes new SG 
wedge region exclusion zones for outside diameter stress corrosion 
cracking (ODSCC) alternate repair criteria (ARC) at tube support plate 
(TSP) intersections and for primary water stress corrosion cracking 
(PWSCC) ARC at dented TSP intersections. The wedge region exclusion 
zones currently approved for the ODSCC ARC and for the PWSCC ARC are 
based on a loss-of-coolant accident (LOCA) plus safe shutdown 
earthquake (SSE) loads analysis performed in 1992. The new wedge region 
exclusion zones are based on new analyses of LOCA plus SSE loads 
completed in 2003 using plant-specific accident loads.
    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.
    Application of a smaller steam generator (SG) wedge region 
exclusion zone will allow more degraded tubes to remain in service 
under alternate repair criteria (ARC). Previously approved ARC limits 
will be applied to tubes outside the exclusion zone, and therefore the 
probability and consequences of tube burst or leakage is not 
significantly increased following a steam line break (SLB).
    Exclusion zones tubes are inspected by bobbin coil every outage and 
by rotating pancake coil (RPC) if the bobbin coil detects degradation. 
SG tubes containing RPC-confirmed crack-like degradation at wedge 
region exclusion zone intersections will be repaired. Because in-
service tube intersections in wedge region exclusion zones do not have 
detectable cracking, they will not be susceptible to in-leakage if 
deformed following a loss-of-coolant-accident (LOCA) plus seismic 
event. Therefore, the consequences of a LOCA plus seismic event are not 
increased.
    Thus, the proposed changes do 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.
    Implementation of revised SG ARC wedge region exclusion zones will 
allow more degraded tubes to remain in service under ARC. 
Implementation of ARC has been previously approved and does not 
introduce any significant change to the plant design basis. A single or 
multiple tube rupture event would not be expected in a SG in which ARC 
has been applied.
    Therefore, the proposed changes do 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.
    Revised wedge region exclusion zones are based on a DCPP-specific 
analysis for the combined effects of a LOCA and safe shutdown 
earthquake (SSE) loads. The number of wedge region tubes that are 
predicted to deform has been decreased when compared to the prior 
analysis, which used highly conservative assumptions. The revised 
analysis incorporates DCPP-specific LOCA and seismic loads that were 
not available when the prior analysis was performed. The revised 
analysis also yields conservative results, such that the number of 
tubes in the exclusion zone (262 per SG) bound the number of tubes 
predicted to deform (120 per SG). Tubes located in the revised wedge 
region exclusion zone will continue to be subject to enhanced eddy 
current inspection requirements and will be excluded from application 
of ARC. Thus, existing tube integrity requirements continue to be met 
for

[[Page 5206]]

these tubes and there is no change to the dose contribution from tube 
leakage. Offsite and control room doses will continue to meet the 
appropriate guidelines and regulations established in Standard Review 
Plan 15.1.5 and 6.4, 10 CFR part 100, and 10 CFR part 50, Appendix A 
General Design Criterion (GDC) 19.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment 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.

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: January 7, 2004.
    Description of amendment requests: The license amendment request 
(LAR) would revise Sections 5.5.9, ``Steam Generator (SG) Tube 
Surveillance Program,'' and 5.6.10, ``Steam Generator (SG) Tube 
Inspection Report'' of the Diablo Canyon Technical Specifications. The 
proposed changes would allow application of a 4-volt alternate repair 
criteria (ARC) in intersections of steam generator (SG) tube hot-legs 
with the four lowest SG tube support plates (TSPs). The 4-volt ARC will 
only apply to Model 51 SG tubes experiencing outside diameter stress 
corrosion cracking (ODSCC) at the intersections of the tube hot legs 
and the four lowest TSPs. In addition, the proposed change includes the 
application of leak-before-break (LBB) to the main steam line (MSL) 
piping inside containment in order to exclude the dynamic effects of a 
main steam line break (MSLB) in the short length of piping upstream of 
the MSL flow restrictor (large MSLB) from consideration for determining 
the loads on the SG TSPs following an MSLB.
    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.
    A 4-volt steam generator (SG) bobbin coil probe voltage-based 
alternate repair criteria (ARC) for axial outside diameter stress 
corrosion cracking (ODSCC) at tube support plate (TSP) locations is 
proposed for the hot-leg region of the SG tube at the 4 lowest TSP 
locations (TSPs 1 through 4). In order to implement the proposed 4-volt 
ARC, sufficient SG tubes will be expanded in the hot-leg region of TSPs 
1 through 4 to limit the TSP deflections following a limiting main 
steam line break (MSLB) event.
    SG tubes pass through holes drilled in the TSP. The inside diameter 
of the drilled holes closely approximates the outside diameter of the 
tubes. Generally, the TSP precludes those tube spans within the drilled 
holes from deforming beyond the diameters of the drilled holes, thus, 
precluding tube burst in the restrained regions. However, design basis 
MSLB events may vertically displace a TSP, removing its support from 
the tube spans passing through it. For TSPs at hot-leg locations in 
which sufficient SG tubes have been expanded at the TSP intersection, 
the deflections of the TSP following a limiting MSLB event are small, 
the TSPs remain essentially stationary during all conditions, and the 
SG tube spans within the drilled TSP holes are restrained. Thus, for 
intersections of SG tube hot-legs and TSPs 1 through 4, axial tube 
burst is eliminated as a credible event and the larger bobbin voltage 
for the proposed 4-volt ARC can be allowed while still meeting the tube 
structural requirements of Regulatory Guide (RG) 1.121 [, ``Bases for 
Plugging Degraded PWR Steam Generator Tubes''].
    For the calculated displacement of the affected TSPs following a 
limiting design basis MSLB event, based on application of leak-before-
break to the main steam system piping inside containment, tube hot-leg 
spans enclosed within TSPs 1 through 4 have a tube burst probability of 
much less than 10-5 collectively. This is orders of magnitude less than 
the 10-2 probability-of-burst criterion specified by Generic Letter 
(GL) 95-05 [, ``Voltage-Based Repair Criteria for Westinghouse Steam 
Generator Tubes Affected by Outside Diameter Stress Corrosion 
Cracking,] and represents negligible axial tube burst probabilities for 
affected tube hot-leg spans intersecting TSPs. Thus, repair limits to 
preclude burst are not needed and tube repair limits for intersections 
of SG tube hot-legs and TSPs 1 through 4 may be based primarily on 
limiting leakage to acceptable levels during accident conditions.
    Cracks that include cellular corrosion may yield under axial loads, 
resulting in tensile tearing of the tube at that location. A tensile 
load requirement to prevent this establishes a structural limit for the 
tube expansion based ODSCC ARC. In order to establish a lower bound for 
the structural limit, tensile tests were used to measure the force 
required to separate a tube that exhibits cellular corrosion. 
Additionally, pulled SG tubes with cellular and/or inter-granular 
attack (IGA) tube wall degradation were evaluated and the tensile 
strength of the tube was conservatively calculated from the remaining 
noncorroded cross-section of the tube. The tensile strength calculation 
assumed that the degraded portions do not contribute to the axial load 
carrying ability of the tube. Data from these tests shows that 
circumferential cracks exhibiting bobbin coil probe indication voltages 
greater than 100 volts at the lower 95 percent confidence level require 
tube pressure differentials above the operating limit of 3-times normal 
operating differential pressure in order to produce circumferential 
ruptures (i.e., axial separation at the plane of the crack due to axial 
tensile tearing). The proposed 4-volt ARC has a safety factor of 25 to 
circumferential ruptures, which ensures the 4-volt ARC does not 
significantly increase the chances of a steam generator tube rupture 
(SGTR) at intersections of SG tube hot-legs and TSPs 1 through 4.
    In addressing the potential combined Loss-of-coolant accident 
(LOCA) and earthquake effects on SG components as required by General 
Design Criterion (GDC) 2 of Appendix A to 10 CFR part 50, analysis has 
shown that SG tube deformation or collapse may occur in certain regions 
of the SG. SG tube collapse reduces RCS flow and could cause partial 
through-wall tube cracks to become full through-wall tube-cracks during 
tube deformation or collapse resulting in potential secondary-to-
primary in-leakage to the reactor coolant system (RCS). Tubes for which 
deformation may occur are excluded from application of the voltage-
based ARC per current TS 5.5.9.d.1.j (iv). TS 5.5.9.d.1.j (iv) will 
continue to apply and is not adversely affected by the 4-volt ARC. 
Therefore tubes for which deformation may occur will not be left in 
service under the 4-volt ARC.
    GL 95-05 states that licensees must perform SG tube postaccident 
leak rate and SG tube burst probability analyses before returning to 
power from outages during which they perform SG

