Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 145-159 [05-24669]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 1 (Tuesday, January 3, 2006)]
[Notices]
[Pages 145-159]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 05-24669]


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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 December 9, 2005 to December 21, 2005. The 
last biweekly notice was published on December 20, 2005 (70 FR 75489).

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

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

[[Page 146]]

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

[[Page 147]]

4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

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

    Date of amendment request: September 19, 2005.
    Description of amendment request: Pursuant to 10 CFR 50.90, Entergy 
Operations, Inc. hereby requests an Operating License amendment for 
Arkansas Nuclear One, Unit 2, to replace the existing steam generator 
(SG) tube surveillance program with that being proposed by the 
Technical Specifications Task Force (TSTF) in TSTF 449, Revision 4. 
Specifically, Technical Specification (TS) 1.1, Definitions; TS 3/
4.4.5, Steam Generators; TS 3.4.6.2, Reactor Coolant System Leakage; TS 
6.5.9, Steam Generator Tube Surveillance Program; and TS 6.6.7, Steam 
Generator Tube Surveillance Reports are being revised to incorporate 
the new Steam Generator Program of TSTF 449, Revision 4. The proposed 
changes are consistent with the Consolidated Line Item Improvement 
Process provided in the May 6, 2005, Federal Register Notice (70 FR 
24126).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change requires a Steam Generator Program that 
includes performance criteria that will provide reasonable assurance 
that the steam generator (SG) tubing will retain integrity over the 
full range of operating conditions (including startup, operation in 
the power range, hot standby, cooldown and all anticipated 
transients included in the design specification). The SG performance 
criteria are based on tube structural integrity, accident induced 
leakage, and operational leakage.
    The structural integrity performance criterion is:
    Structural integrity performance criterion: All in-service steam 
generator tubes shall retain structural integrity over the full 
range of normal operating conditions (including startup, operation 
in the power range, hot standby, and cool down and all anticipated 
transients included in the design specification) and design basis 
accidents. This includes retaining a safety factor of 3.0 against 
burst under normal steady state full power operation primary to 
secondary pressure differential and a safety factor of 1.4 against 
burst applied to the design basis accident primary to secondary 
pressure differentials. Apart from the above requirements, 
additional loading conditions associated with the design basis 
accidents, or combination of accidents in accordance with the design 
and licensing basis, shall also be evaluated to determine if the 
associated loads contribute significantly to burst or collapse. In 
the assessment of tube integrity, those loads that do significantly 
affect burst or collapse shall be determined and assessed in 
combination with the loads due to pressure with a safety factor of 
1.2 on the combined primary loads and 1.0 on axial secondary loads.
    The accident induced leakage performance criterion is:
    The primary to secondary accident induced leakage rate for any 
design basis accidents, other than a SG tube rupture, shall not 
exceed the leakage rate assumed in the accident analysis in terms of 
total leakage rate for all SGs and leakage rate for an individual 
SG. Leakage is not to exceed 1 gpm through any one SG.
    The operational leakage performance criterion is:
    The RCS operational primary to secondary leakage through any one 
SG shall be limited to <=150 gallons per day per SG.
    A steam generator tube rupture (SGTR) event is one of the design 
basis accidents that are analyzed as part of a plant's licensing 
basis. In the analysis of a SGTR event, a bounding primary to 
secondary leakage rate equal to the leakage rate associated with a 
double-ended rupture of a single tube is assumed.
    For other design basis accidents such as main steam line break 
(MSLB) and control element assembly (CEA) ejection, the tubes are 
assumed to retain their structural integrity (i.e., they are assumed 
not to rupture). The accident induced leakage criterion introduced 
by the proposed changes accounts for tubes that may leak during 
design basis accidents. The accident induced leakage criterion 
limits this leakage to no more than the value assumed in the 
accident analysis.
    The SG performance criteria proposed change identify the 
standards against which tube integrity is to be measured. Meeting 
the performance criteria provides reasonable assurance that the SG 
tubing will remain capable of fulfilling its specific safety 
function of maintaining reactor coolant pressure boundary integrity 
throughout each operating cycle and in the unlikely event of a 
design basis accident. The performance criteria are only a part of 
the Steam Generator Program required by the proposed change. The 
program, defined by NEI 97-06, Steam Generator Program Guidelines, 
includes a framework that incorporates a balance of prevention, 
inspection, evaluation, repair, and leakage monitoring.
    The consequences of design basis accidents are, in part, 
functions of the DOSE EQUIVALENT I-131 in the primary coolant and 
the primary to secondary LEAKAGE rates resulting from an accident. 
Therefore, limits are included in the plant technical specifications 
for operational leakage and for DOSE EQUIVALENT I-131 in primary 
coolant to ensure the plant is operated within its analyzed 
condition. The typical analysis of the limiting design basis 
accident assumes that primary to secondary leak rate after the 
accident is 1 gallon per minute with no more than 720 gallons per 
day in any one SG, and that the reactor coolant activity levels of 
DOSE EQUIVALENT I-131 are at the technical specification values 
before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current technical specifications and 
enhances the requirements for SG inspections. The proposed change 
does not adversely impact any other previously evaluated design 
basis accident and is an improvement over the current technical 
specifications.
    Therefore, the proposed change does not affect the consequences 
of a SGTR accident and the probability of such an accident is 
reduced. In addition, the proposed changes do not affect the 
consequences of other design basis events.
    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 performance based requirements are an improvement 
over the requirements imposed by the current technical 
specifications.
    Implementation of the proposed Steam Generator Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
result of the implementation of the Steam Generator Program will be 
an enhancement of SG tube performance. Primary to secondary leakage 
that may be experienced during all plant conditions will be 
monitored to ensure it remains within current accident analysis 
assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change does not impact any other 
plant system or component. The change enhances SG inspection 
requirements.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
also isolate the radioactive fission products in the primary coolant 
from the secondary system. In summary, the safety function of a SG 
is maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical

