Biweekly Notice Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 18371-18380 [E6-5086]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 69 (Tuesday, April 11, 2006)]
[Notices]
[Pages 18371-18380]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E6-5086]


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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 March 17, 2006 to March 30, 2006. The last 
biweekly notice was published on March 28, 2006 (71 FR 15479).

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

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

[[Page 18372]]

provide a brief explanation of the bases for the contention and a 
concise statement of the alleged facts or expert opinion which support 
the contention and on which the petitioner/requestor intends to rely in 
proving the contention at the hearing. The petitioner/requestor must 
also provide references to those specific sources and documents of 
which the petitioner is aware and on which the petitioner/requestor 
intends to rely to establish those facts or expert opinion. The 
petition must include sufficient information to show that a genuine 
dispute exists with the applicant on a material issue of law or fact. 
Contentions shall be limited to matters within the scope of the 
amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff; (3) e-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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

    Date of amendment request: February 7, 2006.
    Description of amendment request: The proposed amendments would 
increase the allowed outage time from 72 hours to 7 days for the 
inoperability of the steam supply to the turbine-driven auxiliary 
feedwater (AFW) pump or the inoperability of the turbine-driven AFW 
pump under certain operating mode restrictions.
    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:

Criterion 1: Does the Proposed Amendment Involve a Significant Increase 
in the Probability or Consequences of an Accident Previously Evaluated?

    Response: No.
    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed amendment to MPS 2 and 3 [Millstone Power Station, 
Unit Nos. 2 and 3] TS [Technical Specification] 3.7.1.2 permits a 7 
day allowed outage time for the inoperability of the necessary steam 
supply to the turbine-driven AFW pump in Modes 1, 2, and 3, or for 
the inoperability of the turbine-driven AFW pump if the 
inoperability occurs in Mode 3 following a refueling outage, if Mode 
2 had not been entered. Extending the allowed outage time does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated because: 1) The proposed amendment 
does not represent a change to the system design, 2) the proposed 
amendment does not prevent the safety function of the AFW [system] 
from being performed since the redundant trains are required to be 
operable, 3) the proposed amendment does not alter, degrade, or 
prevent action described or assumed in any accident described in the 
MPS 2 and 3 FSARs [final safety analysis reports] from being 
performed since the other trains of AFW are required to be operable, 
4) the proposed amendment does not alter any assumptions previously 
made in evaluating radiological consequences, and 5) the proposed 
amendment does not affect the integrity of any fission product 
barrier. No other safety related equipment is affected by the 
proposed change. Therefore, this proposed amendment does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.

Criterion 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 create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment to MPS 2 and 3 TS 3.7.1.2 would allow a 7 
day allowed outage time for the inoperability of the necessary steam 
supply to the turbine-driven AFW pump in Modes 1, 2, and 3, or for 
the inoperability of the turbine-driven AFW pump if the 
inoperability occurs in Mode 3 following a refueling outage, if Mode 
2 had not been entered. Extending the allowed action time does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated because: 1) the proposed amendment 
does not represent a change to the system design, 2) the proposed 
amendment does not alter how equipment is operated or the ability of 
the system to deliver the required AFW flow, and 3) the

[[Page 18373]]

proposed amendment does not affect any other safety related 
equipment. Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

Criterion 3: Does the Proposed Amendment Involve a Significant 
Reduction in a Margin of Safety?

