Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 147-159 [E6-22492]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 1 (Wednesday, January 3, 2007)]
[Notices]
[Pages 147-159]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E6-22492]


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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 8, 2006 to December 21, 2006. The 
last biweekly notice was published on December 19, 2006 (71 FR 75987).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazard 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, Rulemaking, 
Directives and Editing Branch, Division of Administrative Services, 
Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the Commission's 
Public Document Room (PDR), located at One White Flint North, Public 
File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. The filing of requests for a hearing and petitions for leave 
to intervene is discussed below.
    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.

[[Page 148]]

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

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

    Date of amendments request: September 28, 2006.
    Description of amendment request: The proposed amendment would 
modify technical specification (TS) requirements of TS 3.8.3, ``Diesel 
Fuel Oil,'' to include a new Condition A with associated Required 
Action and Completion Time. The proposed Condition A allows the main 
fuel oil storage tank to be inoperable for up to 14 days for the 
purpose of performing inspection, cleaning, or repair activities.
    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 does not alter the assumption of the 
accident analyses or the Technical Specification Bases. The 
inclusion of provisions to permit internal inspection of the main 
fuel oil storage tank during plant operation does not impact the 
availability of the EDGs to perform their intended safety function. 
Furthermore, while the main fuel oil storage tank is out of service, 
the availability of on-site and off-site fuel oil sources ensures 
that an adequate supply of fuel oil remains available. 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 involve a physical change to the 
design of the Diesel Fuel Oil System, nor does it alter the 
assumptions of the accident analyses. The inclusion of provisions to 
permit internal inspection and cleaning of the main fuel oil storage 
tank during plant operation does not introduce any new failure 
modes. 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?

[[Page 149]]

    Response: No.
    The proposed change alters the method of operation of the Diesel 
Fuel Oil System. However, the availability of the EDGs to perform 
their intended safety function is not impacted and the assumptions 
of the accident analyses are not altered. 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Acting Branch Chief: D. Pickett.

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

    Date of amendment request: November 27, 2006.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) by relocating references to 
specific American Society for Testing and Materials (ASTM) standards 
for fuel oil testing to licensee-controlled documents and adding 
alternate criteria to the ``clear and bright'' acceptance test for new 
fuel oil.
    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 by a reference to a generic analysis published in the 
Federal Register on February 22, 2006 (71 FR 9179), which is presented 
below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of any accident previously 
evaluated?
    Response: No.
    The proposed changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. Requirements to perform testing in 
accordance with applicable ASTM standards are retained in the TS as 
are requirements to perform surveillances of both new and stored 
diesel fuel oil. Future changes to the licensee-controlled document 
will be evaluated pursuant to the requirements of 10 CFR 50.59, 
``Changes, tests and experiments,'' to ensure that such changes do 
not result in more than a minimal increase in the probability or 
consequences of an accident previously evaluated. In addition, the 
``clear and bright'' test used to establish the acceptability of new 
fuel oil for use prior to addition to storage tanks has been 
expanded to recognize more rigorous testing of water and sediment 
content. Relocating the specific ASTM standard references from the 
TS to a licensee-controlled document and allowing a water and 
sediment content test to be performed to establish the acceptability 
of new fuel oil will not affect nor degrade the ability of the 
emergency diesel generators (DGs) to perform their specified safety 
function. Fuel oil quality will continue to meet ASTM requirements.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not adversely 
affect the ability of structures, systems, and components (SSCs) to 
perform their intended safety function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of any accident previously evaluated. 
Further, the proposed changes do not increase the types and amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of any 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 changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. In addition, the ``clear and bright'' 
test used to establish the acceptability of new fuel oil for use 
prior to addition to storage tanks has been expanded to allow a 
water and sediment content test to be performed to establish the 
acceptability of new fuel oil. 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 requirements retained in the TS continue to require 
testing of the diesel fuel oil to ensure the proper functioning of 
the DGs.
    Therefore, the changes do 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 changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. Instituting the proposed changes will 
continue to ensure the use of applicable ASTM standards to evaluate 
the quality of both new and stored fuel oil designated for use in 
the emergency DGs. Changes to the licensee-controlled document are 
performed in accordance with the provisions of 10 CFR 50.59. This 
approach provides an effective level of regulatory control and 
ensures that diesel fuel oil testing is conducted such that there is 
no significant reduction in a margin of safety.
    The ``clear and bright'' test used to establish the 
acceptability of new fuel oil for use prior to addition to storage 
tanks has been expanded to allow a water and sediment content test 
to be performed to establish the acceptability of new fuel oil. The 
margin of safety provided by the DGs is unaffected by the proposed 
changes since there continue to be TS requirements to ensure fuel 
oil is of the appropriate quality for emergency DG use. The proposed 
changes provide the flexibility needed to improve fuel oil sampling 
and analysis methodologies while maintaining sufficient controls to 
preserve the current margins of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David G. Pettinari, Legal Department, 688 
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
    NRC Branch Chief: L. Raghavan.