[[Page 5207]]

inspections. Licensees must include the results in a report to the NRC 
within 90 days after restart. If an analysis reveals that postaccident 
leak-rate or burst-probability exceeds limits, the licensee must report 
it to the NRC and assess the safety significance of this finding.
    For the proposed 4-volt ARC, the axial tensile tearing tube rupture 
probability is calculated for indications found at intersections of 
tube hot-legs and TSPs 1 through 4. The sum of MSLB axial tube burst 
probability for cold-leg TSP intersections, MSLB axial tube burst 
probability for hot-leg intersections at TSPs 5 through 7, axial 
tensile tearing tube rupture probability for TSPs 1 through 4, and the 
burst probability for indications left in service under other ARCs must 
be compared to the GL 95-05 reporting value of 10-\2\. Due 
to the negligible burst probability for axial ODSCC indications at 
intersections of tube hot-legs and TSPs 1 through 4, calculation of the 
axial burst probability is not required for these indications.
    The design basis MSLB outside of containment produces the limiting 
radiological consequence from any SG tube leakage due to SG tube 
indications that are postulated to exist at the initiation of an 
accident. Verification prior to each operating cycle, that the sum of 
MSLB leak rates from indications left in service under all ARC 
(including the proposed 4-volt ARC) are less than the leak rate limit 
assumed in the MSLB radiological consequences analysis, will ensure 
that site boundary doses for this accident remain within an acceptable 
fraction of the guidelines of Title 10 of the Code of Federal 
Regulations, Part 100, (10 CFR part 100) and that doses to the control 
room operators remain within the 10 CFR part 50, Appendix A GDC 19 
limits.
    The application of leak-before-break (LBB) to the MSL piping inside 
containment does not alter the way in which plant equipment is operated 
and cannot initiate an accident. The application of LBB to the main 
steam system does not affect the plant operating conditions and will 
not challenge the ability of the main steam system to perform its 
design function or to mitigate an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Use of the proposed SG tube 4-volt ARC does not significantly 
change the operating conditions of the SG. Application of the 4-volt 
ARC does not significantly increase the probability of either single or 
multiple tube ruptures. SG tube integrity remains adequate for all 
plant operating conditions. The GL 95-05 SG tube integrity limits will 
be confirmed through in-service inspection and monitoring of primary-
to-secondary leakage.
    The Diablo Canyon Units 1 and 2 Technical Specifications (TS) 
impose a normal SG primary-to-secondary leak rate limit of 150 gallons 
per day (gpd) per SG to minimize the potential for excessive leakage 
during all plant conditions. The 150 gpd limit provides added margin to 
accommodate contingent leakage should a stress corrosion crack grow at 
a greater than expected rate or extend outside the TSP. The proposed 4-
volt ARC does not adversely impact the TS 150 gpd limit. Normal 
operating leakage is not expected to significantly increase due to 
indications left in service under the proposed 4-volt ARC.
    The application of LBB to the MSL piping inside containment does 
not involve any physical alteration to the plant or any change in which 
the plant is operated which could introduce a new failure mode. The use 
of LBB does not involve plant equipment being operated in a different 
manner.
    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.
    RG 1.121 describes a method for meeting GDCs 14, 15, 31, and 32 of 
Appendix A to 10 CFR 50 by reducing the probability or consequences of 
SGTR through application of criteria for removing degraded tubes from 
service. These criteria set limits of degradation for SG tubing through 
inservice inspection. Analyses show that tube integrity will continue 
to meet the criteria of Regulatory Guide 1.121 after implementation of 
the proposed 4-volt ARC. Even under the worst case ODSCC occurrence at 
TSP elevations left in service under the 4-volt ARC, the 4-volt ARC 
will not cause or significantly increase the probability of a SGTR 
event.
    Verification prior to each operating cycle, that the sum of MSLB 
leak rates from indications left in service under all ARC (including 
the proposed 4-volt ARC) are less than the leak rate limit assumed in 
the MSLB radiological consequences analysis, will ensure that site 
boundary doses for this accident remain within an acceptable fraction 
of the guidelines of 10 CFR 100 and that doses to the control room 
operators remain within the limits of GDC 19 of Appendix A to 10 CFR 
50.
    Inspections conducted for the proposed 4-volt ARC are the same as 
required by GL 95-05 with adjustment of the rotating pancake coil 
inspection requirements for hot-leg TSPs 1 through 4 intersections to 
reflect the higher 4-volt ARC limit. All hot-leg TSPs 1 through 4 
intersections with bobbin coil voltages greater than 4 volts will be 
inspected with Plus Point coil and a Plus Point coil minimum sample 
inspection of intersections with bobbin indications less than or equal 
to 4 volts will be applied to hot-leg TSPs 1 through 4. The Plus Point 
coil data will be evaluated to confirm that the principal degradation 
mechanism continues to be ODSCC.
    Plugging SG tubes reduces RCS flow margin. The 4-volt ARC will 
reduce the number of tubes that must be plugged. Thus, the 4-volt ARC 
will conserve RCS flow margin, preserving operational and safety 
benefits that would otherwise be reduced by unnecessary plugging.
    The application of LBB to the MSL piping inside containment will 
not adversely affect operation of plant equipment and will not result 
in a change to design basis accident initial conditions or the 
setpoints at which protective actions are initiated. With application 
of LBB, the main steam system will continue to perform its function as 
assumed in the accident analyses.
    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: Stephen Dembek.

Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear 
Plant Unit 1, Limestone County, Alabama

    Date of amendment request: November 3, 2003.
    Description of amendment request: The proposed amendment would 
lower the allowable value for Function 7.b, Scram Discharge Volume 
Water Level--High Float Switches in Technical Specification (TS) Table 
3.3.1.1-1, Reactor Protection System Instrumentation. As part of the 
proposed change, the licensee would

[[Page 5208]]

also remove the Low Scram Pilot Air Header Pressure switches from 
service.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. Modifications to the Scram Discharge Instrument Volume 
(SDIV) System are being implemented to ensure that the SDIV high 
water level instrumentation will respond adequately to provide 
redundant, diverse trip functions for a Scram Discharge Volume (SDV) 
inleakage event. The proposed change does not involve any change to 
the design or functional requirements of plant systems and the 
surveillance test methods will be unchanged. The proposed change 
will not give rise to any increase in operating power level, fuel 
operating limits, or effluents. The proposed change does not affect 
any accident precursors. In addition, the proposed change will not 
significantly increase any radiation levels. Since the scram 
function will be successfully performed, lowering the allowable 
value for the Scram Discharge Volume Water Level--High Float 
Switches and removal of the scram pilot air header pressure trip 
system does not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The design criteria for the Scram Discharge System is 
contained in the Safety Evaluation Report on the Boiling Water 
Reactor (BWR) Scram Discharge System, which was transmitted by NRC 
letter dated December 9, 1980, to All BWR Licensees. Modifications 
to the SDV System have been evaluated to demonstrate that the high 
water level instrumentation in the SDIV will respond adequately to 
provide the required trip function. No new system failure modes are 
created as a result of removing the low scram pilot air header trip, 
since the redundant and diverse SDIV high water level instruments 
will initiate a successful reactor scram. Therefore, lowering the 
allowable value for the Scram Discharge Volume Water Level--High 
Float Switches and removal of the scram pilot air header pressure 
trip system does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The water level in the SDIV is monitored by both resistance-
temperature type detectors and float switches. Redundancy and 
diversity in the instrumentation that initiates the scram signal is 
maintained even with the lowering of the allowable value for the 
Scram Discharge Volume Water Level--High Float Switches and removal 
of the low scram pilot air header pressure trip function. 
Modifications to the SDIV System have been evaluated to demonstrate 
that the high water level instrumentation will respond adequately to 
provide the required trip function for an inleakage event. 
Therefore, the proposed amendment does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear 
Plant Unit 1, Limestone County, Alabama

    Date of amendment request: November 10, 2003.
    Description of amendment request: The proposed amendment includes 
the necessary Technical Specification (TS) changes for the planned 
replacement of the power range monitoring portion of the existing 
Neutron Monitoring System with a digital upgrade. These changes would 
expand the current allowable operating domain to the Maximum Extended 
Load Line Limit region of the power/flow chart.
    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.
    Power Range Neutron Monitor (PRNM) Changes:
    The proposed TS changes are associated with the Nuclear 
Measurement Analysis and Control (NUMAC) PRNM retrofit design. The 
proposed changes involve modification of the Limiting Conditions for 
Operation (LCOs) and Surveillance Requirements for equipment 
designed to mitigate events which result in power increase 
transients. For the Average Power Range Monitor (APRM) system, the 
mitigating action is to block control rod withdrawal or initiate a 
reactor scram which terminates the power increase when setpoints are 
exceeded. For the Rod Block Monitor (RBM) system, the mitigating 
action is to block continuous control rod withdrawal prior to 
exceeding the Minimum Critical Power Ratio safety limit during a 
postulated Rod Withdrawal Error event. The worst case failure of 
either the APRM or the RBM systems is failure to initiate its 
mitigating action (failure to scram or block rod withdrawal). 
Failure to initiate these mitigating actions will not increase the 
probability of an accident. Thus, the proposed changes do not 
increase the probability of an accident previously evaluated.
    For the APRM and the RBM systems, the NUMAC PRNM design, 
together with revised operability requirements and revised 
surveillance requirements, results in equipment which continues to 
perform the same mitigation functions conditions with reliability 
equal to or greater than the equipment which it replaces. Because 
there is no change in mitigation functions and because reliability 
of the functions is maintained, the proposed changes do not involve 
an increase in the consequences of an accident previously evaluated.
    APRM and RBM Technical Specification (ARTS) improvements and 
operation in an expanded core power/flow domain, the Maximum 
Extended Load Line Limit (MELLL) Changes:
    The proposed ARTS/MELLL changes permit expansion of the current 
allowable power/flow operating region and will apply a newer 
methodology for assuring that fuel thermal and mechanical design 
limits are satisfied. Operation in the MELLL region with the ARTS 
changes has been evaluated and there is adequate design margin for 
operation in the MELLL region for all events and parameters 
considered. Because operation in the MELLL region maintains adequate 
design margin, the proposed changes do not increase the probability 
of an accident previously evaluated.
    In support of operation in the MELLL region, the proposed change 
modifies flow-biased APRM scram and rod block setpoints and 
implements new RBM power-biased setpoints. No direct credit for the 
flow-biased APRM scram or APRM flow-biased rod block is taken in 
mitigation of any design basis event. Therefore, design margins are 
not degraded by the proposed changes.
    The proposed changes to the RBM system will assure that a Rod 
Withdrawal Error is not a limiting event and that the RBM continues 
to enforce rod blocks under appropriate conditions.
    Therefore, the proposed changes do not increase the probability 
or the 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 PRNM and ARTS/MELLL changes involve modification 
and replacement of the existing power range neutron monitoring 
equipment, modification of the setpoints and operational 
requirements for the APRM and RBM systems, implementation of a new 
methodology for administering compliance with fuel thermal limits, 
and operation in an extended power/flow domain. These proposed 
changes do not modify the basic functional requirements of the 
affected equipment, create any new system interfaces or 
interactions, nor create any new system failure modes or sequence of

[[Page 5209]]

