[Federal Register Volume 69, Number 51 (Tuesday, March 16, 2004)]
[Notices]
[Pages 12361-12376]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 04-5596]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from, January 20, 2004, through March 4, 2004. 
The last biweekly notice was published on March 2, 2004 (69 FR 9857).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the

[[Page 12362]]

Commission make a final No Significant Hazards Consideration 
Determination, any hearing will take place after issuance. The 
Commission expects that the need to take this action will occur very 
infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, hearingdocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to 301-415-3725 or by email to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the

[[Page 12363]]

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 email to pdr@nrc.gov.

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

    Date of amendment request: November 11, 2003.
    Description of amendment request: The proposed amendment would 
amend Appendix A, Technical Specifications (TS), of Facility Operating 
License No. NPF-62 for Clinton Power Station (CPS). The proposed 
changes would revise several CPS TS instrument channel trip setpoint 
Allowable Values.
    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 amendment implements revised Allowable Values for 
the following instrument functions.

 Main Steam Isolation Valve--Closure
 Anticipated Transient Without Scram Recirculation 
Pump Trip Reactor Steam Dome Pressure--High
 Reactor Vessel Pressure--Low (Injection 
Permissive)
 Reactor Vessel Water Level--Low Low Low, Level 1
 Reactor Vessel Water Level--Low Low, Level 2
 High Pressure Core Spray (HPCS) System Reactor 
Vessel Water Level--High, Level 8
 Reactor Core Isolation Cooling (RCIC) Storage 
Tank Level--Low
 HPCS System Suppression Pool Water Level--High 
(Pump Suction Transfer)
 Automatic Depressurization System (ADS) 
Initiation Permissive, Low Pressure Core Spray (LPCS) Pump Discharge 
Pressure--High
 ADS Initiation Permissive, Low Pressure Coolant 
Injection (LPCI) Pumps Discharge Pressure--High
 RCIC System Suppression Pool Water Level--High 
(Pump Suction Transfer)
 Main Steam Line Pressure--Low, and
 Safety Relief Valve (SRV) Relief and Low-Low Set 
(LLS) functions channel calibration surveillance requirement

    The proposed changes do not require modification to the 
facility. There is no impact on the accident analysis as a result of 
the proposed changes to the Allowable Values. The analytical limit, 
which is used as input to the accident analysis, does not change. 
The proposed changes will be implemented through revision of the 
associated surveillance test procedures, where the revised Allowable 
Value will replace the existing value.
    Derivation of the Allowable Value in accordance with Regulatory 
Guide 1.105, ``Instrument Setpoints,'' uses the analytical limit as 
a fixed starting point from which instrument uncertainties are added 
or subtracted, as appropriate. Calculation of the Allowable Value to 
plant-specific parameters provides additional confidence that 
protective instrumentation that passes the surveillance testing 
criteria will perform its design function without exceeding the 
associated safety analysis limit.
    The revised Allowable Values for the affected equipment are not 
considered an initiator to any previously analyzed accident and 
therefore, cannot increase the probability of any previously 
evaluated accident. Implementation of the revised Allowable Values 
will ensure that the instrumentation will perform its required 
function to meet the accident analysis assumptions. The proposed 
Allowable Values will ensure that the fuel is adequately cooled, 
containment and drywell are isolated as required, primary 
containment temperature and pressure design limits are met, and 
overpressurization of the nuclear steam supply system is prevented 
following an accident or transient. The proposed changes do not 
increase the probability of any accident previously evaluated.
    Since the proposed changes ensure the same level of protection 
as assumed in the accident analyses, the conclusions of the accident 
scenarios remain valid. As a result, no changes to radiological 
release parameters are involved. Therefore, the proposed changes do 
not increase the consequences of an accident previously evaluated.
    In summary, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    The proposed changes do not affect the design, functional 
performance or operation of the facility. Similarly, they do not 
affect the design or operation of any structures, systems, or 
components involved in the mitigation of any accidents, nor do they 
affect the design or operation of any component in the facility such 
that new equipment failure modes are created. Setpoints remain the 
same and therefore, there is no impact on the operation of any of 
the associated systems.
    As such the proposed amendment 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 changes do not involve a change to the plant design 
or operation. The proposed changes will be implemented through 
revisions to the associated surveillance test procedures where the 
revised Allowable Value replaces the existing Allowable Value. No 
changes to the instrument setpoints are involved. Since the 
availability of the systems will be maintained and since the system 
designs are unaffected, the proposed changes ensure the 
instrumentation is capable of performing their intended functions. 
The proposed changes do not affect the accident analyses that assume 
the operability of the instrumentation associated with these 
Allowable Values. The margins associated with the analytical limits 
are not impacted by the proposed Allowable Values since the 
analytical limits remain unchanged.
    Therefore, operation of CPS in accordance with the proposed 
changes will 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: Edward J. Cullen, Deputy General Counsel 
Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of amendment request: August 6, 2003, as supplemented on 
February 13, 2004.
    Description of amendment request: This amendment would revise the 
Technical Specifications (TSs) to incorporate reference to the 10 CFR 
50.55a, Codes and Standards, in lieu of the existing criteria of 
Regulatory Guide 1.35.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed revision to Technical Specification 4.4.2.1 and 
associated Bases Section incorporates reference to the criteria of 
10 CFR 50.55a, ``Codes and standards,'' in lieu of the existing 
criteria of Regulatory Guide 1.35. This change provides consistency 
between the Technical Specification tendon surveillance program 
criteria and the regulatory requirements specified in 10 CFR 
50.55a(b)(2)(vi). These regulatory requirements and the associated 
surveillance program ensure that the reactor building tendon 
prestressing system is capable of maintaining the structural 
integrity of the containment during operating

[[Page 12364]]