    Response: No.
    The proposed amendment does not involve a significant reduction 
in a margin of safety.
    The proposed amendment to MPS 2 and 3 TS 3.7.1.2 would allow a 7 
day allowed action time for the inoperability of the necessary steam 
supply to the turbine-driven AFW pump in Modes 1, 2, and 3. 
Extending the allowed action time does not involve a significant 
reduction in a margin of safety because: 1) There is a redundant 
steam supply to the turbine driven AFW pump, 2) the motor-driven AFW 
pumps are required to be operable when Mode 3 is entered, 3) the 
motor-driven AFW pumps can provide sufficient flow to remove decay 
heat and cool the unit to shutdown cooling system entry conditions 
from power operations, 4) the motor-driven AFW pumps are designed to 
supply sufficient water to remove decay heat with steam generator 
pressure at no load conditions to cool the unit to shutdown cooling 
entry conditions, 5) the proposed change does not change or 
introduce any new setpoints at which mitigating functions are 
initiated, 6) no changes to the design parameters of the AFW 
[system] are being proposed, and 7) no changes in system operation 
that would impact an established safety margin are being proposed by 
this change.
    The proposed amendment to MPS 2 and 3 TS 3.7.1.2 would also 
allow a 7 day allowed action time for the inoperability of the 
turbine-driven AFW pump if the inoperability occurs in Mode 3 
following a refueling outage, if Mode 2 had not been entered. 
Extending the allowed action time does not involve a significant 
reduction in a margin of safety because: (1) During a return to 
power operations following a refueling outage, decay heat is at its 
lowest levels, (2) the motor-driven AFW pumps are required to be 
operable when Mode 3 is entered, (3) the motor-driven AFW pumps can 
provide sufficient flow to remove decay heat and cool the unit to 
shutdown cooling system entry conditions from power operations, (4) 
the motor-driven AFW pumps are designed to supply sufficient water 
to remove decay heat with steam generator pressure at no load 
conditions to cool the unit to shutdown cooling entry conditions, 
(5) the proposed change does not change or introduce any new 
setpoints at which mitigating functions are initiated, (6) no 
changes to the design parameters of the AFW are being proposed, and 
(7) no changes in system operation that would impact an established 
safety margin are being proposed by this change
    Therefore, based on the above, 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: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Branch Chief: Darrell J. Roberts.

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

    Date of amendment request: March 1, 2006.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications to reconcile the 10 CFR Part 50 and 
10 CFR Part 72 criticality requirements for the loading and unloading 
of dry spent fuel storage canisters in the spent fuel pool (SFP).
    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) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The applicable accidents are the dropped fuel assembly and drop 
of the 100 ton spent fuel cask into the SFP. This amendment request 
does not change the fuel assemblies or any of the Part 50 
structures, systems, or components involved in fuel assembly or cask 
handling or any of the operations involved. Therefore, this 
amendment request does not affect the probability of an accident 
previously evaluated.
    The proposed change does not increase the consequences of an 
accident previously evaluated for the following reasons: there is no 
increase in radiological source terms for the fuel; there is no 
change to the SFP water level; subcriticality is maintained for 
normal and accident conditions for the spent fuel storage racks and 
for cask loading and unloading; and the same boron concentrations 
that were previously credited for the spent fuel storage racks are 
assumed in the criticality analysis for cask loading and unloading.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Handling of fuel assemblies and the NUHOMS[reg] spent fuel cask 
have been previously evaluated for Oconee. These activities lead to 
evaluation of the fuel handling accident (dropped fuel assembly) and 
drop of the 100 ton spent fuel cask onto spent fuel stored in the 
Oconee SFP. These elements of the license amendment request (LAR) 
are not new, and thus do not create the potential for new or 
different kinds of accidents.
    The new element of this LAR is the application of additional 
criticality controls (i.e., minimum burnup requirements for the fuel 
assemblies) beyond the 10 CFR 72 controls already in place for the 
NUHOMS[reg] spent fuel cask. However, application of such 
criticality controls is not a new activity for Oconee, since similar 
criticality controls are currently applied to the spent fuel storage 
racks. Fuel assembly misloading is not a new accident; as discussed 
in Enclosure 3, Section 6.5, fuel assembly misloading has been 
considered previously for the NUHOMS[reg] spent fuel cask and for 
the Oconee spent fuel pool racks. Furthermore, the criticality 
analysis for cask loading and unloading evaluates the same boron 
concentrations, moderator temperatures, and misloading scenario as 
previously evaluated for the spent fuel storage racks. The analysis 
demonstrates that a criticality accident does not occur under these 
conditions. It is concluded that the possibility of a criticality 
accident is not created since application of criticality controls is 
not new and the analysis demonstrates that criticality does not 
occur. More generally, this supports the conclusion that the 
potential for new or different kinds of accidents is not created.
    (3) Involve a significant reduction in a margin of safety.
    This LAR involves the application of additional criticality 
controls (minimum burnup requirements) to the 10 CFR 72 controls 
already in place for the NUHOMS[reg] spent fuel cask. The 
criticality analysis demonstrates subcriticality margins are 
maintained for normal and accident conditions consistent with 10 CFR 
50.68(b) and other NRC guidance. Margins previously established for 
Oconee's spent fuel storage racks are not altered. Therefore, this 
LAR does not result in a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Branch Chief: Evangelos C. Marinos.