Duke Power Company LLC, 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: April 11, 2006.
    Description of amendment request: The proposed amendments would 
(Item 1) revise the Technical Specifications (TSs) and delete the 
license conditions related to steam generator (SG) tube integrity and 
(Item 2) revise an organizational description in TS 5.2.1 that is 
solely administrative in nature and unrelated to the SG tube integrity 
TSs.
    The changes related to SG tube integrity are consistent with the 
consolidated line-item improvement process (CLIIP), Nuclear Regulatory 
Commission-approved Revision 4 to Technical Specification Task Force 
(TSTF) Standard TS Change Traveler, TSTF-449, ``Steam Generator Tube 
Integrity.''
    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:

[[Page 150]]

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
(Item 1) SG Tube Integrity
    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 (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 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 0.27 gallons per minute with no more 
than 135 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 (main steamline break), rod ejection, or a reactor coolant pump 
locked rotor event, or other previously evaluated accident.
(Item 2) Organization Description Revision in TS 5.2.1
    The proposed change revises an organizational description in TS 
5.2.1 to conform to an application for consent to the indirect transfer 
of control of the renewed facility operating licenses. The proposed 
change does not affect the operation of any equipment, and is solely 
administrative in nature; therefore, the proposed change has no impact 
on any accident probabilities or consequences.
Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated
(Item 1) SG Tube Integrity
    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.
(Item 2) Organization Description Revision in TS 5.2.1
    There are no new accident causal mechanisms created as a result of 
this proposed change. No changes are being made to the plant that will 
introduce any new accident causal mechanisms. This change is solely 
administrative in nature and does not impact any plant systems that are 
accident initiators; therefore, no new accident types are being 
created.
Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety
(Item 1) SG Tube Integrity
    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

[[Page 151]]

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.
(Item 2) Organization Description Revision in TS 5.2.1
    Margin of safety is related to confidence in the ability of the 
fission product barriers to perform their design functions during and 
following an accident situation. This proposed change is solely 
administrative in nature and does not affect the performance of the 
barriers. Consequently, no safety margins will be impacted.
    Attorney for licensee: Lisa F. Vaughn, Associate General Counsel 
and Managing Attorney, Duke Energy Carolinas, LLC, 526 South Church 
Street EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Evangelos C. Marinos.

Duke Power Company LLC, 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: November 16, 2006.
    Description of amendment request: The proposed amendments would 
authorize revision to revise the Updated Final Safety Analysis Report 
(UFSAR) to describe the flood protection measures for the auxiliary 
building.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. This License Amendment Request (LAR) proposes the use of a 
realistic seismic evaluation of the Auxiliary Building sprinkler 
system (high pressure service water) piping which demonstrates that 
these non-Category I (non-seismic) self-actuating sprinkler systems 
will not fail during a Maximum Hypothetical Earthquake (MHE) and 
clarifies Duke's commitment toward Auxiliary Building flood 
protection measures in the Updated Final Safety Analysis Report 
(UFSAR). The proposed change does not affect any Chapter 15 accident 
analyses. Operation in accordance with the amendment authorizing 
this change would not involve any accident initiation sequences or 
change the consequences of any accident analyzed.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. This LAR proposes the use of a realistic seismic evaluation 
of the Auxiliary Building sprinkler system (high pressure service 
water) piping which demonstrate that these non-Category I (non-
seismic) self-actuating sprinkler systems will not fail during a MHE 
and clarifies Duke's commitment toward Auxiliary Building flood 
protection measures in the UFSAR. Operation in accordance with this 
proposed amendment will not result in a change in the parameters 
governing plant operation and will not generate any new accident 
initiators. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    No. This LAR proposes the use of a realistic seismic evaluation 
of the Auxiliary Building sprinkler system (high pressure service 
water) piping, which demonstrates that these non-Category I (non-
seismic) self-actuating sprinkler systems will not fail during a MHE 
and clarifies Duke's commitment toward Auxiliary Building flood 
protection measures in the UFSAR. Operation in accordance with this 
proposed amendment will not result in a change in the parameters 
governing plant operation and will not affect any Chapter 15 
accident analyses. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, Charlotte, NC 28202.
    NRC Branch Chief: Evangelos C. Marinos.