and accident conditions. The reactor building prestressing system is 
not an initiator of any accident. Therefore, this change is not 
related to the probability of any accident previously evaluated. 
This change ensures that the containment tendon surveillance program 
addresses the appropriate regulatory criteria. This change does not 
result in any reduction in the effectiveness of the existing 
surveillance program. The tendon surveillance program will continue 
to ensure that the containment structure is capable of performing 
its intended safety function in the event of a design basis 
accident. Therefore, this change has no affect on the consequences 
of an accident previously evaluated.
    The proposed changes to Technical Specification Definition 1.22, 
Technical Specification 3.1.6.6 and associated Bases, and Technical 
Specification 3.24 Bases are only administrative changes or 
corrections and have no affect on plant design or operations.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed revision to Technical Specification 4.4.2.1 and 
associated Bases Section incorporates reference to the criteria of 
10 CFR 50.55a, ``Codes and standards,'' in lieu of the existing 
criteria of Regulatory Guide 1.35. This change provides consistency 
between the Technical Specification tendon surveillance program 
criteria and the regulatory requirement specified in 10 CFR 
50.55a(b)(2)(vi). The proposed Technical Specification change does 
not result in any reduction in effectiveness of the existing tendon 
surveillance program. The tendon surveillance program will continue 
to satisfy the applicable Technical Specification and regulatory 
required criteria, thus ensuring that the containment structure will 
perform its design safety function. This change has no affect on the 
design and operation of plant structures, systems, and components. 
This change does not introduce any new accident precursors and does 
not involve any alterations to plant configurations, which could 
initiate a new or different kind of accident.
    The proposed changes to Technical Specification Definition 1.22, 
Technical Specification 3.1.6.6 and associated Bases, and Technical 
Specification 3.24 Bases are only administrative changes or 
corrections and have no affect on plant design or operations.
    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 revision to Technical Specification 4.4.2.1 and 
associated Bases Section incorporates reference to the criteria of 
10 CFR 50.55a, ``Codes and standards,'' in lieu of the existing 
criteria of Regulatory Guide 1.35. The change provides consistency 
between the Technical Specification tendon surveillance program 
criteria and the regulatory requirement specified in 10 CFR 
50.55a(b)(2)(vi). The containment examination and inspection 
requirements specified in 10 CFR 50.55a(b)(2)(vi) meet the same 
standards as the criteria specified in Regulatory Guide 1.35. The 
proposed Technical Specification change does not result in any 
reduction in effectiveness of the existing tendon surveillance 
program. The tendon surveillance program will continue to satisfy 
the applicable Technical Specification and regulatory required 
criteria, thus ensuring that the containment structure will perform 
its design safety function in accordance with existing margins of 
safety for containment integrity.
    The proposed changes to Technical Specification Definition 1.22, 
Technical Specification 3.1.6.6 and associated Bases, and Technical 
Specification 3.24 Bases are only administrative changes or 
corrections and have no affect on plant design or operations.
    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: Edward J. Cullen, Jr., Esquire, Vice 
President, General Counsel and Secretary, Exelon Generation Company, 
LLC, 300 Exelon Way, Kennett Square, PA 19348.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendments request: December 15, 2003.
    Description of amendments request: The proposed amendment would 
revise Technical Specification 3.1.8, ``Scram Discharge Volume (SDV) 
Vent and Drain Valves,'' to allow a vent or drain line with one 
inoperable valve to be isolated instead of requiring the valve to be 
restored to Operable status within 7 days.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on February 24, 2003 (68 FR 8637), on possible 
amendments to revise the action for one or more SDV vent or drain lines 
with an inoperable valve, including a model safety evaluation and model 
no significant hazards consideration (NSHC) determination, using the 
consolidated line-item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on April 15, 
2003 (68 FR 18294). The licensee affirmed the applicability of the 
model NSHC determination in its application dated December 15, 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.
    A change is proposed to allow the affected SDV vent and drain 
line to be isolated when there are one or more SDV vent or drain 
lines with one valve inoperable instead of requiring the valve to be 
restored to operable status within 7 days. With one SDV vent or 
drain valve inoperable in one or more lines, the isolation function 
would be maintained since the redundant valve in the affected line 
would perform its safety function of isolating the SDV. Following 
the completion of the required action, the isolation function is 
fulfilled since the associated line is isolated. The ability to vent 
and drain the SDVs is maintained and controlled through 
administrative controls. This requirement assures the reactor 
protection system is not adversely affected by the inoperable 
valves. With the safety functions of the valves being maintained, 
the probability or consequences of an accident previously evaluated 
are not significantly increased.
    Criterion 2--The proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any previously evaluated.
    Criterion 3--The proposed change does not involve a significant 
reduction in the margin of safety.
    The proposed change ensures that the safety functions of the SDV 
vent and drain valves are fulfilled. The isolation function is 
maintained by redundant valves and by the required action to isolate 
the affected line. The ability to vent and drain the SDVs is 
maintained through administrative controls. In addition, the reactor 
protection system will prevent filling of an SDV to the point that 
it has insufficient volume to accept a full scram. Maintaining the 
safety functions related to isolation of the SDV and insertion of 
control rods ensures that the proposed change does not involve a 
significant reduction in the margin of safety.


[[Page 12365]]


    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: William Burton, Acting.

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

    Date of amendment request: February 4, 2004.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications Index and Technical Specifications 
(TS) 4.4.1.3.2, ``Reactor Coolant System Hot Shutdown Surveillance 
Requirements,'' and 3.4.1.4.1.b, ``Reactor Coolant System Cold 
Shutdown--Loops Filled Limiting Condition For Operation.'' The proposed 
change to the Index is an administrative update to restore consistency 
with other sections of the TS. The proposed change to TS 4.4.1.3.2 and 
TS 3.4.1.4.1.b eliminates a requirement that the wide-range 
instrumentation be inoperable before the narrow-range instrumentation 
can be used for confirmation of the minimum steam generator secondary 
side water level. The primary reason for this proposed change to TS 
4.4.1.3.2 and TS 3.4.1.4.1.b is to provide the operational flexibility 
needed for a smooth transition through the applicable range of 
operating conditions.
    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.
    There is no impact on previously evaluated accidents because the 
proposed amendment does not affect the capability of any structure, 
system, or component to perform its design function. The functional 
capability of the narrow range instrumentation is not impacted by 
the operability status of the wide range instrumentation. The 
existing minimum values specified by Technical Specifications for 
the wide range and the narrow range instrumentation conservatively 
incorporate the applicable uncertainties necessary to make either 
instrument suitable for use over the expected range of operating 
conditions. As a result, the proposed amendment does not affect the 
operating procedures and administrative controls that have the 
function of preventing or mitigating any [previously] evaluated 
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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not change the design function or 
operation of any structure, system, or component. The proposed 
amendment does not involve any physical change to plant equipment. 
Use of the narrow range instrumentation while the wide range 
instrumentation is operable does not create any new or different 
failure mechanisms, malfunctions, or accident initiators than those 
already considered in the design and licensing bases.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment does not affect the margin of safety 
because the existing minimum values specified by Technical 
Specifications for the wide range and the narrow range 
instrumentation are not changed. Those minimum values conservatively 
incorporate the applicable uncertainties necessary to make either 
instrument suitable for use over the expected range of operating 
conditions. The calculation of those uncertainties for use of the 
narrow range instrumentation is unaffected by the operating status 
of the wide range instrumentation.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, [Carolina Power & Light Company] concludes 
that the proposed amendment involves no significant hazards 
consideration under the standards set forth in 10 CFR 50.92(c), and, 
accordingly, a finding of ``no significant hazards consideration'' 
is justified.