[[Page 148]]

condition of the tube. The proposed change does not affect tube 
design or operating environment. The proposed change is expected to 
result in an improvement in the tube integrity by implementing the 
Steam Generator Program to manage SG tube inspection, assessment, 
and plugging. The requirements established by the Steam Generator 
Program are consistent with those in the applicable design codes and 
standards and are an improvement over the requirements in the 
current technical specifications.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: September 19, 2005.
    Description of amendment request: Entergy Operations, Inc., 
proposes to amend Technical Specification (TS) 3.6.2.1, ``Containment 
Spray System,'' to allow a one-time extension of the allowable outage 
time (AOT) for the Containment Spray System (CSS) from 72 hours to a 
maximum of 7 days, to be used once for each train or, at most, two 
times during fuel cycles 18 and 19. The proposed change is intended to 
provide flexibility in scheduling CSS maintenance activities, reduce 
refueling outage duration, and improve the availability of CSS 
components important to safety during plant shutdowns.
    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 TS change does not affect the design, operational 
characteristics, function or reliability of the CSS.
    The CSS is primarily designed to mitigate the consequences of a 
Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB). The 
requested change does not affect the assumption used in the 
deterministic LOCA or MSLB analyses.
    The duration of a TS AOT is determined considering that there is 
a minimal possibility that an accident will occur while a component 
is removed from service. A risk informed assessment was performed 
which concluded that the increase in plant risk is small and 
consistent with the guidance contained in Regulatory Guide 1.177 
[``An Approach for Plant-Specific Risk-Informed Decisionmaking: 
Technical Specifications''].
    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 previously evaluated?
    Response: No.
    The proposed change does not involve a change in the design, 
configuration, or method of operation of the plant that could create 
the possibility of a new or different kind of accident. The proposed 
change extends the AOT currently allowed by the TS to 7 days.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The Containment Heat Removal System (CHRS) consists of the CSS 
and the Containment Cooling System (CCS). The CHRS functions to 
rapidly reduce the containment pressure and temperature after a 
postulated LOCA or MSLB accident by removing thermal energy from the 
containment atmosphere. The CHRS also assists in limiting off-site 
radiation levels by reducing the pressure differential between the 
containment atmosphere and the outside atmosphere, thereby reducing 
the driving force for leakage of fission products from the 
containment.
    The CHRS is designed so that either both trains of the CSS, or 
one train of CSS and one train of CCS will provide adequate heat 
removal to attenuate the post-accident pressure and temperature 
conditions imposed upon the containment following a LOCA or MSLB.
    The proposed change includes administrative controls that will 
be established to ensure one train of CSS and one train of CCS will 
be available during the extended CSS AOT.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: David Terao.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of amendment request: October 18, 2005.
    Description of amendment request: The proposed amendment would 
revise applicability requirements related to single control rod 
withdrawal allowances in shutdown modes.
    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 special operation allowances do not 
involve the modification of any plant equipment or affect basic 
plant operation. The relevant design basis analyses are associated 
with refueling operations. The refueling interlocks are designed to 
back up procedural core reactivity controls during refueling 
operations to prevent an inadvertent criticality during refueling 
operations. The relaxations proposed in relocating and revising 
single controlrod withdrawal allowances during the Refueling MODE 
with the reactor vesselhead fully tensioned, to the proposed special 
operations allowances consistent with NUREG-1433 recommendations, 
will not increase the probability of an accident compared to a 
withdrawal of a rod while in Refueling MODE with the reactor vessel 
head removed. This is because the proposed special operations will 
allow the withdrawal of only one control rod at a time while 
requiring the one-rod-out interlock to be OPERABLE and other 
requirements imposed to ensure that all other rods remain fully 
inserted. This requirement coupled with the reactivity margin 
requirement for the most reactive rod fully withdrawn or removed, is 
adequate to prevent inadvertent criticality when a single rod is 
withdrawn for maintenance or testing. As such, there is no 
significant increase in the probability of an accident previously 
evaluated. Since no criticality is assumed to occur, the 
consequences of analyzed events are therefore not affected. 
Therefore, the proposed change does not involve a significant 
increase in the consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The proposed change does not involve any physical 
alteration of existing plant equipment or the installation of new 
equipment. The basic operation of installed equipment is unchanged 
and no new accident initiators or failure modes are introduced as a 
result of these changes. The methods governing plant operation and