Duke Power Company LLC, et al., 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: April 11, 2006.
    Description of amendment request: The proposed amendment would add 
Technical Specification (TS) Limiting Condition for Operation (LCO) 
3.0.8 to allow a delay time for entering a supported system TS when the 
inoperability is due solely to an inoperable snubber. The proposed 
changes are consistent with approval of TS Task Force (TSTF) change 
TSTF-372, Revision 4, ``Addition of LCO 3.0.8, Inoperability of 
Snubbers.''
    The NRC staff issued a notice of availability of a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination for referencing in license amendment applications in the 
Federal Register on November 24, 2004 (69 FR 68412).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an inoperable snubber if risk is assessed and managed. The 
postulated seismic event requiring snubbers is a low-probability 
occurrence and the overall TS system safety function would still be 
available for the vast majority of anticipated challenges. 
Therefore, the probability of an accident previously evaluated is 
not significantly increased, if at all. The consequences of an 
accident while relying on allowance provided by proposed LCO 3.0.8 
are no different than the consequences of an accident while relying 
on the TS required actions in effect without the allowance provided 
by proposed LCO 3.0.8. Therefore, the consequences of an accident 
previously evaluated are not significantly affected by this change. 
The addition of a requirement to assess and manage the risk 
introduced by this change will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

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

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to inoperable snubbers, if risk is 
assessed and managed, will not introduce new failure modes or 
effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns. Thus, this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.

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

    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber, if risk is assessed and managed. The

[[Page 152]]

postulated seismic event requiring snubbers is a low-probability 
occurrence and the overall TS system safety function would still be 
available for the vast majority of anticipated challenges. The risk 
impact of the proposed TS changes was assessed following the three-
tiered approach recommended in Regulatory Guide 1.177. A bounding 
risk assessment was performed to justify the proposed TS changes. 
This application of LCO 3.0.8 is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk. 
The net change to the margin of safety is insignificant. Therefore, 
this change does not involve a significant reduction in a margin of 
safety. \

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Lisa F. Vaughn, Associate General Counsel 
and Managing Attorney, Duke Energy Carolinas, LLC, 526 South Church 
Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Evangelos C. Marinos.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