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

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

    Date of amendment request: June 3, 2003.
    Description of amendment request: Pursuant to Title 10 of the Code 
of Federal Regulations, Section 50.90, Duke Energy Corporation 
requested an amendment to the McGuire Nuclear Station Facility 
Operating Licenses and Technical Specifications. The proposed change 
would add a note to Limiting Condition of Operation 3.7.11, ``Auxiliary 
Building Filtered Ventilation Exhaust System (ABFVES)'', that would 
allow the Auxiliary Building pressure boundary to be opened 
intermittently under administrative control. Changes to the 
corresponding Bases would also be made to establish the administrative 
controls that are required to minimize the consequences of the open 
pressure boundary.
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No, the Auxiliary Building Filtered Ventilation Exhaust System 
(ABFVES) is not assumed to be an initiator of any analyzed accident. 
Therefore, the proposed change contained in this license amendment 
request has no significant impact on the probability of occurrence 
of any previously analyzed accident.
    The ABFVES provides a means of filtering air from the area of 
the active emergency core cooling system (ECCS) components, thereby 
providing environmental control for temperature and humidity in the 
ECCS pump room area and the Auxiliary Building. During emergency 
operations, the ABFVES exhausts air from the mechanical penetration 
area and the ECCS pump room area and discharges it through the 
system filters. For cases where the Auxiliary Building pressure 
boundary is opened intermittently under administrative controls, 
appropriate compensatory measures would be required by the proposed 
Technical Specification to ensure the pressure boundary can be 
rapidly restored. Based on the compensatory measures available to 
the plant operators and the administrative controls required to 
rapidly restore an opened pressure boundary, the accident 
consequences do not cause a significant increase in dose above the 
applicable General Design Criter[i]a, Standard Review Plan, or 10 
CFR [Part] 100 limits.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No, there are no changes being made to actual plant hardware 
which will result in

[[Page 12366]]

any new accident causal mechanisms. Also, no changes are being made 
to the way in which the plant is being operated. Therefore, no new 
accident causal mechanisms will be generated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    No, margin of safety is related to the ability of the fission 
product barriers to perform their design functions during and 
following accident conditions. These barriers include the fuel 
cladding, the reactor coolant system, and the containment system. 
The performance of these barriers will not be significantly degraded 
by the proposed changes. When the Auxiliary Building pressure 
boundary is open on an intermittent basis, as permitted by the 
changes proposed in this license amendment request, administrative 
controls would be in place to ensure that the integrity of the 
pressure boundary could be rapidly restored. Therefore, it is 
expected that the plant, and the operating personnel, would maintain 
the ability to mitigate design basis events, and that none of the 
fission product barriers would be significantly affected by this 
change. Therefore, the proposed change is not considered to result 
in 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: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: October 15, 2003.
    Description of amendment request: The amendments would add a new 
Technical Specification (TS) 3.9.7, ``Unborated Water Source isolation 
Valves,'' and would revise TS 3.9.2, ``Nuclear Instrumentation,'' to 
delete the requirement for Boron Dilution Mitigation System automatic 
valve actuations and makeup water pump trip during Mode 6 and to agree 
with the wording of NUREG-1431, ``Standard Technical Specifications 
Westinghouse Plants,'' Revision 2. The licensee proposed these changes 
to provide configuration control of the dilution valves during Mode 6 
to preclude the possibility of a boron dilution event and to provide an 
opportunity to conduct maintenance on the volume control tank valves, 
refueling water storage tank valves, and their respective power 
supplies.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Operation of the facilities in accordance with this amendment 
would not involve a significant increase in the probability or 
consequences of an accident previously evaluated. The BDMS [Boron 
Dilution Mitigation System] system is designed to mitigate the 
consequences of an inadvertent boron dilution event. The probability 
of the dilution accident will be reduced by administratively 
isolating potential dilution flow paths. Thus, with the proposed 
changes, boron dilution is not considered a credible accident during 
refueling.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Operation of the facilities in accordance with this amendment 
would not create the possibility of a new or different kind of 
accident from any accident previously evaluated. No new accident 
causal mechanisms are created as a result of this proposed 
amendment. No changes are being made to any structure, system, or 
component which will introduce any new accident causal mechanisms. 
This amendment request does not impact any plant systems that are 
accident initiators and does not impact any safety analysis.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    Operation of the facilities in accordance with this amendment 
would not involve a significant reduction in a margin of safety. The 
design criterion and margin of safety for the current BDMS is that 
the dilution event is terminated prior to the loss of all shutdown 
margin. The same criterion will be met following the isolation of 
dilution valves. Therefore, there is no 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: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of amendment request: February 18, 2004.
    Description of amendment request: The proposed amendment deletes 
requirements from the Technical Specifications (TS) to maintain 
hydrogen recombiners and hydrogen and oxygen monitors. Licensees were 
generally required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island Nuclear Station] Action Plan 
Requirements,'' and Regulatory Guide (RG) 1.97, ``Instrumentation for 
Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs 
Conditions During and Following an Accident.'' Implementation of these 
upgrades was an outcome of the lessons learned from the accident that 
occurred at TMI, Unit 2. Requirements related to combustible gas 
control were imposed by Order for many facilities and were added to or 
included in the 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 February 18, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was

[[Page 12367]]

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 monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
RG 1.97 Category 1, is intended for key variables that most directly 
indicate the accomplishment of a safety function for design-basis 
accident events. The hydrogen and oxygen monitors no longer meet the 
definition of Category 1 in RG 1.97. As part of the rulemaking to 
revise 10 CFR 50.44 the Commission found that Category 3, as defined 
in RG 1.97, is an appropriate categorization for the hydrogen 
monitors because the monitors are required to diagnose the course of 
beyond design-basis accidents. Also, as part of the rulemaking to 
revise 10 CFR 50.44, the Commission found that Category 2, as 
defined in RG 1.97, is an appropriate categorization for the oxygen 
monitors, because the monitors are required to verify the status of 
the inert containment.
    The regulatory requirements for the hydrogen and oxygen monitors 
can be relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, 
classification of the oxygen monitors as Category 2, and removal of 
the hydrogen and oxygen monitors from TS will not prevent an 
accident management strategy through the use of the SAMGs, the 
emergency plan (EP), the emergency operating procedures (EOP), 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.

    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: January 15, 2004.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) Section 5.5.12, ``Primary 
Containment Leakage Rate Testing Program,'' to reflect a one-time 
deferral of the primary containment Type A test to no later than 
February 27, 2011, for Unit 2, and no later than July 13, 2009, for 
Unit 3.
    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?
    The proposed change will revise Dresden Nuclear Power Station 
(DNPS) Units 2 and 3 Technical Specifications (TS) Section 5.5.12, 
``Primary Containment Leakage Rate Testing Program,'' to reflect a 
one-time deferral of the primary containment Type A test to no later 
than February 27, 2011, for Unit 2, and no later than July 13, 2009, 
for Unit 3. The current Type A test interval of 10 years, based on 
past performance, would be extended on a one-time basis to 15 years 
from the last Type A test.
    The function of the primary containment is to isolate and 
contain fission products released from the reactor coolant system 
(RCS) following a design basis loss-of-coolant accident (LOCA) and 
to confine the postulated release of radioactive material to within 
limits. The test interval associated with Type A testing is not a 
precursor of any accident previously evaluated. Therefore, extending 
this test interval on a one-time basis from 10 years to 15 years 
does not result in an increase in the probability of occurrence of 
an accident. The successful performance history of Type A testing 
provides assurance that the DNPS primary containments will not 
exceed allowable leakage rate values specified in the TS and will 
continue to perform their design function following an accident. The 
risk assessment of the proposed change has concluded that there is 
an insignificant increase in total population dose rate and an 
insignificant increase in the conditional containment failure 
probability.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change for a one-time extension of the Type A tests 
for DNPS Units 2 and 3 will not affect the control parameters