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

    Date of amendment request: February 14, 2006.
    Description of amendment request: The proposed change will modify 
the Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specification (TS) 
Surveillance Requirement 4.6.1.1.a. Specifically, the proposed change 
will eliminate the requirement to verify containment isolation valves 
that are maintained locked, sealed, or otherwise secured closed from 
the monthly

[[Page 18374]]

position verification. The proposed change will result in reducing 
radiological exposure to Operations, Health Physics, and Security 
personnel.
    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 accident mitigation features of the plant for previously 
evaluated accidents are not affected by the proposed change. No 
changes are proposed to the physical components or to the 
containment isolation function.
    Repositioning of manual containment isolation valves is 
procedurally controlled and governed by the note that is contained 
in TS 3.6.3.1, Containment Isolation Valves, which allows opening 
locked or sealed closed valves on an intermittent basis. The valve 
position is tracked until it is restored to its original position 
(locked or deactivated position, as appropriate). While the valve 
remains open, an individual, in constant communication with the 
control room staff, is stationed at the valve. If an accident were 
to occur, the control room staff would direct the individual 
stationed at the valve to close the valve thereby precluding the 
release of radioactivity outside containment.
    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 change the design, method of 
operation, or configuration of the plant. The procedural controls 
that establish the ANO-2 containment valve program controls and 
include the administrative controls that are associated with the 
note in TS 3.6.3.1, ensure containment integrity is appropriately 
established such that no new or different types of accidents are 
created.
    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 proposed change does not change the design basis for any 
equipment in the plant. The proposed change will exclude 
verification of the normally locked, sealed, or otherwise secured 
closed valves, blind flanges, and the deactivated automatic valves; 
however, the administrative controls applied to these components 
ensure that containment integrity is maintained.
    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.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment requests: March 7, 2006.
    Description of amendment requests: The proposed amendments would 
modify the Technical Specifications (TS) of the units by expanding 
Section 5.5.2, Leakage Monitoring Program, to include the Liquid Waste 
Disposal System, the Waste Gas System, and the Post-Accident 
Containment Hydrogen Monitoring System. These systems are currently in 
the licensee's own leakage monitoring program but are not listed in TS 
Section 5.5.2. The licensee also proposed to make an editorial change 
to the section.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff reviewed the licensee's analysis, and 
performed its own as follows:

    (1) Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of any accident 
previously evaluated?
    No. The proposed change would only add the three subject systems 
to the listing in Section 5.5.2. The licensee is currently 
performing leakage monitoring of these systems under its own 
program. Leakage monitoring of these three systems, whether listed 
in the TS or not, does not have any impact on the initiation of any 
accident previously analyzed, or on the scenarios and radiological 
consequences of these accidents. Consequently, 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 proposed change is purely administrative, and does not 
involve any change to the design or operation of a system, 
structure, or component. Consequently, the proposed change leads to 
no possibility to create 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?
    No. The proposed change would not change any assumption, 
analysis method, calculation model, or acceptance criterion. 
Accordingly, the proposed change does not involve a significant 
reduction in a margin of safety.