events that could lead to an accident. The worst case failure of the 
affected equipment is failure to perform a mitigation action, and 
failure of this equipment to perform a mitigating action does not 
create the possibility of a new or different kind of accident. No 
new external threats or release pathways are created. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    PRNM Changes: These proposed TS changes are associated with the 
NUMAC PRNM retrofit design. The NUMAC PRNM change does not impact 
reactor operating parameters or the functional requirements of the 
PRNM system. The replacement equipment continues to provide 
information, enforce control rod blocks, and initiate reactor scrams 
under appropriate specified conditions. The proposed change does not 
revise any safety margin requirements. The replacement APRM/RBM 
equipment has improved channel trip accuracy compared to the current 
system, and meets or exceeds system requirements previously assumed 
in setpoint analysis. Thus, the ability of the new equipment to 
enforce compliance with margins of safety equals or exceeds the 
ability of the equipment which it replaces. Therefore, the proposed 
changes do not involve a reduction in a margin of safety.
    ARTS/MELLL Changes: Operation in the MELLL region does not 
affect the ability of the plant safety-related trips or equipment to 
perform their functions, nor does it cause any significant increase 
in offsite radiation doses resulting from any analyzed event. 
Analyses have demonstrated that, for operation in the MELLL region, 
adequate margin to design limits is maintained. Implementation of 
the ARTS improvements provides flow- and power-dependent thermal 
limits which maintain existing margins of safety in normal 
operation, anticipated operational occurrences, and accident events. 
Implementation of power-biased RBM setpoints improves the margin of 
safety in a postulated Rod Withdraw Error (RWE) by assuring that the 
RWE is not a limiting event. Thus, the proposed changes do not 
involve a 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: December 18, 2003.
    Brief description of amendments: The proposed amendments would 
revise Technical Specification (TS) 3.9.6, ``Residual Heat Removal 
(RHR) and Coolant Circulation Low Water Level'' to correct completion 
times for Actions B.2 and B.3. Action B.2 should have a completion time 
of immediately and Action B.3 should have a completion time of 4 hours.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR), Section 50.91(a), 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 proposed change is considered to be a correction of an 
editorial error. The proposed revision to TS 3.9.6 is consistent 
with the current CPSES [Comanche Peak Steam Electric Station] 
licensing basis. Therefore the proposed change does 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 change is considered to be an editorial correction 
and does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change is considered to be an editorial correction 
and does not involve a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: December 31, 2003.
    Brief description of amendments: The proposed amendments would 
revise Technical Specification (TS) 3.8.1, ``AC Sources--Operating'' to 
extend the allowable completion times for the required actions 
associated with restoration of an inoperable diesel generator (DG) and 
an inoperable offsite circuit (i.e., startup transformer). The proposed 
amendments will also revise TS 3.8.9, ``Distribution Systems--
Operating'' to extend the allowable completion times for the required 
actions associated with restoration of an inoperable alternating 
current (AC) electrical power distribution system (i.e., 6.9 kV AC 
safety bus).
    Basis for proposed no significant hazards consideration 
determination: As required by title 10 of the Code of Federal 
Regulations (10 CFR), Section 50.91(a), 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 proposed Technical Specification changes do not 
significantly increase the probability of occurrence of a previously 
evaluated accident because the 6.9 kV AC components (i.e., Diesel 
Generators (DGs), startup transformers (STs), and safety-related 
(Class 1E) busses) are not initiators of previously evaluated 
accidents involving a loss of offsite power. The proposed changes to 
the Technical Specification Action Completion Times do not affect 
any of the assumptions used in the deterministic or the 
Probabilistic Safety Assessment (PSA) analysis[.]
    The proposed Technical Specification changes will continue to 
ensure the 6.9 kV AC components perform their function when called 
upon. Extending the Technical Specification Completion Times to 10 
days does not affect the design of the DGs, the operational 
characteristics of the DGs, the interfaces between the DGs and other 
plant systems, the function, or the reliability of the DGs. Thus, 
the DGs will be capable of performing either accident mitigation 
function and there is no impact to the radiological consequences of 
any accident analysis. To fully evaluate the effect of the changes 
to the 6.9 kV AC components, Probabilistic Safety Analysis (PSA) 
methods and deterministic analysis were utilized. The results of 
this analysis show no significant increase in the Core Damage 
Frequency.
    The Configuration Risk Management Program (CRMP) in Technical 
Specification 5.5.18 is an administrative program that

[[Page 5210]]

assesses risk based on plant status. Adding the requirement to 
implement the CRMP for Technical Specification 3.8.1 and 3.8.9 
requires the consideration of other measures to mitigate 
consequences of an accident occurring while a 6.9 kV AC component is 
inoperable.
    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 changes do not result in a change in the manner in 
which the electrical distribution subsystems provide plant 
protection. There are no design changes associated with the proposed 
changes. The changes to Completion Times do not change any existing 
accident scenarios, nor create any new or different accident 
scenarios.
    The changes do not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. In addition, 
the changes do not impose any new or different requirements or 
eliminate any existing requirements. The changes do not alter 
assumptions made in the safety analysis. The proposed changes are 
consistent with the safety analysis assumptions and current plant 
operating practice.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by these changes. The proposed changes will not 
result in plant operation in a configuration outside the design 
basis. The calculated impact on risk is insignificant and is 
consistent with the acceptance criteria contained in Regulatory 
Guides 1.174 and 1.177. The proposed activities involve[ ] changes 
to certain Completion Times. The proposed changes remain bounded by 
the existing Surveillance Requirement Completion Times and therefore 
have no impact to the margins of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

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

    Date of application request: June 27, 2003, as revised by letter 
dated December 19, 2003, (previously published in the Federal Register 
on July 22, 2003 [68 FR 43396]).
    Description of amendment request: The licensee's letter dated 
December 19, 2003, revises the original amendment application dated 
June 27, 2003. The original amendment request was described as that 
which would revise the Technical Specifications (TSs) to (1) extend the 
allowed outage time (AOT) or required action completion time (CT) for 
an inoperable diesel generator (DG) by adding the phrase ``OR 108 hours 
once per cycle for each DG'' to the completion time for Required Action 
B.4 in TS 3.8.1, ``AC Sources-Operating,'' and (2) delete the second CT 
given in certain required actions in TS 3.6.6, ``Containment Spray and 
Cooling Systems;'' TS 3.7.5, ``Auxiliary Feedwater (AFW) System;'' TS 
3.8.1; and TS 3.8.9, ``Distribution System--Operating.'' The revised 
application dated December 19, 2003, requests changes to only Required 
Actions A.3 and B.4 for TS 3.8.1 to extend the AOT, or required action 
CT, for an inoperable DG.
    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 for increasing the ``second'' Completion 
Times under TS 3.8.1 do not affect the design, operational 
characteristics, or intended functions of the equipment addressed by 
TS 3.8.1. With no direct effects on the subject equipment (or any 
other plant equipment or features), the proposed ``second'' 
Completion Time changes are not associated with any initiating 
condition for any accident previously evaluated, and therefore would 
not affect the probability of such accidents. Further, the 
consequences of evaluated accidents are independent of mitigating 
equipment allowed outage times as long as adequate availability of 
the equipment is ensured.
    ``Second'' Completion Times are primarily administrative in 
nature and are only intended to prevent successive, overlapping or 
contiguous entries and exits from Conditions within a Technical 
Specification LCO [Limiting Condition for Operation], which could 
otherwise result in an extended period of time for which the LCO is 
not met. The new, extended ``second'' Completion Times preserve this 
intent and were determined by the same method used to establish the 
original/existing second Completion Time limits, albeit with a 
longer, risk-informed Completion Time established for an inoperable 
diesel generator.
    The proposed changes to the ``second'' Completion Times of TS 
3.8.1 support the extended Completion Time/AOT specified for an 
inoperable diesel generator as proposed in AmerenUE's June 27, 2003 
amendment application (Reference 1 [in AmerenUE's revised 
application dated December 19, 2003]). The acceptability and 
conformance to regulatory guidance for that change is addressed in 
[AmerenUE's June 27, 2003 amendment application], and the 
conclusions reached therein, including those reached with respect to 
significant hazards consideration, remain unchanged for that 
proposed change.
    Therefore, the proposed changes do 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 changes are primarily administrative in nature and 
do not involve a change in the design, configuration, or operational 
characteristics of the plant.
    No physical alteration of the plant is involved, as no new or 
different type of equipment is to be installed. The changes do not 
alter any assumptions made in the safety analyses, and no alteration 
in the procedures for ensuring that the plant remains within 
analyzed limits is involved. As such, no new failure modes or 
mechanisms that could cause a new or different kind of accident from 
any previously evaluated are being introduced.
    Therefore, the proposed changes do 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 proposed changes to the affected second Completion Times do 
not alter the manner in which safety limits or limiting safety 
system settings are determined. The safety analysis acceptance 
criteria are not impacted by [these] change[s], and the proposed 
changes will not permit plant operation in a configuration [that is] 
outside the design basis.
    The proposed, extended second Completion Time limits were 
established in the same manner as the original limits, and meet the 
same intent, except that a longer risk-informed DG AOT has been used 
to establish the proposed second Completion Time limits[.] The basis 
and acceptability of that time limit is addressed in the June 27, 
2003 amendment application (as supported by this supplemental/
revision[, dated December 19, 2003]), and the conclusions reached 
therein still apply, including those reached with respect to [no] 
significant hazards consideration. [The June 27, 2003, amendment 
application stated: ``Further, with regard to plant risk, the risk 
assessment performed for the DG AOT extension determined that the 
quantifiable increase in plant risk is acceptably small.'']
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are