[[Page 12368]]

governing unit operation or the response of plant equipment to 
transient and accident conditions. The proposed change does not 
introduce any new equipment or modes of system operation. No 
installed equipment will be operated in a new or different manner. 
As such, no new failure mechanisms are introduced.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    DNPS Units 2 and 3 are General Electric BWR/3 plants with Mark I 
primary containments. The Mark I primary containment consists of a 
drywell, which encloses the reactor vessel, reactor coolant 
recirculation system, and branch lines of the RCS; a toroidal-shaped 
pressure suppression chamber containing a large volume of water; and 
a vent system connecting the drywell to the water space of the 
suppression chamber. The primary containment is penetrated by 
access, piping, and electrical penetrations.
    The integrity of the primary containment penetrations and 
isolation valves is verified through Type B and Type C local leak 
rate tests (LLRTs) and the overall leak-tight integrity of the 
primary containment is verified by a Type A integrated leak rate 
test (ILRT) as required by 10 CFR 50, Appendix J, ``Primary Reactor 
Containment Leakage Testing for Water-Cooled Power Reactors.'' The 
tests are performed to verify the essentially leak-tight 
characteristics of the primary containment at the design basis 
accident pressure. The proposed change for a one-time extension of 
the Type A tests do not affect the method for Type A, B, or C 
testing, or the test acceptance criteria. In addition, based on 
previous Type A testing results, EGC does not expect additional 
degradation, during the extended period between Type A tests, which 
would result in a significant reduction in a margin of safety.
    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of application for amendment request: January 15, 2004.
    Description of amendment request: Modify Technical Specification 
Surveillance Requirement 3.4.3.2, SR 3.5.1.10, and SR 3.6.1.6.1 to 
provide an alternative means for testing the main steam Electromatic 
relief valves and the dual function Target Rock safety/relief valves.
    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?
    The proposed changes modify Technical Specifications (TS) 
Surveillance Requirement (SR) 3.4.3.2, SR 3.5.1.10, and SR 3.6.1.6.1 
to provide an alternative means for testing the main steam line 
relief valves, automatic depressurization system valves, and low set 
relief valves. Accidents are initiated by the malfunction of plant 
equipment, or the catastrophic failure of plant structures, systems 
or components. The performance of relief valve testing is not a 
precursor to any accident previously evaluated and does not change 
the manner in which the valves are operated. The proposed testing 
requirements will not contribute to the failure of the relief valves 
nor any plant structure, system or component. Exelon Generation 
Company, LLC has determined that the proposed change in testing 
methodology provides an equivalent level of assurance that the 
relief valves are capable of performing their intended safety 
functions. Thus, the proposed changes do not affect the probability 
of an accident previously evaluated.
    The performance of relief valve testing provides confidence that 
the relief valves are capable of depressurizing the reactor pressure 
vessel (RPV). This will protect the reactor vessel from 
overpressurization and allow the combination of the Low Pressure 
Coolant Injection and Core Spray systems to inject into the RPV as 
designed. The low set relief logic causes two low set relief valves 
to be opened at a lower pressure than the relief mode pressure 
setpoints and causes the low set relief valves to stay open longer, 
such that reopening of more than one valve is prevented on 
subsequent actuations. Thus, the low set relief function prevents 
excessive short duration relief valve cycles with valve actuation at 
the relief setpoint, which limits induced thrust loads on the relief 
valve discharge line for subsequent actuations of the relief valve. 
The proposed changes do not affect any function related to the 
safety mode of the duel function safety/relief valves. The proposed 
changes involve the manner in which the subject valves are tested, 
and have no effect on the types or amounts of radiation released or 
the predicted offsite does in the events of an accident. The 
proposed testing requirements are sufficient to provide confidence 
that the relief valves are capable of performing their intended 
safety functions. In addition, a stuck open relief valve accident is 
analyzed in the Updated Final Safety Analysis Report. Since the 
proposed testing requirements do not alter the assumptions for the 
stuck open relief valve accident, the radiological consequences of 
any accident previously evaluated are not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes do not affect the assumed accident 
performance of the main steam relief valves, nor any plant 
structure, system, or component previously evaluated. The proposed 
changes do not install any new equipment, and installed equipment is 
not being operated in a new or different manner. The proposed change 
in test methodology will ensure that the valves remain capable of 
preforming their safety functions due to meeting the testing 
requirements of the American Society of Mechanical Engineers Boiler 
and Pressure Vessel Code, with the exception of opening the valve 
following installation or maintenance for which a relief request has 
been submitted, proposing an acceptable alternative. No setpoints 
are being changed which would alter the dynamic response of plant 
equipment. Accordingly, no new failure modes are introduced.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed changes will allow testing of the valve actuation 
electrical circuitry, including the solenoid, and mechanical 
actuation components, without causing the relief valve to open. The 
relief valves will be manually actuated prior to installation in the 
plant. Therefore, all modes of relief valve operation will be tested 
prior to entering the mode of operation requiring the valve to 
perform their safety functions. The proposed changes do not affect 
the valve setpoint or the operational criteria that directs the 
relief valves to be manually opened during plants transients. There 
are no changes proposed which alter the setpoints at which 
protective actions are initiated, and there is no change to the 
operability requirements for equipment assumed to operate for 
accident mitigation.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Vice President, 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.

[[Page 12369]]