[[Page 149]]

testing remain consistent with current safety analysis assumptions. 
These changes do not adversely affect existing plant safety margins 
or the reliability of the equipment assumed to operate in the safety 
analysis. The requirements imposed during these Special Operations 
ensure the existing analyses and equipment operating conditions 
remain bounding. 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 margin of safety is not reduced because the 
proposed requirements offer similar protection to those imposed 
during normal refueling activities. The proposed special operation 
allowances do not involve the modification of any plant equipment or 
affect basic plant operation. The proposed allowances limit the 
withdrawal of only one control rod at a time. This allowance is 
controlled by the reactor mode switch in the refuel position, or 
other precautions to prevent the withdrawal or removal of more than 
one rod and the requirement that adequate reactivity margin be 
maintained. These requirements are adequate to prevent an 
inadvertent criticality. These changes do not adversely affect 
existing plant safety margins or the reliability of the equipment 
assumed to operate in the safety analysis. As such, there are no 
changes being made to safety analysis assumptions, safety limits or 
safety system settings that would adversely affect plant safety as a 
result of the proposed change. Therefore, the proposed change does 
not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J.M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts, 02360-5599.
    NRC Branch Chief: Richard Lauder.

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

    Date of amendment request: January 10, 2005.
    Description of amendment request: The proposed change will delete 
the License Conditions concerning emergency core cooling system pump 
suction strainers from Appendix C of the Limerick Generating Station, 
Unit No. 1 Facility Operating License that were added by Amendment No. 
128.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No. The proposed change is administrative in nature. 
The proposed change does not involve the modification of any plant 
equipment nor does it affect basic plant operation. The proposed 
change will have no impact on any safety related structures, systems 
or components. The License Conditions proposed for deletion pertain 
to actions that have been completed and are obsolete, or involve 
activities that are controlled in accordance with other regulatory 
processes, i.e., 10 CFR 50.59 and 10 CFR 50.65.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The proposed change is administrative in nature. 
The proposed change has no impact on the design, function or 
operation of any plant structure, system or component and does not 
affect any accident analyses. The License Conditions in Appendix C 
can be deleted because they are obsolete or involve activities that 
are controlled in accordance with other regulatory processes.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The proposed change is administrative in nature, 
does not negate any existing requirement, and does not adversely 
affect existing plant safety margins or the reliability of the 
equipment assumed to operate in the safety analysis. As such, there 
is no change being made to safety analysis assumptions, safety 
limits or safety system settings that would adversely affect plant 
safety as a result of the proposed change. Margins of safety are 
unaffected by deletion of the License Conditions.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Brad Fewell, Assistant General Counsel, 
Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 
19348.
    NRC Branch Chief: Darrell J. Roberts.