[[Page 5211]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Section Chief: Stephen Dembek.

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

    Date of application request: December 15, 2003.
    Description of amendment request: The amendment would revise 
Technical Specifications (TSs) 3.3.9, ``Boron Dilution Mitigation 
System (BDMS),'' and 3.9.2, ``Unborated Water Source Isolation 
Valves.'' The proposed changes would replace the phrase ``unborated 
water'' by the word ``dilution'' in several places and delete 
references to isolation valves BGV0178 and BGV0601. A Note would also 
be added to TS 3.9.2 about dilution source path valves may be 
unisolated.
    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 do not involve a significant increase in 
the probability or consequences of an inadvertent boron dilution 
accident by isolating the BTRS [boron thermal regeneration system] 
anion resin vessels in MODE 6 or by isolating the purge line for 
detector SJRE001 during flushing activities in MODE 6. By 
recognizing these potential dilution sources and by making TS 3.3.9 
and TS 3.9.2 more generic for consideration of all potential 
dilution sources, plant administrative controls are revised such 
that the plant is put in a safer condition than before. Specific 
isolation [valve numbers] are removed from TS 3.3.9 and TS 3.9.2. 
They are relocated from the [Technical] Specifications to the 
appropriate TS Bases. This is an administrative only change and is 
consistent with the [Improved] Standard Technical Specifications, 
NUREG-1431[, that the Callaway Technical Specifications are based 
upon]. Allowing a dilution source path to be unisolated under 
administrative controls, described in TS Bases 3.9.1 during 
refueling decontamination activities, is acceptable as allowed by 
Amendment [No.] 97 to the Callaway Operating License and does not 
involve a significant increase in the probability or consequences of 
an inadvertent boron dilution accident.
    Therefore, the proposed changes do 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 changes do not create the possibility of a new or 
different kind of accident. Although other potential dilution 
sources are identified for administrative control, the evaluation of 
a MODE 6 dilution event remains unchanged. Isolating the BTRS anion 
vessels or isolating the purge line for detector SJRE001 during 
flushing activities in MODE 6 and making TS 3.3.9 and TS 3.9.2 more 
generic does not impact the operability of any safety related 
equipment required for plant operation. No new equipment will be 
added and no new limiting single failures are created. The plant 
will continue to be operated within the envelope of the existing 
safety analysis. In addition specific isolation [valve numbers] are 
removed from TS 3.3.9 and TS 3.9.2. They are relocated from the 
[Technical] Specifications to the appropriate TS Bases. This is an 
administrative only change and is consistent with the [Improved] 
Standard Technical Specifications, NUREG-1431 [, that the Callaway 
Technical Specifications are based upon]. Allowing a dilution source 
path to be unisolated under administrative controls, described in TS 
Bases 3.9.1 during refueling decontamination activities, is 
acceptable as allowed by Amendment [No.] 97 to the Callaway 
Operating License and does not create the possibility of a new or 
different kind of inadvertent boron dilution accident.
    Therefore, the proposed changes do 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 changes do not reduce the margin of safety. 
Although other potential dilution sources are identified for 
administrative control and TS 3.3.9 and TS 3.9.2 are made more 
generic for consideration of all potential dilution sources, the 
evaluated margin of safety for a dilution event in MODE 6 remains 
the same. Recognition of other potential dilution sources, isolation 
of the BTRS anion resin beds and the purge line for detector SJRE001 
during flushing activities in MODE 6, places the plant in a safer 
condition than before. In addition specific isolation [valve 
numbers] are removed from TS 3.3.9 and TS 3.9.2. They are relocated 
from the [Technical] Specifications to the appropriate TS Bases. 
This is an administrative only change and is consistent with the 
[Improved] Standard Technical Specifications, NUREG-1431 [, that the 
Callaway Technical Specifications are based upon]. Finally, allowing 
a dilution source path to be unisolated under administrative 
controls, described in TS Bases 3.9.1 during refueling 
decontamination activities, is acceptable as allowed by Amendment 
[No.] 97 to the Callaway Operating License and does not involve a 
significant reduction in a margin of safety due to an inadvertent 
boron dilution event.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of application request: December 17, 2003.
    Description of amendment request: The amendment would revise 
Technical Specifications (TSs) 3.3.1, ``Reactor Trip System (RTS) 
Instrumentation,'' 3.3.2, ``Engineered Safety Feature Actuation System 
(ESFAS) Instrumentation,'' and 3.3.9, ``Boron Dilution Mitigation 
System (BDMS).'' The purpose of the amendment is to adopt the 
completion time, test bypass time, and surveillance frequency time 
changes approved by the NRC in Topical Reports WCAP-14333-P-A, 
``Probabilistic Risk Analysis of the RPS [reactor protection system] 
and ESFAS Test Times and Completion Times,'' and WCAP-15376-P-A, 
``Risk-Informed Assessment of the RTS and ESFAS Surveillance Test 
Intervals and Reactor Trip Breaker Test and Completion Times.'' The 
proposed changes would revise the required actions for certain action 
conditions; increase the completion times for several required actions 
(including some notes); delete notes in certain required actions; 
increase frequency time intervals (including certain notes) in several 
surveillance requirements (SRs); add an action condition and required 
actions; revise notes in certain SRs; and revise Table 3.3.2-1. There 
are also several administrative corrections to the format of the TSs 
(e.g., moving the ``AND'' in the required actions for Condition O in TS 
3.3.2).
    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. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since no 
hardware changes are proposed. The same