    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: January 27, 2004.
    Description of amendment request: The proposed change would revise 
Technical Specification 3.4.5 to allow repair of steam generator tubes 
by installation of leak limiting Alloy 800 sleeves.
    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?
    No. The leak limiting Alloy 800 sleeves are designed using the 
applicable American Society for Mechanical Engineers (ASME) Boiler 
and Pressure Vessel Code [ASME Code] and, therefore, meet the design 
objectives of the original steam generator (SG) tubing. The applied 
stresses and fatigue usage for the sleeves are bounded by the limits 
established in the ASME Code. Mechanical testing has shown that the 
structural strength of sleeves under normal, upset, emergency, and 
faulted conditions provides margin to the acceptance limits. These 
acceptance limits bound the most limiting (three times normal 
operating pressure differential) burst margin recommended by NRC 
Regulatory Guide 1.121, ``Bases for Plugging Degraded PWR Steam 
Generator Tubes.'' Burst testing of sleeve-tube assemblies has 
confirmed the analytical results and demonstrated that no 
unacceptable levels of primary-to-secondary leakage are expected 
during any plant condition.
    The leak limiting Alloy 800 sleeve depth-based structural limit 
is determined using NRC guidance and the pressure stress equation of 
ASME Code, Section III with additional margin added to account for 
the configuration of long axial cracks. An Alloy 800 sleeved tube 
will be plugged on detection of an imperfection in the sleeve or in 
the pressure boundary portion of the original tube wall in the leak 
limiting sleeve/tube assembly.
    Evaluation of the repaired SG tube testing and analysis 
indicates no detrimental effects on the leak limiting Alloy 800 
sleeve or sleeved tube assembly from reactor system flow, primary or 
secondary coolant chemistries, thermal conditions or transients, or 
pressure conditions as may be experienced at Beaver Valley Power 
Station (BVPS) Unit [No.] 1. Corrosion testing and historical 
performance of sleeve-tube assemblies indicates no evidence of 
sleeve or tube corrosion considered detrimental under anticipated 
service conditions.
    The implementation of the proposed change has no significant 
effect on either the configuration of the plant or the manner in 
which it is operated. The consequences of a hypothetical failure of 
the leak limiting Alloy 800 sleeve-tube assembly is bounded by the 
current SG tube rupture (SGTR) analysis described in the BVPS Unit 
No. 1 Updated Final Safety Analysis Report. Due to the slight 
reduction in the inside diameter caused by the sleeve wall 
thickness, primary coolant release rates through the parent tube 
would be slightly less than assumed for the SGTR analysis and 
therefore, would result in lower total primary fluid mass release to 
the secondary system. A main steam line break or feedwater line 
break will not cause a SGTR since the sleeves are analyzed for a 
maximum accident differential pressure greater than that predicted 
in the BVPS Unit No. 1 safety analysis. The sleeve-tube assembly 
leakage during plant operation would be minimal and is well within 
the allowable Technical Specification leakage limits.
    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?
    No. The leak limiting Alloy 800 sleeves are designed using the 
applicable ASME Code as guidance, and therefore meet the objectives 
of the original SG tubing. As a result, the functions of the SG will 
not be significantly affected by the installation of the proposed 
sleeve. The proposed sleeves do not interact with any other plant 
systems. Any accident as a result of potential tube or sleeve 
degradation in the repaired portion of the tube is bounded by the 
existing SGTR accident analysis. The continued integrity of the 
installed sleeve-tube assembly is periodically verified by Technical 
Specification requirements and a sleeved tube will be plugged on 
detection of an imperfection in the sleeve or in the pressure 
boundary portion of the tube wall in the leak limiting sleeve/tube 
assembly.
    Implementation of the proposed change has no significant effect 
on either the configuration of the plant, or the manner in which it 
is operated.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The repair of degraded SG tubes with leak limiting Alloy 800 
sleeves restores the structural integrity of the degraded tube under 
normal operating and postulated accident conditions. The reduction 
in core cooling margin due to the addition of Alloy 800 sleeves is 
not significant because the cumulative effect of all repaired 
(sleeved) and plugged tubes will continue to be less than the 
currently allowed core cooling margin threshold established by the 
total steam generator tube plugging level. The design safety factors 
utilized for the sleeves are consistent with the safety factors in 
the ASME Boiler and Pressure Vessel Code used in the original SG 
design. The sleeve and portions of the installed sleeve-tube 
assembly that represent the reactor coolant pressure boundary will 
be monitored and a sleeved tube will be plugged on detection of an 
imperfection in the sleeve or in the pressure boundary portion of 
the original tube wall in the leak limiting sleeve/tube assembly. 
Use of the previously identified design criteria and design 
verification testing assures that the margin to safety is not 
significantly different from the original SG tubes.
    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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: January 26, 2004.
    Description of amendment request: The proposed change would revise 
the BVPS-1 and 2 Updated Final Safety Analysis Report (UFSAR) 
description of the design-basis bounding limitations for the ultimate 
heat sink design. The proposed change would allow the design 
descriptions in the BVPS-1 and 2 UFSARs to credit the current Technical 
Specification (TS) 3.7.5.1 requirement at each unit to shut down when 
the Ohio River level reaches a low level below 654 feet mean sea level 
(msl). This UFSAR revision would preclude design consideration for 
design-basis accidents associated with power operation from occurring 
when the Ohio River level is below 654 feet msl since the units would 
be required to be shut down.
    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?
    No. The proposed change will revise the BVPS Unit No. 1 and Unit 
No. 2 UFSAR

[[Page 12370]]

description of the design basis bounding limitations for the 
ultimate heat sink design. FENOC's proposed change will allow the 
design description in each BVPS Unit's UFSAR to credit the current 
[TS] 3.7.5.1 requirement at each BVPS Unit to shutdown when the Ohio 
River level reaches a low level below 654 feet Mean Sea Level (msl). 
This UFSAR revision will, therefore, preclude design consideration 
for design bases accidents associated with power operation from 
occurring when the Ohio River level is below 654' msl since the 
plant will already be shutdown. This LAR [license amendment request] 
does not propose any Technical Specification changes nor any 
physical plant changes.
    Since no physical plant changes nor any instrument setpoint 
changes are being requested, it [the proposed change] would not 
result in an increase in [the] probability of an accident previously 
evaluated. Since the proposed change only clarifies the limiting 
design basis ultimate heat sink scenario, consistent with both 
Units' original licensing bases, it would not result in a 
significant increase in the consequences of an accident previously 
evaluated.
    In conclusion, the request to amend the UFSARs for BVPS Unit 
Nos. 1 and 2 to clarify the limiting design basis ultimate heat sink 
scenario, consistent with both Units' original licensing bases, does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    No. The proposed changes only clarif[y] the limiting design 
basis ultimate heat sink scenario, consistent with both Units' 
original licensing bases. Since this is not a change to [the] 
original licensing bases and the design for the River Water System, 
Service Water System, Intake Structure, and [the] ultimate heat sink 
will remain valid for all credible plant conditions, this does not 
induce a new mechanism that would result in a different kind of 
accident from those previously analyzed.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed changes [sic] only clarif[y] the limiting 
design basis ultimate heat sink scenario, consistent with both 
Units' original licensing bases. The proposed bounding conditions 
bound the credible BVPS Unit 1 and Unit 2 operating conditions. The 
design for the River Water System, Service Water System, Intake 
Structure, and ultimate heat sink continue to meet General Design 
Criteria 2 and 44 and the recommendations of Regulatory Guide 1.27, 
Revision 2.
    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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: January 28, 2004.
    Description of amendment request: The proposed amendment would 
delete requirements from the Technical Specifications (TSs) to maintain 
hydrogen recombiners and hydrogen monitors. Licensees were generally 
required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' 
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' Implementation of these upgrades 
was an outcome of the lessons learned from the accident that occurred 
at TMI, Unit 2. Requirements related to combustible gas control were 
imposed by Order for many facilities and were added to or included in 
the 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 January 28, 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 monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
RG 1.97 Category 1, is intended for key variables that most directly 
indicate the accomplishment of a safety function for design-basis 
accident events. The hydrogen 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.
    The regulatory requirements for the hydrogen 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 and removal of 
the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the SAMGs, the emergency plan 
(EP), the emergency operating procedures (EOP), 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 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 monitor requirements,

[[Page 12371]]

including removal of these requirements from TS, will not result in 
any failure mode not previously analyzed. The hydrogen recombiner 
and hydrogen monitor equipment was intended to mitigate a design-
basis hydrogen release. The hydrogen recombiner and hydrogen 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 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.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from the TSs 
will not result in a significant reduction in their functionality, 
reliability, and availability.