    Date of amendment request: September 26, 2006.
    Description of amendment request: The proposed amendment would 
allow up to eight AREVA NP Inc. Modified Advanced Mark-BW(A) fuel 
assemblies containing M5 alloy to be placed in nonlimitng Braidwood 
Station, Unit No. 1 core regions for evaluation during Cycle 14, 15, 
and 16. The proposed amendment would also remove all references to 
Joseph Oat spent fuel storage racks that have been physically removed 
from the spent fuel pool.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS [technical specification] change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The AREVA Advanced Mark-BW(A) fuel is similar in design to the 
Westinghouse fuel that will be co-resident in the core. The Advanced 
Mark-BW(A) fuel assemblies are also similar in design to the 
Advanced Mark-BW assemblies using M5 alloy material for the 
cladding, structural tubing, and grids generically approved for use 
in Westinghouse 3- and 4-loop designed pressurized water reactors 
with 17 x 17 fuel rod arrays. The AREVA Advanced Mark-BW(A) fuel 
assemblies will be placed in nonlimiting regions (i.e., locations) 
of the core. The Cycle 14, 15, and 16 reload designs will meet all 
applicable design criteria. EGC [Exelon Generation Company, LLC] 
will use the NRC-approved standard reload design models and methods 
to demonstrate that all applicable design criteria will be met. 
Evaluations will be performed as part of the cycle specific reload 
safety analysis for the operation of the AREVA Advanced Mark-BW(A) 
fuel to confirm that the acceptance criteria of the existing safety 
analyses continue to be met. Operation of the AREVA Advanced Mark-
BW(A) fuel will not significantly increase the predicted 
radiological consequences of accidents postulated in the Updated 
Final Safety Analysis Report.
    The proposed change regarding removal of all references in TS to 
the Joseph Oat spent fuel racks is administrative and deletes 
unnecessary wording relating to equipment that is physically removed 
from the Braidwood Station spent fuel pool and therefore does not 
alter the design, configuration, operation, or function of any plant 
system, structure or component. As a result, the administrative 
change does not affect the outcome of any previously evaluated 
accidents.
    Based on the above discussion, the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The AREVA Advanced Mark-BW(A) fuel is similar in design to the 
Westinghouse fuel that will be co-resident in the core. The Advanced 
Mark-BW(A) fuel assemblies are also similar in design to the 
Advanced Mark-BW assemblies using M5 alloy material for the 
cladding, structural tubing, and grids generically approved for use 
in Westinghouse 3- and 4-loop designed pressurized water reactors 
with 17 x 17 fuel rod arrays. The Braidwood Station, Unit [No.] 1 
cores in which the fuel operates will be designed to meet all 
applicable design criteria and ensure that all pertinent licensing 
basis criteria are met. Demonstrated adherence to these standards 
and criteria precludes new challenges to components and systems that 
could introduce a new type of accident. The reload core designs for 
the cycles in which the AREVA Advanced Mark-BW(A) fuel will operate 
will demonstrate that the use of up to eight AREVA Advanced Mark-
BW(A) fuel assemblies in nonlimiting core regions (i.e., locations) 
is acceptable. The relevant design and performance criteria will 
continue to be met and no new single failure mechanisms will be 
created. The use of AREVA Advanced Mark-BW(A) fuel does not involve 
any alteration to plant equipment or procedures that would introduce 
any new or unique operational modes or accident precursors.
    The proposed change regarding removal of all references in TS to 
Joseph Oat spent fuel racks is administrative and deletes 
unnecessary wording relating to equipment that is physically removed 
from the Braidwood Station spent fuel pool and therefore does not 
alter the design, configuration, operation, or function of an plant 
system, structure or component. As a result, the administrative 
change does not create any new or different kind of accident.
    Based on this evaluation, the proposed change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    Operation of Braidwood Station, Unit [No.] 1 with up to eight 
AREVA Advanced Mark-BW(A) fuel assemblies in nonlimiting core 
regions (i.e., locations) does not change the performance 
requirements on any system or component such that any design 
criteria will be exceeded. The normal limits on core operation 
defined in the Braidwood Station TS will remain applicable for the 
use of up to eight AREVA Advanced Mark-BW(A) fuel assemblies during 
Cycles 14, 15, and 16. The reload core designs for the cycles in 
which the AREVA Advanced Mark-BW(A) fuel will operate will 
specifically evaluate any pertinent differences, including both 
mechanical design differences and the past irradiation history, 
between the AREVA Advanced Mark-BW(A) fuel product, and the 
Westinghouse fuel product that will be co-resident in the core. The 
use of up to eight AREVA Advanced Mark-BW(A) fuel assemblies will be 
specifically evaluated during the reload design process using reload 
design models and methods as approved by the NRC.
    The proposed change regarding removal of all references in TS to 
Joseph Oat spent fuel racks is administrative and deletes 
unnecessary wording relating to equipment that is physically removed 
from the Braidwood Station spent fuel pool and therefore does not 
alter the design, configuration, operation, or function of an plant 
system, structure or component. As a result, the administrative 
change does not affect the ability of any operable structure, 
system, or component to perform its designated safety function.
    Based on this evaluation, the proposed change does not involve a 
significant reduction in a margin of safety.

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

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: October 19, 2006.
    Description of amendment request: The proposed amendments would

[[Page 153]]

revise Technical Specification 4.6.2.1.d to allow the frequency of air 
or smoke flow testing of the containment spray nozzles to be reduced 
from 10 years to an activity-related frequency following maintenance 
that could cause a blockage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed change revises the surveillance frequency from 
once per 10 years to following activities that could result in 
nozzle blockage. The containment spray system nozzles are passive 
components and are not considered as an initiator of any analyzed 
event. The proposed change will not impact the ability of the 
containment spray system to mitigate the consequences of an 
accident. Industry experience indicates that containment spray 
systems of similar design are highly reliable and not susceptible to 
plugging due to the open design of the nozzles, the location of the 
nozzles high in the containment dome, and the corrosion resistant 
materials used for construction of the system. The alternative 
frequency of this surveillance has no impact on the probability of 
failure of associated active systems. Therefore, there is no 
significant increase in the probability or consequences of 
previously evaluated accidents due to the extended surveillance 
frequency.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed amendment provides an alternative frequency for 
performance of the spray nozzle surveillance test. The containment 
spray nozzles are used for accident mitigation only. Potential 
unidentified blockage of the containment spray nozzles will not 
result in the initiation of an accident. The change does not involve 
a physical alteration of the plant nor does it result in an 
operational condition different from that which has already been 
considered in the accident analyses. Therefore, the change does not 
create the possibility of a new or different kind of accident or 
malfunction from any accident previously evaluated.
    3. Does this change involve a significant reduction in margin of 
safety?
    No. The alternative frequency of spray nozzle testing has no 
significant impact on the consequences of any analyzed accident and 
does not significantly change the failure probability of any 
equipment that provides protection for the health and safety of the 
public. The containment spray system will continue to be capable of 
maintaining containment temperature and pressure below design 
values. Therefore, there is no 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Douglas V. Pickett (Acting).