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

    Date of amendment request: September 29, 2005.
    Description of amendment request: The proposed amendment would 
eliminate operability requirements for Secondary Containment, Secondary 
Containment Isolation Valves, the Standby Gas Treatment System, and 
Secondary Containment Isolation Instrumentation when handling 
irradiated fuel that has decayed for 24 hours since critical reactor 
operations and when performing Core Alterations. Similar technical 
specification relaxations are proposed for the Control Room Emergency 
Filter System and its initiation instrumentation after a decay period 
of 7 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment involves implementation of the 
Alternative Source Term (AST) for the fuel handling accident (FHA) 
at Cooper Nuclear Station (CNS). There are no physical design 
modifications to the plant associated with the proposed amendment. 
The FHA AST calculation does not impact the initiators of an FHA in 
any way.
    The changes also do not impact the initiators for any other 
design[-]basis accident (DBA) or events. Therefore, because DBA 
initiators are not being altered by adoption of the AST analyses the 
probability of an accident previously evaluated is not affected.
    With respect to consequences, the only previously evaluated 
accident that could be affected is the FHA. The AST is an input to 
calculations used to evaluate the consequences of the accident, and 
does not, in and of itself, affect the plant response or the actual 
pathways to the environment utilized by the radiation/activity 
released by the fuel. It does, however, better represent the 
physical characteristics of the release, so that appropriate 
mitigation techniques may be applied. For the FHA, the AST analyses 
demonstrate acceptable doses that are within regulatory limits after 
24 hours of radioactive decay since reactor shutdown, without credit 
for Secondary Containment, the Standby Gas Treatment System, 
Secondary Containment Isolation Valves, or Secondary Containment 
Isolation Instrumentation, and that the Control Room Emergency 
Filter System (CREFS) and CREFS Instrumentation need not be credited 
after a 7[-]day period of decay. Therefore, the consequences of an

[[Page 150]]

accident previously evaluated are not significantly increased.
    Based on the above conclusions, this proposed amendment does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment does not involve a physical alteration of 
the plant. No new or different types of equipment will be installed 
and there are no physical modifications to existing equipment 
associated with the proposed changes. The proposed changes to the 
control of Engineered Safety Features during handling of irradiated 
fuel do not create new initiators or precursors of a new or 
different kind of accident. New equipment or personnel failure modes 
that might initiate a new type of accident are not created as a 
result of the proposed amendment.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously analyzed.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed amendment is associated with the implementation of 
a new licensing basis for the CNS FHA. Approval of this change from 
the original source term to an AST derived in accordance with the 
guidance of Regulatory Guide (RG) 1.183 is being requested. The 
results of the FHA analysis, revised in support of the proposed 
license amendment, are subject to revised acceptance criteria. The 
AST FHA analysis has been performed using conservative 
methodologies, as specified in RG 1.183. Safety margins have been 
evaluated and analytical conservatism has been utilized to ensure 
that the analysis adequately bounds the postulated limiting event 
scenario. The dose consequences of the limiting FHA remain within 
the acceptance criteria presented in 10 CFR 50.67, the Standard 
Review Plan, and RG 1.183.
    The proposed changes continue to ensure that the doses at the 
Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) 
boundary, as well as the Control Room, are within the corresponding 
regulatory limits. For the FHA, RG 1.183 conservatively sets the EAB 
and LPZ limits below the 10 CFR 50.67 limit, and sets the Control 
Room limit consistent with 10 CFR 50.67.
    Since the proposed amendment continues to ensure the doses at 
the EAB, LPZ and Control Room are within corresponding regulatory 
limits, the proposed license amendment does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: October 12, 2005.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 3.4.9, ``RCS [reactor 
coolant system] Pressure and Temperature (P/T) Limits,'' curves 3.4.9-
1, ``Pressure/Temperature Limits for Non-Nuclear Heatup or Cooldown 
Following Nuclear Shutdown,'' 3.4.9-2, ``Pressure/Temperature Limits 
for Inservice Hydrostatic and Inservice Leakage Tests, and 3.4.9-3, 
``Pressure/Temperature Limits for Criticality,'' to remove the cycle 
operating restriction and replace it with a limitation of 30 effective 
full-power years (EFPY).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The 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 fluence evaluation 
for the P/T curves for 30 EFPY was performed using the NRC-approved 
Radiation Analysis Modeling Application (RAMA) fluence methodology. 
The curves generated from this method 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 an NRC accepted means of ensuring protection against 
brittle reactor vessel fracture, and compliance with 10 CFR 50 
Appendix G. The curves are the same as approved in Amendment Number 
204, CNS is only requesting to remove the one cycle limitation and 
limit their use to 30 EFPY based on the shift in the Adjusted 
Reference Temperature (ART) using the new fluence values. 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?
    Response: No.
    The proposed change updates existing P/T operating limits to 
correspond to the current NRC guidance. The proposed TS change 
extends the use of the current, NRC-approved P/T curves beyond the 
end of Cycle 23 to 30 EFPY. The proposed change does not involve a 
physical change to the plant, add any new equipment or any new mode 
of operation. These TS changes demonstrate compliance with the 
brittle fracture requirements of 10 CFR 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 a 
margin of safety?
    Response: No.
    The proposed change revises the existing CNS P/T curves to limit 
their use to 30 EFPY based on fluence calculation using the NRC-
approved Radiation Analysis Modeling Application (RAMA) fluence 
methodology. The curves have not been recalculated. Limiting the use 
of the P/T curves to 30 EFPY, based on the recalculation of the 
fluence per the NRC-approved (RAMA) fluence methodology does not 
affect a margin of safety. These changes do not affect any system 
used to mitigate accidents or transients.
    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 C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: David Terao.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of amendment request: September 16, 2005.
    Description of amendment request: The proposed amendment would 
revise the surveillance requirements (SRs) for the emergency Diesel 
Generators (EDGs) to provide more margin to the acceptance criterion. 
The new SR