    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.
    Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Richard J. Laufer

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

    Date of amendment request: February 3, 2004
    Description of amendment request: This amendment request proposes 
to revise a footnote to clarify a surveillance requirement and 
associated bases for emergency diesel generator testing.
    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. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    FPL Energy Seabrook, LLC (FPLE Seabrook) proposes to revise 
footnote (* * *) of Technical Specification (TS) Surveillance 
Requirement (SR) 4.8.1.1.2a.5 to remove the link created between 
actions b. and c. of TS 3.8.1.1 and the loaded surveillance testing 
requirements of SR 4.8.1.1.2a.6. This revision to footnote (* * *) 
is a change to the Technical Specifications that does not modify the 
physical design or operation of the plant and will not create a 
possibility of an accident. Strict compliance with the footnote 
requires paralleling the only operable EDG [emergency diesel 
generator] unit with the off-site grid upon entry into action 
statement[s] b. or c. of TS 3.8.1.1. Operation of the only operable 
EDG unit in this manner may increase its vulnerability for failure 
if power from the off-site grid is disturbed or lost. EDG unit 
availability for subsequent emergency demands may also be adversely 
affected.
    The proposed change will eliminate the undesirable link that 
presently exists between action statement[s] b. and c. of TS 3.8.1.1 
and SR 4.8.1.1.2a.6 but will maintain the primary purpose of the SR, 
which is to ensure that the EDG unit is capable of starting from 
standby conditions and attaining rated voltage and frequency. 
Additionally, the proposed change is consistent with the methodology 
used in NRC [Nuclear Regulatory Commission] NUREG-1431, Revision 3, 
``Standard Technical Specifications Westinghouse Plants.'' 
Therefore, the proposed change does not involve a significant 
increase [in] the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change does not affect any plant structures, 
systems, or components. The operation of plant systems and equipment 
will not be affected by this proposed change. The proposed change to 
footnote (* * *) does not have the capability to initiate accidents. 
The proposed change will eliminate the undesirable link that 
presently exists between action statement[s] b. and c. of TS 3.8.1.1 
and SR 4.8.1.1.2a.6. However, the proposed change will maintain the 
primary purpose of the SR and supporting footnote, which is to 
ensure that the EDG unit is capable of starting from standby 
conditions and attaining rated voltage and frequency. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The proposed changes do not involve a change in the operational 
limits or physical design of the plant. The proposed changes do not 
change the function or operation of plant equipment or affect the 
response of that equipment if it is called on to operate. The 
performance capability of the EDG units will not be affected. The 
proposed change will maintain the primary purpose of the SR and 
supporting footnote, which is to ensure that the EDG unit is capable 
of starting from standby conditions and attaining rated voltage and 
frequency. Therefore, the proposed change does not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis, and based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    Acting NRC Section Chief: Darrell J. Roberts.

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

    Date of amendment request: January 29, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.4.9 Pressure Temperature (P/T) 
Curve figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 for Heatup/Cooldown-Core 
not Critical, Pressure Test and Heatup/Cooldown-Core Critical 
conditions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed revisions to the Cooper Nuclear Station (CNS) P/T 
curves are based on the recommendations in Regulatory Guide (RG) 
1.99, Revision 2, and are therefore in accordance with the latest 
Nuclear Regulatory Commission (NRC) guidance. The evaluation for the 
P/T curves for 32 EFPY [Effective Full Power Years] was performed 
using the approved methodologies of 10 CFR [Part] 50, Appendix G. 
The curves generated from these methods provide guidance to ensure 
that the P/T limits will not be exceeded during any phase of reactor 
operation. Accordingly, the proposed revision to the CNS P/T curves 
is based on

[[Page 12372]]

an NRC accepted means of ensuring protection against brittle reactor 
vessel fracture, and compliance with 10 CFR [Part] 50 Appendix G. 
Therefore, this proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Based on the above, NPPD [Nebraska Public Power District] 
concludes that the proposed TS change to TS 3.4.9 P/T curves, 
figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 does not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change updates existing P/T operating limits to 
correspond to the current NRC guidance. The proposed TS change 
provides more operating flexibility in the P/T curves for in-service 
leakage and hydrostatic pressure testing, non-nuclear heatup and 
cooldown, and criticality, with the benefits primarily in the area 
of pressure test being performed at a lower temperature. The 
proposed change does not involve a physical change to the plant, add 
any new equipment or any new mode of operation. These changes 
demonstrate compliance with the brittle fracture requirements of 10 
CFR [Part] 50 Appendix G, and therefore do not create the 
possibility for a new or different kind of accident from any 
accident previously evaluated.
    Based on the above, NPPD concludes that the proposed TS change 
to TS 3.4.9 P/T curves, figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    The proposed change to the CNS P/T curves does not create a 
significant reduction in the margin of safety. The proposed change 
revises the existing CNS P/T curves to be consistent with 
recommendations of RG 1.99, Revision 2, the current NRC guidance 
given to ensure compliance with 10 CFR [Part] 50 Appendix G.
    For P/T curve development ASME [American Society of Mechanical 
Engineers] Section Xl Code [Boiler and Pressure Vessel Code] Case N-
640 uses the Kic fracture toughness curve as the lower bound for 
fracture toughness. P/T curves based on the Kic fracture toughness 
limits enhance industrial safety by expanding the P/T window in the 
low-temperature operating region. The potential benefits are a 
reduction in the duration of the pressure test and, associated 
increase in personnel safety, while conducting inspections in 
primary containment. Therefore, operational flexibility is gained 
while maintaining an adequate margin of safety to Reactor Pressure 
Vessel brittle fracture. As stated above, the development of the P/T 
curves to 32 EFPY was performed per the guidelines of 10 CFR [Part] 
50 Appendix G, and thus, the margin of safety is not significantly 
reduced as the result of the proposed TS change.
    Based on the above, NPPD concludes that the proposed TS change 
to TS 3.4.9 P/T curves, figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 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: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: February 2, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.1.8, ``Scram Discharge Volume (SDV) 
Vent and Drain Valves,'' to allow a vent or drain line with one 
inoperable valve to be isolated instead of requiring the valve to be 
restored to Operable status within 7 days.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on February 24, 2003 (68 FR 8637), on possible 
amendments to revise the action for one or more SDV vent or drain lines 
with an inoperable valve, including a model safety evaluation and model 
no significant hazards consideration (NSHC) determination, using the 
consolidated line-item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on April 15, 
2003 (68 FR 18294). The licensee affirmed the applicability of the 
model NSHC determination in its application dated February 2, 2004
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    A change is proposed to allow the affected SDV vent and drain 
line to be isolated when there are one or more SDV vent or drain 
lines with one valve inoperable instead of requiring the valve to be 
restored to operable status within 7 days. With one SDV vent or 
drain valve inoperable in one or more lines, the isolation function 
would be maintained since the redundant valve in the affected line 
would perform its safety function of isolating the SDV. Following 
the completion of the required action, the isolation function is 
fulfilled since the associated line is isolated. The ability to vent 
and drain the SDVs is maintained and controlled through 
administrative controls. This requirement assures the reactor 
protection system is not adversely affected by the inoperable 
valves. With the safety functions of the valves being maintained, 
the probability or consequences of an accident previously evaluated 
are not significantly increased.
    Criterion 2--The proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any previously evaluated.
    Criterion 3--The proposed change does not involve a significant 
reduction in the margin of safety.
    The proposed change ensures that the safety functions of the SDV 
vent and drain valves are fulfilled. The isolation function is 
maintained by redundant valves and by the required action to isolate 
the affected line. The ability to vent and drain the SDVs is 
maintained through administrative controls. In addition, the reactor 
protection system will prevent filling of an SDV to the point that 
it has insufficient volume to accept a full scram. Maintaining the 
safety functions related to isolation of the SDV and insertion of 
control rods ensures that the proposed change does not involve a 
significant reduction in the margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: February 9, 2004
    Description of amendment request: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
technical specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of Title 10 of the Code of