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: October 19, 2006.
    Description of amendment request: The proposed amendments would 
revise various Technical Specifications (TSs) to address requirements 
that should have been changed as part of previously approved 
amendments. These amendments included TS changes regarding relocation 
of administrative requirements to licensee controlled programs such as 
the Topical Quality Assurance Report (TQAR), handling of recently 
irradiated fuel in accordance with TS Task Force change traveler TSTF-
51, and Auxiliary Feedwater Actuation System (AFAS) trip and bypass 
requirements. The proposed amendments also correct some typographical 
errors.
    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) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    These proposed license amendments require no plant hardware or 
operational modifications. The proposed changes either correct 
various administrative errors or incorporate changes that have been 
justified by previously approved license amendments and should have 
been made as part of those submittals. Therefore, operation of the 
facility in accordance with the proposed amendments would not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    No modifications to either plant hardware or operational 
procedures are required to support these proposed license 
amendments; hence, no new failure modes are created. The proposed 
changes either correct various administrative errors or incorporate 
changes that have been justified by previously approved license 
amendments and should have been made as part of those submittals. 
Therefore, operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The TS corrections proposed by these license amendments are 
administrative in nature in that they either correct typographical 
errors (e.g., letter dates and transient limits) or are justified by 
previous license amendments (i.e., relocation of administrative 
programs to the TQAR, the implementation of TSTF-51 for recently 
irradiated fuel, and correct inconsistencies introduced by AFAS trip 
and bypass requirements). Therefore, operation of the facility in 
accordance with the proposed amendment would 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: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Douglas V. Pickett (Acting).

Indiana Michigan Power Company (I&M), Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Unit 1 and 2, Berrien County, Michigan

    Date of amendment request: November 3, 2006.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TS) to reflect a proposed plant 
modification that will replace the reactor coolant system resistance 
temperature detectors (RTDs) and bypass piping with fast response 
thermowell detectors mounted directly in the primary loop piping. The 
specific TS requirements affected include the notes in Unit 2 TS 
surveillance requirement for channel calibration of the overtemperature 
differential temperature (OT[Delta]T) and overpower differential 
temperature (OP[Delta]T) reactor trip system functions. The proposed 
change also affects the Unit 1 and Unit 2 TS allowable values for 
OT[Delta]T and OP[Delta]T reactor trip systems functions.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 154]]

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 of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The resistance temperature detectors (RTD) bypass system is the 
hardware associated with Reactor Coolant System instrumentation 
having control, indication, and protection functions. The RTD bypass 
system is not considered a precursor to any previously analyzed 
accident. The system is relied upon to mitigate the consequences of 
some accidents. The new system replacing the RTD bypass system will 
perform the same control, indication, and protection functions, and, 
similarly, will not be considered a precursor to any accident. The 
capability of the system to mitigate the consequences of the 
previously analyzed accidents will not be significantly affected. 
Therefore, replacement of the existing RTD bypass system with the 
new system will not increase the probability of occurrence of an 
accident, and will not increase 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 replacement of the existing RTD bypass with the new system 
would not create new failure modes, and the replacement system is 
not an initiator of any new or different kind of accident. The 
proposed deletion of the note in Technical Specification (TS) 
Surveillance Requirement 3.3.1.15, and proposed changes to Allowable 
Values in TS Table 3.3.1-1 do not affect the interaction of the 
replacement system with any system whose failure or malfunction can 
initiate an 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.
    Margins of safety are established in the design of components, 
the configuration of components to meet certain performance 
parameters, and in the models and associated assumptions used to 
analyze the system's performance. The replacement system will 
continue to perform the same temperature detection function to the 
same level of reliability as defined in the Donald C. Cook Nuclear 
Plant Updated Final Safety Analysis Report. Therefore, the proposed 
amendment does not involve a significant reduction in a margin of 
safety.