[[Page 5212]]

reactor trip system (RTS) and engineered safety feature actuation 
system (ESFAS) instrumentation will continue to be used. The 
protection systems will continue to function in a manner consistent 
with the plant design basis. These changes to the Technical 
Specifications [in the amendment] do not result in a condition where 
the design, material, and construction standards that were 
applicable prior to the change are altered.
    The proposed changes will not modify any system interface. The 
proposed changes will not affect the probability of any event 
initiators [because the proposed changes are not event initiators]. 
There will be no degradation in the performance of or an increase in 
the number of challenges imposed on safety-related equipment assumed 
to function during an accident situation. There will be no change to 
normal plant operating parameters or accident mitigation 
performance. The proposed changes will not alter any assumptions or 
change any mitigation actions in the radiological consequence 
evaluations in the FSAR [Callaway Final Safety Analysis Report].
    The determination that the results of the proposed changes are 
acceptable [to be considered for plant-specific Technical 
Specifications] was established in the NRC Safety Evaluations 
prepared for WCAP-14333-P-A (issued by letter dated July 15, 1998) 
and for WCAP-15376-P-A (issued by letter dated December 20, 2002). 
Implementation of the proposed changes will result in an 
insignificant risk impact. Applicability of these conclusions has 
been verified through plant-specific reviews and implementation of 
the generic analysis results in accordance with the respective NRC 
Safety Evaluation conditions [for the two WCAPs].
    The proposed changes to the Completion Times, test bypass times, 
and Surveillance Frequencies reduce the potential for inadvertent 
reactor trips and spurious ESF [engineered safety feature] 
actuations, and therefore do not increase the probability of any 
accident previously evaluated. The proposed changes do not change 
the response of the plant to any accidents and have an insignificant 
impact on the reliability of the RTS and ESFAS signals. The RTS and 
ESFAS will remain highly reliable and the proposed changes will not 
result in a significant increase in the risk of plant operation. 
This is demonstrated by showing that the impact on plant safety as 
measured by the increase in core damage frequency (CDF) is less than 
1.0E-06 per year and the increase in large early release frequency 
(LERF) is less than 1.0E-07 per year. In addition, for the 
Completion Time changes, the incremental conditional core damage 
probabilities (ICCDP) and incremental conditional large early 
release probabilities (ICLERP) are less than 5.0E-07 and 5.0E-08, 
respectively. These changes meet the acceptance criteria in 
Regulatory Guides 1.174 and 1.177. Therefore, since the RTS and 
ESFAS will continue to perform their [safety] functions with high 
reliability as originally assumed, and the increase in risk as 
measured by [Delta]CDF, [Delta]LERF, ICCDP, ICLERP risk metrics is 
within the acceptance criteria of existing [NRC] regulatory 
guidance, there will not be a significant increase in the 
consequences of any accidents.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components (SSCs) 
from performing their intended [safety] function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed changes do not affect the source term, 
containment isolation, or radiological release assumptions used in 
evaluating the radiological consequences of an accident previously 
evaluated. The proposed changes are consistent with safety analysis 
assumptions and resultant consequences.
    Therefore, [the] change[s do] not increase 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.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. The proposed changes will not affect the normal method of 
plant operation. No performance requirements will be affected or 
eliminated. The proposed changes will not result in physical 
alteration to any plant system nor will there be any change in the 
method by which any safety-related plant system performs its safety 
function. There will be no setpoint changes or changes to accident 
analysis assumptions.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of these changes. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of these changes.
    Therefore, the proposed changes do 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 changes do not affect the acceptance criteria for 
any analyzed event nor is there a change to any Safety Analysis 
Limit (SAL). There will be no effect on the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined nor will there be any effect on those plant 
systems necessary to assure the accomplishment of protection 
functions. There will be no impact on the overpower limit, DNBR 
[departure from nucleate boiling ratio] limits, FQ [heat 
flux hot channel factor], F[Delta]H [nuclear enthalpy rise hot 
channel factor], LOCA PCT [loss-of-coolant accident peak cladding 
temperature], peak local power density, or any other margin of 
safety. The radiological dose consequence acceptance criteria listed 
in the [NRC] Standard Review Plan will continue to be met.
    Redundant RTS and ESFAS trains are maintained, and diversity 
with regard to the signals that provide reactor trip and engineered 
safety features actuation is also maintained. All signals credited 
as primary or secondary, and all operator actions credited in the 
accident analyses will remain the same. The proposed changes will 
not result in plant operation in a configuration outside the design 
basis. The calculated impact on risk is insignificant and meets the 
acceptance criteria contained in Regulatory Guides 1.174 and 1.177. 
Although there was no attempt to quantify any positive human factors 
benefit due to increased Completion Times and bypass test times, it 
is expected that there would be a net benefit due to a reduced 
potential for spurious reactor trips and actuations associated with 
testing.
    Implementation of the proposed changes is expected to result in 
an overall improvement in safety, as follows:
    (a) Reduced testing will result in fewer inadvertent reactor 
trips, less frequent actuation of ESFAS components, less frequent 
distraction of operations personnel without significantly affecting 
RTS and ESFAS reliability.
    (b) Improvements in the effectiveness of the operating staff in 
monitoring and controlling plant operation will be realized. This is 
due to less frequent distraction of the operators and shift 
supervisor to attend to instrumentation Required Actions with short 
Completion Times.
    (c) Longer repair times associated with increased Completion 
Times will lead to higher quality repairs and improved reliability.
    (d) The Completion Time extensions for the reactor trip breakers 
will provide the utilities additional time to complete test and 
maintenance activities while at power, potentially reducing the 
number of forced outages related to compliance with reactor trip 
breaker Completion Times, and provide consistency with the 
Completion Times for the logic trains.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: December 15, 2003.
    Description of amendment request: The licensee is proposing to 
revise Technical Specification (TS) Section 3.3.1, ``Reactor Trip 
System (RTS)

[[Page 5213]]