[[Page 151]]

acceptance criterion will allow the EDG frequency to be within 2 percent of the rated value. The current acceptance limit is 
nominally 1 percent of rated frequency.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change. The EDG are not an initiator of any 
accident previously evaluated. As a result, the probability of any 
accident previously evaluated is not significantly increased. The 
consequences of any accident previously evaluated are not increased, 
as the EDG will continue to meet their safety function, as specified 
in the accident analysis, in a highly reliable manner.
    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.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. The changes do not alter assumptions made in the safety 
analysis for the EDG performance. The proposed changes remain 
consistent with the safety analysis assumptions (e.g., UFSAR Section 
8.3.1.4).
    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 change revises the acceptance criterion for EDG 
Surveillances to match that in the NRC's guidelines (Safety Guide 9) 
and the Improved Standard Technical Specifications (NUREG-1433, Rev 
3). Because the EDG can perform to the specified acceptance 
criterion as stated in the UFSAR Section 8.3.1.4; the EDG will 
continue to meet their specified safety function in the safety 
analysis, in a highly reliable manner.
    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: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Branch Chief: L. Raghavan.

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

    Date of amendment request: November 21, 2005.
    Description of amendment request: The proposed amendments to 
Prairie Island Nuclear Generating Plant (PINGP) Units 1 and 2 Operating 
Licenses, would allow extension of the Completion Time associated with 
Technical Specification (TS) 3.8.1 Required Action B4, from 7 days to 
14 days and for concomitant TS changes. The proposed amendment would 
also allow online performance of emergency diesel generator maintenance 
activities that are currently performed during refueling outages, to 
provide additional flexibility.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    This license amendment request proposes Technical Specification 
changes to extend the Technical Specification 3.8.1, ``AC Sources-
Operating,'' Completion Time for an inoperable emergency diesel 
generator to 14 days. These changes allow an emergency diesel 
generator to be inoperable for 7 days more than Technical 
Specification 3.8.1 currently provides. A minor format correction on 
the Technical Specification 3.8.1 Actions Table is also proposed.
    The emergency diesel generators are safety related components 
which provide backup electrical power supply to the onsite 
Safeguards Distribution System. The emergency diesel generators are 
not accident initiators, thus allowing an emergency diesel generator 
to be inoperable for an additional 7 days for performance of 
maintenance or testing does not increase the probability of a 
previously evaluated accident.
    Deterministic and probabilistic risk assessments evaluated the 
effect of the proposed Technical Specification changes on the 
availability of an electrical power supply to the plant emergency 
safeguards features systems. These assessments concluded that the 
proposed Technical Specification changes do not involve a 
significant increase in the risk of power supply unavailability.
    The plant emergency safeguards features systems consist of two 
trains for 100% redundancy within each unit. Accident analyses 
demonstrate that only one emergency safeguards features train is 
required for accident mitigation. Thus, with one train inoperable 
the other train is capable of performing the required safety 
function. Design basis analyses are not required to be performed 
assuming extended loss of all power supplies to the plant emergency 
safeguards features systems. Thus this change does not involve a 
significant increase in the consequences of a previously analyzed 
accident.
    The Technical Specification format correction is an 
administrative change and does not involve a significant increase in 