    The Nuclear Regulatory Commission (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: Kimberly Harshaw, Esquire, One Cook Place, 
Bridgman, MI 49106.
    NRC Branch Chief: L. Raghavan.

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

    Date of amendment requests: December 14, 2006.
    Description of amendment requests: The proposed amendments would 
delete Section 2.G of the Diablo Canyon Power Plant Facility Operating 
Licenses, which requires reporting of violations of the requirements in 
Sections 2.C, 2.E, and 2.F of the Facility Operating License.
    The NRC staff issued a notice of opportunity to comment in the 
Federal Register on August 29, 2005 (70 FR 51098), on possible 
amendments to eliminate the license condition involving reporting of 
violations of other requirements (typically in License Condition 2.C) 
in the operating license, including a model safety evaluation and model 
no significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. The NRC staff subsequently 
issued a notice of availability of the model for referencing in license 
amendment applications in the Federal Register on November 4, 2005 (70 
FR 67202). The licensee affirmed the applicability of the NSHC 
determination in its application dated December 14, 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:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change involves the deletion of a reporting 
requirement. The change does not affect plant equipment or operating 
practices and therefore does not significantly increase the 
probability or consequences of an accident previously evaluated.
    2. Does the 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 that it deletes a 
reporting requirement. The change does not add new plant equipment, 
change existing plant equipment, or affect the operating practices 
of the facility. Therefore, the change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change deletes a reporting requirement. The change 
does not affect plant equipment or operating practices and therefore 
does not involve a significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Antonio Fernandez, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: December 15, 2006.
    Description of amendment request: The amendment request would 
revise the Technical Specifications (TSs) to adopt NRC-approved 
Revision 4 to Technical Specification Task Force (TSTF) Standard 
Technical Specification Change Traveler TSTF-372, ``Addition of LCO 
[Limiting Condition for Operation] 3.0.8, Inoperability of Snubbers.'' 
The amendment would add (1) a new LCO 3.0.8 addressing when one or more 
required snubbers are unable to perform their associated support 
function(s) (i.e., the snubber is inoperable) and (2) a reference to 
LCO 3.0.8 in LCO 3.0.1 on when LCOs shall be met.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on November 24, 2004 (69 FR 68412), on possible 
license amendments adopting TSTF-372 using the NRC's consolidated line 
item improvement process (CLIIP) for amending licensee's TSs, which 
included a model safety evaluation (SE) and model no significant 
hazards consideration (NSHC) determination. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on May 4, 2005 
(70 FR 23252), which included the resolution of public comments on the 
model SE. The May 4, 2005, notice of availability referenced the 
November 24, 2004, notice. The licensee has affirmed the applicability 
of the following NSHC determination in its application.
    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:


[[Page 155]]



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

    The proposed change allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an inoperable snubber if risk is assessed and managed. The 
postulated seismic event requiring snubbers is a low-probability 
occurrence and the overall TS system safety function would still be 
available for the vast majority of anticipated challenges. 
Therefore, the probability of an accident previously evaluated is 
not significantly increased, if at all. The consequences of an 
accident while relying on allowance provided by proposed LCO 3.0.8 
are no different than the consequences of an accident while relying 
on the TS required actions in effect without the allowance provided 
by proposed LCO 3.0.8. Therefore, the consequences of an accident 
previously evaluated are not significantly affected by this change. 
The addition of a requirement to assess and manage the risk 
introduced by this change will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

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

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Allowing delay times for entering [a] supported system TS when 
inoperability is due solely to inoperable snubbers, if risk is 
assessed and managed, will not introduce new failure modes or 
effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns. Thus, this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.

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

    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber, if risk is assessed and managed. The postulated seismic 
event requiring snubbers is a low-probability occurrence and the 
overall TS system safety function would still be available for the 
vast majority of anticipated challenges. The risk impact of the 
proposed TS changes was assessed following the three-tiered approach 
recommended in [NRC] RG [Regulatory Guide] 1.177. A bounding risk 
assessment was performed to justify the proposed TS changes. This 
application of LCO 3.0.8 is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk[, 
which is required by the proposed LCO 3.0.8]. The net change to the 
margin of safety is insignificant. Therefore, this change does not 
involve a significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: David Terao.

Notice of Issuance of