    Based on the NRC staff's analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment requests involve no 
significant hazards consideration.
    Attorney for licensee: James M. Petro, Jr., Esquire, One Cook 
Place, Bridgman, MI 49106.
    NRC Branch Chief: L. Raghavan.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of amendment request: February 16, 2006.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) requirements related to steam 
generator (SG) tube integrity. The change is consistent with NRC-
approved Revision 4 to Technical Specification Task Force (TSTF) 
Standard Technical Specification Change Traveler, TSTF-449, ``Steam 
Generator Tube Integrity.'' The availability of this TS improvement was 
announced in the Federal Register on May 6, 2005 (70 FR 24126) as part 
of the consolidated line item improvement process (CLIIP).
    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 March 2, 
2005 (70 FR 10298) as part of the CLIIP. The licensee affirmed the 
applicability of the model NSHC determination in its application dated 
February 16, 2006.
    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 requires a SG Program that includes 
performance criteria that will provide reasonable assurance that the 
SG tubing will retain integrity over the full range of operating 
conditions (including startup, operation in the power range, hot

[[Page 18375]]

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.
    A SGTR [steam generator tube rupture] 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 operational LEAKAGE rate limits 
in the licensing basis plus the leakage rate associated with a 
double-ended rupture of a single tube is assumed.
    For other design basis accidents such as MSLB [main steam line 
break], rod ejection, and reactor coolant pump locked rotor the 
tubes are assumed to retain their structural integrity (i.e., they 
are assumed not to rupture). These analyses typically assume that 
primary to secondary leakage for all SGs is 1 gallon per minute or 
increases to 1 gallon per minute as a result of accident induced 
stresses. 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 to the TS 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 SG Program required by the proposed change to the TS. The 
program, defined by NEI [Nuclear Energy Institute] 97-06, Steam 
Generator Program Guidelines, includes a framework that incorporates 
a balance of prevention, inspection, evaluation, repair, and leakage 
monitoring. The proposed changes do not, therefore, significantly 
increase the probability of an accident previously evaluated.
    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 [500 gallons per 
day or 720 gallons per day] in any one SG, and that the reactor 
coolant activity levels of dose equivalent I-131 are at the TS 
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 TSs 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 TSs.
    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 an MSLB, rod ejection, or a reactor coolant pump 
locked rotor event, or other previously evaluated accident.

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

    The proposed performance based requirements are an improvement 
over the requirements imposed by the current technical 
specifications. Implementation of the proposed SG 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 SG 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.

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

    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 
isolate the radioactive fission products in the primary coolant from 
the secondary system. In summary, the safety function of an SG is 
maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical 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 SG Program to manage SG tube 
inspection, assessment, repair, and plugging. The requirements 
established by the SG Program are consistent with those in the 
applicable design codes and standards and are an improvement over 
the requirements in the current TSs.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the 
TS.

    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: 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: February 13, 2006.
    Description of amendment request: The proposed amendments would 
make miscellaneous administrative changes by revising Technical 
Specifications (TS) 3.0 ``Surveillance Requirement (SR) 
Applicability''; and TS Chapter 5.0, ``Administrative Controls''. The 
proposed changes will improve TS usability, conformance with the 
industry standard, NUREG-1431, ``Standard Technical Specifications, 
Westinghouse Plants'', Revision 3.0 (NUREG-1431) and accuracy.
    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 administrative changes 
to the Prairie Island Nuclear Generating Plant Technical 
Specifications as follows: Technical Specification 3.0, 
``Surveillance Requirement (SR) Applicability'', revise page headers 
and correct capitalization; and Technical Specification Chapter 5.0, 
``Administrative Controls'', correct Topical Report numbers and make 
format corrections.
    The proposed changes are administrative and do not affect plant 
operation maintenance or testing. These changes do not affect any 
plant systems which are accident initiators and thus these 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 administrative changes 
to the Prairie Island Nuclear Generating Plant Technical 
Specifications as follows: Technical Specification 3.0, 
``Surveillance Requirement (SR) Applicability'', revise page headers 
and