the probability or consequences of an accident.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This license amendment request proposes Technical Specification 
changes to extend the Technical Specification 3.8.1, ``AC Sources-
Operating,'' Completion Time for an inoperable emergency diesel 
generator to 14 days. These changes allow an emergency diesel 
generator to be inoperable for 7 days more than Technical 
Specification 3.8.1 currently provides. A minor format correction on 
the Technical Specification 3.8.1 Actions Table is also proposed.
    The proposed Technical Specification changes do not involve a 
change in the plant design, system operation, or procedures involved 
with the emergency diesel generators. The proposed changes allow an 
emergency diesel generator to be inoperable for additional time. 
There are no new failure modes or mechanisms created due to plant 
operation for an extended period to perform emergency diesel 
generator maintenance or testing. Extended operation with an 
inoperable emergency diesel generator does not involve any 
modification in the operational limits or physical design of plant 
systems. There are no new accident precursors generated due to the 
extended allowed Completion Time.
    The Technical Specification format correction is an 
administrative change and does not create the possibility of a new 
or different kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    This license amendment request proposes Technical Specification 
changes to extend the Technical Specification 3.8.1, ``AC Sources-
Operating,'' Completion Time for an inoperable emergency diesel 
generator to 14 days. These changes allow an emergency diesel 
generator to be inoperable for 7 days

[[Page 152]]

more than Technical Specification 3.8.1 currently provides. A minor 
format correction on the Technical Specification 3.8.1 Actions Table 
is also proposed.
    Currently, if an inoperable emergency diesel generator is not 
restored to operable status within 7 days, Technical Specification 
3.8.1 will require unit shutdown to MODE 3 within 6 hours and MODE 5 
within 36 hours. The proposed Technical Specification changes will 
allow steady state plant operation at 100% power for an additional 7 
days.

    There is some risk associated with continued operation for an 
additional 7 days with one emergency diesel generator inoperable. 
This risk is judged to be small and reasonable consistent with the 
risk associated with operations for 7 days with one emergency diesel 
generator inoperable as allowed by the current Technical 
Specifications. Specifically, the remaining operable emergency 
diesel generator and paths are adequate to supply electrical power 
to the onsite Safeguards Distribution System. An emergency diesel 
generator is required to operate only if both offsite power sources 
fail and there is an event which requires operation of the plant 
emergency safeguards features such as a design basis accident. The 
probability of a design basis accident occurring during this period 
is low.
    Deterministic and probabilistic risk assessments evaluated the 
effect of the proposed Technical Specification changes on the 
availability of an electrical power supply to the plant emergency 
safeguards features systems. These assessments concluded that the 
proposed Technical Specification changes do not involve a 
significant increase in the risk of power supply unavailability.
    There is also some risk associated with the Technical 
Specification unit shutdown evolutions. Plant load change evolutions 
require additional plant operations activities which introduce 
equipment challenges, increase the risk of plant trip and increase 
the risk for operational errors. Also unit shutdown does not remove 
the desirability of having emergency diesel generator backup for the 
4 kV safeguards buses, but rather places dependence on the operable 
4 kV bus by requiring operation of the residual heat removal system. 
Thus, possible additional risk associated with continuing operation 
an additional 7 days with an inoperable emergency diesel generator 
may be offset by avoiding the additional risk associated with unit 
shutdown.
    Therefore, based on the considerations given above, the proposed 
changes do not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Branch Chief: L. Raghavan.