Instrumentation,'' and TS 3.3.2, ``Engineered Safety Feature Actuation 
System (ESFAS) Instrumentation,'' to adopt completion time, test bypass 
time (in Notes for several Required Actions), and surveillance 
frequency changes approved by the NRC in WCAP-14333-P-A, Revision 1, 
``Probabilistic Risk Analysis of the RPS and ESFAS Test Times and 
Completion Times,'' dated October 1998, and WCAP-15376-P-A, Revision 1, 
``Risk-Informed Assessment of the RTS and ESFAS Surveillance Test 
Intervals and Reactor Trip Breaker Test and Completion Times,'' dated 
March 2003.
    As part of this amendment, for TS 3.3.1, the Required Actions for 
Condition D, one power range neutron flux-high channel inoperable, are 
revised, and a Note for the Required Actions for Condition R is 
deleted.
    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.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since no 
hardware changes are proposed. The same reactor trip system (RTS) 
and engineered safety feature actuation system (ESFAS) 
instrumentation will continue to be used. The protection systems 
will continue to function in a manner consistent with the plant 
design basis. These changes to the Technical Specifications do not 
result in a condition where the design, material, and construction 
standards that were applicable prior to the change are altered.
    The proposed changes will not modify any system interface. The 
proposed changes will not affect the probability of any event 
initiators. There will be no degradation in the performance of or an 
increase in the number of challenges imposed on safety-related 
equipment assumed to function during an accident situation. There 
will be no change to normal plant operating parameters or accident 
mitigation performance. The proposed changes will not alter any 
assumptions or change any mitigation actions in the radiological 
consequence evaluations in the USAR [Updated Safety Analysis 
Report].
    The determination that the results of the proposed changes are 
acceptable was established in the NRC Safety Evaluations prepared 
for WCAP-14333-P-A (issued by letter dated July 15, 1998) and for 
WCAP-15376-P-A (issued by letter dated December 20, 2002). 
Implementation of the proposed changes will result in an 
insignificant risk impact. Applicability of these conclusions has 
been verified through plant-specific reviews and implementation of 
the generic analysis results in accordance with the respective NRC 
Safety Evaluation conditions.
    The proposed changes to the Completion Times, test bypass times, 
and Surveillance Frequencies reduce the potential for inadvertent 
reactor trips and spurious ESF [engineered safety feature] 
actuations, and therefore do not increase the probability of any 
accident previously evaluated. The proposed changes do not change 
the response of the plant to any accidents and have an insignificant 
impact on the reliability of the RTS and ESFAS signals. The RTS and 
ESFAS will remain highly reliable and the proposed changes will not 
result in a significant increase in the risk of plant operation. 
This is demonstrated by showing that the impact on plant safety as 
measured by the increase in core damage frequency (CDF) is less than 
1.0E-06 per year and the increase in [the] large early release 
frequency (LERF) is less than 1.0E-07 per year. In addition, for the 
Completion Time changes, the incremental conditional core damage 
probabilities (ICCDP) and incremental conditional large early 
release probabilities (ICLERP) are less than 5.0E-07 and 5.0E-08, 
respectively. These changes meet the acceptance criteria in 
Regulatory Guides 1.174 and 1.177. Therefore, since the RTS and 
ESFAS will continue to perform their functions with high reliability 
as originally assumed, and the increase in risk as measured by 
[Delta]CDF, [Delta]LERF, ICCDP, ICLERP risk metrics is within the 
acceptance criteria of existing regulatory guidance, there will not 
be a significant increase in the consequences of any accidents.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components (SSCs) 
from performing their intended function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
The proposed changes are consistent with safety analysis assumptions 
and resultant consequences.
    Therefore, [the proposed changes do] not increase 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.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. The proposed changes will not affect the normal method of 
plant operation. No performance requirements will be affected or 
eliminated. The proposed changes will not result in [a] physical 
alteration to any plant system nor will there be any change in the 
method by which any safety-related plant system performs its safety 
function. There will be no setpoint changes or changes to accident 
analysis assumptions.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of these changes. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of these changes.
    Therefore, the proposed changes do 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 changes do not affect the acceptance criteria for 
any analyzed event nor is there a change to any Safety Analysis 
Limit (SAL). There will be no effect on the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined nor will there be any effect on those plant 
systems necessary to assure the accomplishment of protection 
functions. There will be no impact on the overpower limit, DNBR 
limits, FQ, F[Delta]H, LOCA [loss-of-coolant accident] 
PCT [peak cladding temperature], peak local power density, or any 
other margin of safety. The radiological dose consequence acceptance 
criteria listed in the Standard Review Plan will continue to be met.
    Redundant RTS and ESFAS trains are maintained, and diversity 
with regard to the signals that provide reactor trip and engineered 
safety features actuation is also maintained. All signals credited 
as primary or secondary, and all operator actions credited in the 
accident analyses will remain the same. The proposed changes will 
not result in plant operation in a configuration outside the design 
basis. The calculated impact on risk is insignificant and meets the 
acceptance criteria contained in Regulatory Guides 1.174 and 1.177. 
Although there was no attempt to quantify any positive human factors 
benefit due to increased Completion Times and bypass test times, it 
is expected that there would be a net benefit due to a reduced 
potential for spurious reactor trips and actuations associated with 
testing.
    Implementation of the proposed changes is expected to result in 
an overall improvement in safety, as follows:
    (a) Reduced testing will result in fewer inadvertent reactor 
trips, less frequent actuation of ESFAS components, [and] less 
frequent distraction of operations personnel without significantly 
affecting RTS and ESFAS reliability.
    (b) Improvements in the effectiveness of the operating staff in 
monitoring and controlling plant operation will be realized. This is 
due to less frequent distraction of the operators and shift 
supervisor to attend to instrumentation Required Actions with short 
Completion Times.
    (c) Longer repair times associated with increased Completion 
Times will lead to higher quality repairs and improved reliability.
    (d) The Completion Time extensions for the reactor trip breakers 
will provide the utilities additional time to complete test and 
maintenance activities while at power, potentially reducing the 
number of forced outages related to compliance with reactor trip 
breaker Completion Times, and provide consistency with the 
Completion Times for the logic trains.

[[Page 5214]]

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

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

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

    Date of application for amendment: January 29, 2003, as 
supplemented by letter dated September 15, 2003.
    Brief description of amendment: The amendment proposes a one-time 
Technical Specification change to extend the test interval for the next 
Appendix J Type A test and the next drywell bypass leakage rate test 
from 10 to 15 years.
    Date of issuance: January 8, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 160.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 10, 2003 (68 FR 
34661).
    The supplemental letter of September 15, 2003, 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 amendment is contained 
in a Safety Evaluation dated January 8, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 29, 2003.
    Brief description of amendments: The amendments revised the Updated 
Final Safety Analysis Report to implement the Boiling Water Reactor 
Vessel and Internals Project reactor pressure vessel integrated 
surveillance program as the basis for demonstrating compliance with the 
requirements of Appendix H to 10 CFR part 50.
    Date of issuance: January 14, 2004.
    Effective date: As of the date of issuance.
    Amendment Nos.: 229 and 257.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
revised the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: August 19, 2003 (68 FR 
49814).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 14, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 10, 2003, as 
supplemented December 30, 2003.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.7.3, ``Control Room Emergency Filtration (CREF) 
System,'' Surveillance Requirement (SR) 3.7.3.6, to permit a one-time 
deferral of SR 3.7.3.6 until startup from the next refueling outage 
(RF-10) to preclude a mid-cycle shutdown solely for the performance of 
this SR. SR 3.7.3.6 requires verifying that unfiltered in-leakage from 
CREF system duct work outside the control room envelope that is at 
negative pressure during accident conditions is within limits. This SR 
is required to be performed every 36 months, and can be performed only 
when the CREF system is not required to be OPERABLE (i.e., in MODES 4 
or 5, with no operations with a potential for draining the reactor 
vessel and with no fuel movement of recently irradiated fuel in 
progress).
    Date of issuance: January 16, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 158.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 25, 2003 (68 
FR 66134).
    The December 30, 2003, supplemental letter provided additional 
clarifying information that was within the scope of the original 
application and did not change the Nuclear Regulatory Commission 
staff's initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 16, 2004.
    No significant hazards consideration comments received: No.