[[Page 18376]]

correct capitalization; and Technical Specification Chapter 5.0, 
``Administrative Controls'', correct Topical Report numbers and make 
format corrections.
    The proposed changes are administrative and thus do not create 
new failure modes or mechanisms and do not generate new accident 
precursors. 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 administrative changes 
to the Prairie Island Nuclear Generating Plant Technical 
Specifications as follows: Technical Specification 3.0, 
``Surveillance Requirement (SR) Applicability'', revise page headers 
and correct capitalization; and Technical Specification Chapter 5.0, 
``Administrative Controls'', correct Topical Report numbers and make 
format corrections.
    The proposed Technical Specification changes are administrative 
and do not affect plant operation, maintenance or testing. 
Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: 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: February 16, 2006.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) requirements related to steam 
generator tube integrity. The change is consistent with NRC-approved 
Revision 4 to Technical Specification Task Force (TSTF) Standard 
Technical Specification Change Traveler, TSTF-449, ``Steam Generator 
Tube Integrity.'' The availability of this TS improvement was announced 
in the Federal Register on May 6, 2005 (70 FR 24126) as part of the 
consolidated line item improvement process (CLIIP).
    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 March 2, 
2005 (70 FR 10298) as part of the CLIIP. The licensee affirmed the 
applicability of the model NSHC determination in its application dated 
February 16, 2006.
    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:

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

    The proposed change requires a SG Program that includes 
performance criteria that will provide reasonable assurance that the 
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.
    A 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 operational leakage rate limits in the licensing basis plus the 
leakage rate associated with a double-ended rupture of a single tube 
is assumed.
    For other design basis accidents such as MSLB, rod ejection, and 
reactor coolant pump locked rotor the tubes are assumed to retain 
their structural integrity (i.e., they are assumed not to rupture). 
These analyses typically assume that primary to secondary leakage 
for all SGs is 1 gallon per minute or increases to 1 gallon per 
minute as a result of accident induced stresses. 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 to the TS 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 SG Program required by the proposed change to the TS. 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 proposed 
changes do not, therefore, significantly increase the probability of 
an accident previously evaluated.
    The consequences of design basis accidents are, in part, 
functions of the dose equivalent 1-131 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 1-131 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 [500 gallons per 
day or 720 gallons per day] in any one SG, and that the reactor 
coolant activity levels of dose equivalent 1-131 are at the TS 
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 TSs 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 TSs.
    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 an MSLB, rod ejection, or a reactor coolant pump 
locked rotor event, or other previously evaluated accident.

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

    The proposed performance based requirements are an improvement 
over the requirements imposed by the current technical 
specifications. Implementation of the proposed SG 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 SG 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.

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

    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 
isolate the

[[Page 18377]]

radioactive fission products in the primary coolant from the 
secondary system. In summary, the safety function of an SG is 
maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical 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 SG Program to manage SG tube 
inspection, assessment, repair, and plugging. The requirements 
established by the SG Program are consistent with those in the 
applicable design codes and standards and are an improvement over 
the requirements in the current TSs.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the 
TS. Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