[[Page 12373]]

Federal Regulations (10 CFR), Part 50, Section 50.65(a)(4). Limiting 
Condition for Operation (LCO) 3.0.4 exceptions in individual TSs would 
be eliminated, several notes or specific exceptions are revised to 
reflect the related changes to LCO 3.0.4, and Surveillance Requirement 
(SR) 4.0.4 is revised to reflect the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on August 2, 
2002 (67 FR 50475), on possible amendments concerning TSTF-359, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on April 4, 2003 (68 FR 16579). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated February 9, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The 
addition of a requirement to assess and manage the risk introduced 
by this change will further minimize possible concerns. Therefore, 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident from any Previously 
Evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Thus, this change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in a Margin of Safety.
    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS LCO. The 
risk associated with this allowance is managed by the imposition of 
required actions that must be performed within the prescribed 
completion times. The net effect of being in a TS condition on the 
margin of safety is not considered significant. The proposed change 
does not alter the required actions or completion times of the TS. 
The proposed change allows TS conditions to be entered, and the 
associated required actions and completion times to be used in new 
circumstances. This use is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk. 
The change also eliminates current allowances for utilizing required 
actions and completion times in similar circumstances, without 
assessing and managing risk. The net change to the margin of safety 
is insignificant. Therefore, this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
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: February 9, 2004.
    Description of amendment request: The proposed amendment revises TS 
5.5.7, ``Reactor Coolant Pump Flywheel Inspection Program,'' to extend 
the allowable inspection interval to 20 years.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on June 24, 2003 (68 FR 37590), on possible amendments 
to extend the inspection interval for reactor coolant pump (RCP) 
flywheels, including a model safety evaluation and model no significant 
hazards consideration (NSHC) determination, using the consolidated 
line-item improvement process. The NRC staff subsequently issued a 
notice of availability of the models for referencing in license 
amendment applications in the Federal Register on October 22, 2003 (68 
FR 60422). The licensee affirmed the applicability of the model NSHC 
determination in its application dated February 9, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change to the RCP flywheel examination frequency 
does not change the response of the plant to any accidents. The RCP 
will remain highly reliable and the proposed change will not result 
in a significant increase in the risk of plant operation. Given the 
extremely low failure probabilities for the RCP motor flywheel 
during normal and accident conditions, the extremely low probability 
of a loss-of-coolant accident (LOCA) with loss of offsite power 
(LOOP), and assuming a conditional core damage probability (CCDP) of 
1.0 (complete failure of safety systems), the core damage frequency 
(CDF) and change in risk would still not exceed the NRC's acceptance 
guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per 
year). Moreover, considering the uncertainties involved in this 
evaluation, the risk associated with the postulated failure of an 
RCP motor flywheel is significantly low. Even if all four RCP motor 
flywheels are considered in the bounding plant configuration case, 
the risk is still acceptably low.
    The proposed change does 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; alter or prevent the ability 
of structures, systems, components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits; or affect the source 
term, containment isolation, or radiological release assumptions 
used in evaluating the radiological consequences of an accident 
previously evaluated. Further, the proposed change does not increase 
the type or amount of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiation exposure. The proposed change is 
consistent with the safety analysis assumptions and resultant 
consequences. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Criterion 2--The proposed change does not create the possibility 
of a new or different

[[Page 12374]]

kind of accident from any accident previously evaluated.
    The proposed change in flywheel inspection frequency does not 
involve any change in the design or operation of the RCP. Nor does 
the change to examination frequency affect any existing accident 
scenarios, or create any new or different accident scenarios. 
Further, the change does not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or alter the methods governing normal plant operation. In 
addition, the change does not impose any new or different 
requirements or eliminate any existing requirements, and does not 
alter any assumptions made in the safety analysis. The proposed 
change is consistent with the safety analysis assumptions and 
current plant operating practice. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Criterion 3--The proposed change does not involve a significant 
reduction in a margin of safety.
    The proposed change does 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 this change. The proposed change will not result 
in plant operation in a configuration outside of the design basis. 
The calculated impact on risk is insignificant and meets the 
acceptance criteria contained in RG 1.174. There are no significant 
mechanisms for inservice degradation of the RCP flywheel. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: 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 email to pdr@nrc.gov.

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

    Date of application for amendments: March 28, 2003, as supplemented 
December 5, 2003.
    Brief description of amendments: These amendments revise the 
Technical Specifications by eliminating the requirements associated 
with hydrogen recombiners and hydrogen monitors.
    Date of issuance: March 2, 2004.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 262 and 239.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 13, 2003 (68 FR 
25651)
    The December 5, 2003, supplemental letter provided clarifying 
information that did not enlarge the scope of the amendment as noticed 
in the original Federal Register notice or change the no significant 
hazards consideration.
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated March 2, 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: July 21, 2003, as supplemented 
February 5, 2004.
    Brief Description of amendments: The amendment revised the Updated 
Final Safety Analysis Report (UFSAR) to describe temporary operation of 
the turbine building ventilation system in a once-through versus 
recirculation configuration during outages.
    Date of issuance: February 26, 2004.
    Effective date: Effective as of the date of issuance shall be 
implemented in accordance with 10 CFR 50.71(e).
    Amendment Nos.: 230 and 258.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
approved changes to the UFSAR.
    Date of initial notice in Federal Register: August 5, 2003 (68 FR 
46241). The February 5, 2004, supplemental letter provided clarifying 
information only and did not change the initial proposed no significant 
hazards consideration or expand the scope of the initial application. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated February 26, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 20, 2003, as supplemented 
by letters dated June 10, September 30, and October 22, 2003
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) to update the heatup, cooldown, 
criticality, and inservice test pressure and temperature limits for the 
reactor coolant system of each unit to a maximum of 34 Effective Full 
Power Years. Additionally, the amendments revise the Low Temperature 
Overpressure (LTOP) System TSs in order to reflect the revised 
pressure-temperature limits and the revised LTOP enable temperature.
    Date of issuance: March 4, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.