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

    Date of amendment request: September 30, 2005.
    Description of amendment request: Omaha Public Power District 
(OPPD) proposes to change the licensing basis by replacing EMF-
2087(P)(A), Revision 0, ``SEM/PWR-98: ECCS [Emergency Core Cooling 
System] Evaluation Model for PWR [pressurized-water reactor] LBLOCA 
[large break loss-of-coolant accident] Applications,'' Siemens Power 
Corporation, June 1999, with the AREVA Topical Report EMF-2103(P)(A), 
``Realistic Large Break LOCA Methodology,'' Framatome ANP, Inc. in the 
Fort Calhoun Station, Unit 1 (FCS) Core Operating Limit Report (COLR). 
Currently, fuel for the FCS is supplied by AREVA. AREVA has performed 
an FCS-specific LBLOCA analysis using their realistic LBLOCA 
methodology for Cycle 24 and beyond.
    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 amendment replaces EMF-2087(P)(A), Revision 0, 
``SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications,'' 
Siemens Power Corporation, June 1999 (Reference 8.6 [of the 
licensee's amendment request]), with the AREVA Topical Report EMF-
2103(P)(A), ``Realistic Large Break LOCA Methodology,'' Framatome 
ANP, Inc. (Reference 8.1 [of the licensee's amendment request]) in 
the FCS COLR. AREVA Topical Report EMF-2103(P)(A) will also replace 
EMF-2087(P)(A) in OPPD topical report OPPD-NA-8303 (Reference 8.5 
[of the licensee's amendment request]). This amendment will allow 
the use of the RLBLOCA [realistic large break loss-of-coolant 
accident] methodology to perform the FCS LBLOCA analysis. The 
proposed amendment will not affect any previously evaluated 
accidents because they are analyzed using applicable NRC[-]approved 
methodologies to ensure all required safety limits are met.
    The proposed amendment does not affect any acceptance criteria 
for any postulated accidents or anticipated operational occurrences 
(AOOs) analyzed and listed in the FCS Updated Safety Analysis Report 
(USAR). The proposed change will not increase the likelihood of a 
malfunction of a structure, system or components (SSC) since the 
change does not involve operation of SSCs in a manner or 
configuration different from those previously evaluated.
    The results from the FCS RLBLOCA analysis have demonstrated the 
adequacy of the ECCS, and these results satisfy the regulatory 
criteria set forth in 10 CFR 50.46(b).
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not result in changes in the operation 
or overall configuration of the facility. The proposed amendment 
does not involve a change in the design function or the operation of 
SSCs involved. The proposed amendment does not involve the operation 
or configuration of the SSCs different from those previously 
analyzed. The proposed amendment to add the RLBLOCA methodology to 
the FCS COLR and OPPD topical report OPPD-NA-8303 (Reference 8.5 [of 
the licensee's amendment request]) does not create any new or 
different kind of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    AREVA has performed the RLBLOCA analysis for FCS and 
demonstrated that the Emergency Core Cooling System (ECCS) is 
adequate to mitigate the consequences of a[n] LBLOCA. The analysis 
has concluded that the acceptance criteria for the ECCS are met with 
significantly increased margins.
    All required safety limits will continue to be analyzed using 
methodologies approved by the Nuclear Regulatory Commission.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: October 5, 2005.
    Description of amendment request: The proposed amendment would 
delete

[[Page 153]]

requirements from the Technical Specifications (TSs) to maintain 
hydrogen recombiners and hydrogen and oxygen monitors. A notice of 
availability for this TS improvement using the consolidated line item 
improvement process was published in the Federal Register on September 
25, 2003 (68 FR 55416).
    Licensees were generally required to implement upgrades as 
described in NUREG-0737, ``Clarification of TMI [Three Mile Island] 
Action Plan Requirements,'' and Regulatory Guide (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 in 1979. 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 Title 10 of the 
Code of Federal Regulations (10 CFR) Section 50.44, ``Combustible gas 
control for nuclear 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 Nuclear Regulatory Commission (NRC) staff issued a notice of 
availability of a model no significant hazards consideration (NSHC) 
determination for referencing license amendment applications in the 
Federal Register on September 25, 2003 (68 FR 55416). The licensee 
affirmed the applicability of the model NSHC determination in its 
application dated October 5, 2005.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The 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 NRC has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen and oxygen monitors are no longer required to mitigate 
design-basis accidents and, therefore, the hydrogen monitors do not 
meet the definition of a safety-related component as defined in 10 
CFR 50.2. RG 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 NRC 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 NRC 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 th