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

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

    Date of amendment request: February 21, 2006.
    Description of amendment request: The amendment would revise 
Technical Specification 5.5.9, ``Steam Generator (SG) Tube Surveillance 
Program,'' to exclude portions of the SG tube below the top of the 
tubesheet in the SGs from periodic tube inspections based on the 
application of structural analysis and leak rate evaluation results to 
re-define the primary-to-secondary 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 proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The proposed change 
that alters the steam generator inspection criteria does not have a 
detrimental impact on the integrity of any plant structure, system, 
or component that initiates an analyzed event. The proposed change 
will not alter the operation of, or otherwise increase the failure 
probability of any plant equipment that initiates an analyzed 
accident.
    Of the applicable accidents previously evaluated, the limiting 
transients[,] with respect to the proposed [change] to the steam 
generator tube inspection criteria, are the steam generator tube 
rupture (SGTR) event and the steam line break (SLB) accident.
    During the SGTR event, the required structural integrity margins 
of the steam generator tubes will be maintained by the presence of 
the steam generator tubesheet. Steam generator tubes are 
hydraulically expanded in the tubesheet area. Tube rupture in tubes 
with cracks in the tubesheet is precluded by the constraint provided 
by the tubesheet. This constraint results from the hydraulic 
expansion process, thermal expansion mismatch between the tube and 
the tubesheet[,] and from the differential pressure between the 
primary and secondary side [of the steam generator]. Based on this 
design, the structural margins against burst, discussed in 
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR 
[Pressurized-Water Reactor] Steam Generator Tubes,'' are maintained 
for both normal and postulated accident conditions.
    The proposed change does not affect other systems, structures, 
components or operational features. Therefore, the proposed [change 
results] in no significant increase in the probability [or] the 
occurrence of a[n] SGTR accident.
    At normal operating pressures, leakage from primary water stress 
corrosion cracking (PWSCC) [of a tube] below the proposed inspection 
depth is limited by both the tube-to-tubesheet crevice and the 
limited crack opening permitted by the tubesheet constraint. 
Consequently, negligible normal operating leakage is expected from 
cracks within the tubesheet region. The consequences of an SGTR 
event are affected by the primary-to-secondary leakage flow during 
the event. Primary-to-secondary leakage flow through a postulated 
ruptured tube is not affected by the proposed change since the 
tubesheet enhances the tube integrity in the region of the hydraulic 
expansion by precluding tube deformation beyond its initial 
hydraulically expanded outside diameter.
    The probability of an SLB is unaffected by the potential failure 
of a steam generator tube as this failure is not an initiator for an 
SLB.
    The consequences of an SLB are also not significantly affected 
by the proposed change. During an SLB accident, the reduction in 
pressure above the tubesheet on the secondary side of the steam 
generator creates a uniformly distributed axial (out of plane) load 
on the tubesheet due to the reactor coolant system pressure on the 
primary [side] of the tubesheet. The resulting bending action causes 
contraction of the tube holes below the tubesheet neutral axis, 
adding to the constraint of the tubes in the tubesheet, thereby 
further restricting primary-to-secondary leakage.
    Primary-to-secondary leakage from tube degradation in the 
tubesheet area during the limiting accident (i.e., an SLB) is 
limited by flow restrictions resulting from the crack and tube-to-
tubesheet contact pressures that provide a restricted leakage path 
above the indications and also limit the degree of potential crack 
face opening as compared to free span indications. The primary-to-
secondary leak rate from tube degradation in the tubesheet region 
during postulated SLB accident conditions will be no more than twice 
that allowed during normal operating conditions when the pressure 
boundary is relocated [by the amendment] to the lesser of the H* or 
B* [tubesheet inspection] depths. Since normal operating leakage 
would be limited to 300 gpd [gallons per day] (0.2 gpm [gallons per 
minute]) through any one steam generator per TS 3.4.13, ``RCS 
[Reactor Coolant System] Operational leakage,'' the associated 
accident condition leak rate, assuming all leakage to be from lower 
tubesheet indications, would be limited to 150 gpd per steam 
generator. This value is well within the assumed accident leakage 
rate of 1.0 gpm discussed in WCGS [(Wolf Creek Generating Station)] 
Updated Safety Analysis Report, Table 15.1-3, ``Parameters Used in 
Evaluating the Radiological Consequences of a Main Steam Line 
Break.'' Therefore, the consequences of an SLB accident remain 
unaffected.
    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 accident from any accident previously evaluated?
    Response: No.
    The proposed change does not introduce any new equipment, create 
new failure modes for existing equipment, or create any new limiting 
single failures. Plant operation will not be altered, and all safety 
functions will continue to perform as previously assumed in accident 
analyses. [Excluding portions of the tube below the proposed 
tubesheet inspection depths does not introduce a new or different 
kind of accident to the steam generator tube because the required 
structural margins of the tubes for both normal and accident 
conditions are maintained.] 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 proposed [change maintains] the required structural margins 
of the steam generator tubes for both normal and accident 
conditions. Nuclear Energy Institute (NEI) 97-06, ``Steam Generator 
Program Guidelines,'' and RG 1.121, ``Bases for Plugging Degraded 
PWR Steam Generator Tubes,'' are used as the bases in the 
development of the tubesheet inspection depth methodology for 
determining that steam generator tube integrity considerations