[[Page 5215]]

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: March 24, 2003, as supplemented 
by letters dated June 25 and October 15, 2003.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) to relocate certain reactor coolant 
system cycle-specific parameter limits from the TSs to the Core 
Operating Limits Report, and revises the minimum allowable reactor 
coolant system flow rate.
    Date of issuance: January 14, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 219 and 201.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 18, 2003 (68 
FR 54749), November 18, 2003 (68 FR 65090).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 14, 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: November 14, 2002, as 
supplemented by letter dated April 14, 2003.
    Brief description of amendments: The amendments revised the 
Technical Specification 3.3.1 ``Reactor Protective System (RPS) 
Instrumentation,'' Surveillance Requirement 3.3.1.3 to add a 
correlation slope to the formula for axial power imbalance error.
    Date of Issuance: January 15, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 337, 337 and 338.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75870).
    The supplement dated April 14, 2003, provided clarifying 
information that did not change the scope of the November 14, 2002, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 15, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 1, 2003, as supplemented 
December 10, 2003.
    Brief description of amendments: The amendments revise Appendix A, 
Technical Specifications (TS), of Facility Operating License Nos. NPF-
11 and NPF-18. Specifically, the changes delete one and add two 
references to the list of analytical methods in TS 5.6.5, ``Core 
Operating Limits Report (COLR),'' that can be used to determine core 
operating limits. The deleted reference is to an analytical method that 
is no longer applicable to LaSalle County Station (LSCS). The new 
references will allow LSCS to use General Electric Company (GE) methods 
for the determination of fuel assembly critical power of Framatome 
Advanced Nuclear Fuel, Inc. (Framatome) Atrium-9B and Atrium-10 fuel. 
The changes are the result of a LSCS decision to insert GE14 fuel 
during the upcoming refueling outage at LSCS Unit 1 in January 2004. 
GE's safety analysis methodologies have been previously used at LSCS 
and GE14 fuel is currently in use at other Exelon Generation Company, 
LLC (Exelon), stations.
    The first added reference, ``GEXL96 Correlation for Atrium-9B 
Fuel,'' lists a method that was previously approved by the NRC for use 
by licensees. The second added reference, ``GEXL97 Correlation for 
Atrium-10 Fuel,'' lists a GE method for determining the critical power 
for Atrium-10 fuel. This correlation had not been previously reviewed 
and approved by the NRC for use by licensees. Additionally, editorial 
changes are made to existing references.
    Date of issuance: January 9, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 164 and 150.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2003 (68 
FR 64135). The supplement dated December 10, 2003, provided clarifying 
information that did not change the scope of the July 1, 2003, 
application nor the initial no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 9, 2004.
    No significant hazards consideration comments received: No.

GPU Nuclear Corporation and Saxton Nuclear Experimental Corporation 
(SNEC), Docket No. 50-146, Saxton Nuclear Experimental Facility (SNEF)

    Date of application for amendment: April 22, 2002, as supplemented 
on December 5, 2002, and September 30 and December 22, 2003.
    Brief description of amendment: The amendment allows removal of the 
upper half of the SNEF containment vessel and makes a change to the 
organization to add the position of Vice-President GPU Nuclear 
Oversight to reflect the merger of GPU Inc. and FirstEnergy Corp.
    Date of Issuance: January 9, 2004.
    Effective Date: January 9, 2004.
    Amendment No.: 19.
    Amended Facility License No. DPR-4: Amendment changed the Technical 
Specifications.
    Date of initial notice in the Federal Register: January 7, 2003, 
with a correction notice published on January 22, 2003. The letters of 
September 30 and December 22, 2003, supplied clarifying information 
that did not expand the scope of the January 5, 2003, or January 22, 
2003, Federal Register Notices. The Commission's related evaluation of 
the amendment is contained in a Safety Evaluation dated January 9, 
2004.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant, 
Van Buren County, Michigan

    Date of application for amendment: October 17, 2002, as 
supplemented December 10, 2003.
    Brief description of amendment: The amendment revises Technical 
Specification Table 3.3.1-2 by modifying a constant in the variable 
thermal margin/low pressure (TM/LP) trip equation. The change reduces 
calculated values for the variable TM/LP trip setpoint, and results 
from improvements in plant equipment used to establish the TM/LP trip 
setpoint. Ultrasonic feedwater flow measurement devices, which were 
recently installed at Palisades, result in less uncertainty applied in 
the methodology used for determining core power level. The devices used 
to calculate the TM/LP trip setpoint were previously replaced with 
digital thermal margin monitors having less uncertainty.
    Date of issuance: January 8, 2004.

[[Page 5216]]

    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 214.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 2, 2003 (68 
FR 52235).
    The December 10, 2003, letter provided additional information in 
support of the initial application, did not expand the scope of the 
application as originally noticed, and did not effect the NRC's 
original proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated January 8, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 18, 2003, as revised by letter 
dated August 28, 2003, and supplemental letters dated October 31 and 
December 15, 2003.
    Brief description of amendment: The amendment revises the renewed 
operating license and technical specifications to increase the licensed 
rated power by 1.6 percent from 1500 megawatts thermal (MWt) to 1524 
MWt.
    Date of issuance: January 16, 2004.
    Effective date: January 16, 2004, and shall be implemented within 
30 days of the date of issuance. Modifications associated with the 
measurement uncertainty recapture power uprate will be completed prior 
to implementation. This includes: (1) Implementation of control room 
alarm functions, and (2) Figure 2-1 of the Pressure-Temperature Limits 
Report will be revised prior to the reactor vessel reaching 39.9 
effective full power years of operation.
    Amendment No.: 224.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the Operating License and Technical Specifications.
    Date of initial notice in Federal Register: September 18, 2003 (68 
FR 54751).
    The October 31 and December 15, 2003, 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 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 January 16, 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 No. 50-366, Edwin I. Hatch Nuclear 
Plant, Unit 2, Appling County, Georgia

    Date of application for amendment: December 4, 2002, as 
supplemented by letters dated June 24, and October 23, 2003.
    Brief description of amendment: The amendment revised the Technical 
Specification regarding the turbine building high temperature primary 
containment isolation value specified in Table 3.3.6.1-1, Item 1f.
    Date of issuance: January 12, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 181.
    Renewed Facility Operating License No. NPF-5: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 21, 2003 (68 FR 
2807).
    The supplements dated June 24 and October 23, 2003, provided 
clarifying information that did not change the scope of the December 4, 
2002, application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 12, 2004.
    No significant hazards consideration comments received: No.

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 application for amendments: February 26, 2003, as 
supplemented by letter dated July 25, 2003.
    Brief description of amendments: The amendments revised the 
Technical Specifications Section 5.5.17, ``Containment Leakage Rate 
Testing Program,'' to reflect a one time deferral of the Type-A 
Containment Integrated Leak Rate Test (ILRT). The 10-year interval 
between ILRTs is to be extended to 15 years from the previous ILRTs 
that were completed in March 2002 for Unit 1 and March 1995 for Unit 2.
    Date of issuance: January 12, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 130 and 108.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 13, 2003 (68 FR 
25658).
    The supplement dated July 25, 2003, provided clarifying information 
that did not change the scope of the February 26, 2003, application nor 
the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 12, 2004.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 26th day of January 2004.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 04-2017 Filed 2-2-04; 8:45 am]
BILLING CODE 7590-01-P