[[Page 12375]]

    Amendment Nos.: 212 and 206.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 23, 2003 (68 
FR 74264).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 4, 2004.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: June 30, 2003, as supplemented by letter 
dated December 16, 2003.
    Brief description of amendment: The amendment revises the control 
room emergency ventilation system surveillance requirements (SRs) by 
modifying an existing SR related to the makeup flow rate to show that 
it is applicable to the VSF-9 train and by adding a new makeup flow 
rate SR that is applicable to the 2VSF-9 train.
    Date of issuance: March 2, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 221.
    Renewed Facility Operating License No. DPR-51: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: July 22, 2003 (68 FR 
43384).
    The December 16, 2003, supplemental letter provided clarifying 
information that did not change the scope of the original Federal 
Register notice or the original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 2, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 28, 2003, as supplemented 
by letters dated October 23 and December 5, 2003.
    Brief description of amendments: The amendments revise the 
technical specifications to reduce the main steam line low pressure 
primary containment isolation allowable value.
    Date of issuance: February 18, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 206/198, 219/213.
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 23, 2003 (68 
FR 74265). The October 23 and December 5, 2003, submittals provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 18, 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: March 31, 2003, as supplemented 
June 26, 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 change increases the upper limit 
associated with TS Table 3.3.5.1-1, ``Emergency Core Cooling System 
Instrumentation,'' Function 3.e, ``HPCS System Flow Rate--Low 
(Bypass),'' Allowable Value from less than or equal to (<=) 1704 
gallons per minute (gpm) to <= 2194 gpm.
    The change increases the Allowable Value band to account for 
instrumentation deadband, as-left setting tolerances and setpoint drift 
and to resolve historical difficulties during calibration. The current 
Allowable Value was initially provided in the LaSalle County Station TS 
during conversion to Improved Technical Specifications (ITS) format. 
This value was based on vendor supplied data and believed at the time 
to adequately account for these parameters. The upper Allowable Value 
limit is being increased based on historical performance data for the 
High Pressure Core Spray (HPCS) system flow switches. The increase in 
the allowed bypass flow rate does not affect the capability of the HPCS 
system in performing its intended safety function.
    Date of issuance: March 4, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 165 and 151.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 13, 2003 (68 FR 
25654). The supplement dated June 26, 2003, provided clarifying 
information that did not change the scope of the March 31, 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 March 4, 2004.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: December 16, 2003 as 
supplemented January 29 and February 13, 2004.
    Brief description of amendment: This amendment revised the 
Technical Specifications to allow a one-time extension of the steam 
generator tube inservice inspection interval from March 9, 2004, to 
March 31, 2005.
    Date of issuance: February 26, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 262.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 6, 2004 (69 FR 
695).
    The supplements dated January 29 and February 13, 2004, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 26, 2004.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: November 20, 2003, as 
supplemented by letter dated February 5, 2004.
    Brief description of amendment: The amendment revised Section 
2.1.1.2 of the Technical Specifications to reflect the results of 
cycle-specific calculations performed for the upcoming Operating Cycle 
10, which would employ a mixed core consisting of predominantly GE11 
fuel bundles with some new GE14 fuel bundles.

[[Page 12376]]

    Date of issuance: February 25, 2004.
    Effective date: As of the date of issuance, to be implemented prior 
to startup from Refueling Outage 9.
    Amendment No.: 112.
    Facility Operating License No. NPF-69: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 23, 2003 (68 
FR 74267).
    The supplemental letter of February 5, 2004, provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The staff's related evaluation of 
the amendment is contained in a Safety Evaluation dated February 25, 
2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 22, 2003, as supplemented 
July 9, November 5, December 15, 2003, and January 30, February 9, and 
February 20, 2004.
    Brief description of amendment: The amendment revised the Kewaunee 
Nuclear Power Plant operating license and technical specifications to 
increase the licensed rated power by 6.0 percent from 1673 megawatts 
thermal to 1772 megawatts thermal.
    Date of issuance: February 27, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 172.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 10, 2003 (68 FR 
34670).
    The supplemental letters 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 February 27, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 2, 2003, as supplemented 
by letters dated August 8 and November 13, 2003.
    Brief description of amendments: The amendments revise certain 
operational requirements of the Diablo Canyon Nuclear Plant Technical 
Specifications for the ventilation filter testing program, the control 
room ventilation system, the auxiliary building ventilation system, and 
the fuel handling building ventilation system. The amendments also 
incorporate a selective implementation of the alternative source term.
    Date of issuance: February 27, 2004.
    Effective date: February 27, 2004, and shall be implemented within 
180 days from the date of issuance.
    Amendment Nos.: Unit 1--163; Unit 2--165.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 24, 2003 (68 FR 
37579).
    The August 8 and November 13, 2003, supplemental letters provided 
additional clarifying information, did not expand the scope of the 
application as originally noticed, and did not change the NRC staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 27, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 29, 2003, as supplemented 
January 12, 2004.
    Brief description of amendment: This amendment revises the 
Technical Specifications (TSs) references in the Surveillance 
Requirement (SR) 4.0.5 and associated Basis, and Bases 3/4.4.2, 3/
4.4.6, and 3/4.4.10. In the current plant TSs, the reference for 
inservice testing (IST) and inservice inspection (ISI) activities is 
the American Society of Mechanical Engineers Boiler and Pressure Vessel 
Code (ASME BPV Code), Section XI. The licensee proposed to reference 
the ASME Code for Operation and Maintenance of Nuclear Power Plants 
(ASME OM Code) and the ASME BPV Code, Section XI for IST activities and 
ISI activities respectively. These changes reflect the fact that the 
pump and valve testing requirements previously contained in Subsections 
IWP and IWV of the ASME BPV Code, Section XI, have been replaced by the 
requirements in the 1998 Edition of the ASME OM Code, 2000 Addenda, for 
the licensee's third 120-month IST interval. These TS changes are 
required to implement the IST program update in accordance with the 
requirements of 10 CFR.55a(f)(5)(ii). The licensee also proposed 
certain other language changes.
    Date of issuance: February 18, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 166.
    Facility Operating License No. NPF-12: Amendment revised the TSs.
    Date of initial notice in Federal Register: October 14, 2003 (68 FR 
59219). The supplemental letter provided clarifying information that 
was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 18, 2004.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear 
Plant, Unit 3, Limestone County, Alabama

    Date of application for amendments: October 1, 2003, as 
supplemented December 19, 2003.
    Description of amendment request: The amendment revised the safety 
limit minimum critical power ratio values in Technical Specification 
(TS) 2.1.1.2.
    Date of issuance: February 24, 2004.
    Effective date: February 24, 2004.
    Amendment No.: 246.
    Facility Operating License No. DPR-68: Amendment revised the TSs.
    Date of initial notice in Federal Register: October 28, 2003 (68 FR 
61481). The December 19, 2003, letter provided clarifying information 
that did not change the scope of the original request or the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 24, 2004.
    No significant hazards consideration comments received: No.


    Dated at Rockville, Maryland, this 8th day of March 2004.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 04-5596 Filed 3-15-04; 8:45 am]
BILLING CODE 7590-01-P