[[Page 18378]]

are maintained within acceptable limits. RG 1.121 describes a method 
acceptable to the NRC for meeting General Design Criteria (GDC) 14, 
``Reactor coolant pressure boundary,'' GDC 15, ``Reactor coolant 
system design,'' GDC 31, ``Fracture prevention of reactor coolant 
pressure boundary,'' and GDC 32, ``Inspection of reactor coolant 
pressure boundary,'' by reducing the probability and consequences of 
a[n] SGTR. RG 1.121 concludes that by determining the limiting safe 
conditions for tube wall degradation[,] the probability and 
consequence of a[n] SGTR are reduced. This RG uses safety factors on 
loads for tube burst that are consistent with the requirements of 
Section III of the American Society of Mechanical Engineers (ASME) 
Code.
    For axially oriented cracking located within the tubesheet, tube 
burst is precluded due to the presence of the tubesheet. For 
circumferentially oriented cracking, Westinghouse letter LTR-CDME-
05-209-P, ``Steam Generator Tube Alternate Repair Criteria for the 
Portion of the Tube Within the Tubesheet at the Wolf Creek 
Generating Station,'' [provided in the application,] defines a 
length of degradation-free expanded tubing that provides the 
necessary resistance to tube pullout due to the pressure induced 
forces, with applicable safety factors applied. Application of the 
limited tubesheet inspection depth criteria will preclude 
unacceptable primary-to-secondary leakage during all plant 
conditions. The methodology for determining leakage provides for 
large margins between calculated and actual leakage values in the 
proposed limited tubesheet inspection depth criteria.
    Therefore, the proposed [change does not] involve a significant 
reduction in any margin [of] safety.

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

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) The 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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

    Date of application for amendment: February 25, 2005.
    Brief description of amendment: The amendment deleted the reporting 
requirement in the Facility Operating License (FOL) related to 
reporting violations of other requirements in the operating license.
    Date of issuance: February 24, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 172.
    Facility Operating License No. NPF-62: The amendment revised the 
FOL.
    Date of initial notice in Federal Register: April 26, 2005 (70 FR 
21450).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 24, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 25, 2005.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to exclude the containment purge valve leakage 
rates from the summation of secondary containment bypass leakage rates.
    Date of issuance: March 21, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of the date of issuance.
    Amendment No.: 173.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 26, 2005 (70 FR 
21451).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 21, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 13, 2005, as supplemented 
on November 29, 2005, and January 20 and February 13, 2006.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 1.1, ``Definitions,'' TS 3.4.13, ``RCS [reactor 
coolant system] Operational Leakage,'' TS 5.5.9, ``Steam Generator Tube 
Surveillance Program,'' and TS 5.6.9, ``Steam Generator [SG] Tube 
Inspection Report,'' and add a new specification (TS 3.4.18) for SG 
Tube Integrity. The changes are consistent with TS Task Force (TSTF) 
Change TSTF-449, Revision 4, ``Steam Generator Tube Integrity.''
    Date of issuance: March 9, 2006.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: 278 and 255.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 6, 2005 (70 FR 
72669).
    The November 29, 2005, and January 20 and February 13, 2006, 
supplements provided additional information that clarified the 
application, did not expand the scope of