[Federal Register Volume 72, Number 155 (Monday, August 13, 2007)]
[Notices]
[Pages 45272-45274]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-15766]


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NUCLEAR REGULATORY COMMISSION


[Docket No. 50-414]

Duke Power Company, LLC.; Notice of Consideration of Issuance of 
Amendment to Facility Operating License, Proposed No Significant 
Hazards Consideration Determination, and Opportunity for a Hearing
    The U.S. Nuclear Regulatory Commission (the Commission) is 
considering issuance of an amendment to Facility Operating License No. 
NPF-52 issued to Duke Power Company, LLC. (the licensee) for operation 
of the Catawba Nuclear Station, Unit 2 located in York County, South 
Carolina.
    The proposed amendment would revise the Catawba Nuclear Station, 
Unit 2, Technical Specification Section 5.5.9 concerning modifications 
to the steam generator tube repair criteria. Before issuance of the 
proposed license amendment, the Commission will have made findings 
required by the Atomic Energy Act of 1954, as amended (the Act), and 
the Commission's regulations.
    The Commission has made a proposed determination that the amendment 
request involves no significant hazards consideration. Under the 
Commission's regulations in Title 10 of the Code of Federal Regulations 
(10 CFR), Part 50, Section 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. 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:

First Standard

    A. Does operation of the facility in accordance with the 
proposed amendment 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 SG [steam generator] tube repair 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 consideration to the proposed change to the SG tube 
repair criteria, are the SG tube rupture event and the steam line 
break event.
    During the SG tube rupture event, the required structural 
integrity margins of the SG tubes will be maintained by the presence 
of the SG tubesheet. SG tubes are hydraulically expanded in the 
tubesheet area. Tube rupture in tubes with cracks in the tubesheet 
region of the tube is precluded by the constraint provided by the 
tubesheet. This constraint results from the hydraulic expansion 
process, thermal expansion mismatch between the tube and tubesheet, 
and the differential pressure between the primary and secondary 
side. Based on this design, the structural margins against burst, 
discussed in the TS 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 changes 
result in no significant increase in the probability of the 
occurrence of a SG tube rupture event.
    At normal operating pressures, leakage from stress corrosion 
cracking below the proposed limited tube repair 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 a SG tube rupture event are 
affected by the primary-to-secondary leakage flow during the event. 
Primary-to-secondary leakage flow through a postulated broken 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 a steam line break event is unaffected by the 
potential failure of a SG tube, as this failure is not an initiator 
for a steam line break event.
    The consequences of a steam line break event are also not 
significantly affected by the proposed change. During a steam line 
break event, the reduction in pressure above the tubesheet on the 
shell side of the SG creates an axially uniformly distributed load 
on the tubesheet due to the reactor coolant system pressure on the 
underside of the tubesheet. The resulting bending action constrains 
the tubes in the tubesheet, thereby restricting primary-to-secondary 
leakage below the midplane.
    Primary-to-secondary leakage from tube degradation in the 
tubesheet area during the limiting accident (i.e., a steam line 
break event) 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 steam line 
break event conditions will be no more than twice that allowed 
during normal operating conditions when the pressure boundary is 
relocated to the 17-inch depth. Since normal operating leakage is 
limited to 75 gallons per day through any one SG per the proposed 
license condition, the associated accident condition leak rate, 
assuming all leakage to be from lower tubesheet indications, would 
be limited to 150 gallons per day per SG. This is the value that is 
assumed in the steam line break dose analysis.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Second Standard

    B. Does operation of the facility in accordance with the 
proposed amendment create the possibility of a new or different kind 
of 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 be performed as previously assumed in 
accident analyses. Therefore, the proposed change does not create 
the possibility of a new or different kind of accident from any 
previously evaluated.

Third Standard

    C. Does operation of the facility in accordance with the 
proposed amendment involve a significant reduction in the margin of 
safety?
    Response: No.
    The proposed change maintains the required structural margins of 
the SG tubes for both normal and accident conditions. NEI [Nuclear 
Energy Institute] 97-06 and the Catawba TS are used as the bases in 
the development of the limited tubesheet tube repair depth 
methodology for determining that SG tube integrity considerations 
are maintained within acceptable limits. Regulatory Guide 1.121 
describes a method acceptable to the NRC for meeting General Design 
Criterion (GDC) 14, ``Reactor coolant pressure boundary,'' GDC 15, 
``Reactor

[[Page 45273]]

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 SG tube rupture event. By determining the limiting 
safe conditions for tube wall degradation, the probability and 
consequences of a SG tube rupture event are reduced. Safety factors 
are used for 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, the analysis referenced in 
support of this proposed amendment 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 tube repair 
depth criterion (17 inches) will preclude unacceptable primary-to-
secondary leakage during all plant conditions.
    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.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 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 Register notice. Written comments may also be 
delivered to Room 6D59, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Documents may be examined, and/or copied for a fee, at the NRC's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1 F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
    The filing of requests for 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 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed by the above date, 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 requestors/petitioner's interest. The petition must 
also identify 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 
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 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 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, 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

[[Page 45274]]

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.
    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(c)(1)(i)-(viii).
    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, [email protected]; 
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 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to Ms. Lisa F. Vaughn, 
Associate General Counsel and Managing Attorney, Duke Energy Carolinas, 
LLC, 526 South Church Street, EC07H, Charlotte, North Carolina 28202, 
attorney for the licensee.
    For further details with respect to this action, see the 
application for amendment dated April 30, 2007 (ADAMS Accession No. 
ML071280284), which is available for public inspection at the 
Commission's PDR, located at One White Flint North, Public File Area O1 
F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly 
available records will be accessible from the Agencywide Documents 
Access and Management System's (ADAMS) Public Electronic Reading Room 
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to ADAMS or who encounter 
problems in accessing the documents located in ADAMS, should contact 
the NRC PDR Reference staff by telephone at 1-800-397-4209, 301-415-
4737, or by e-mail to [email protected].

    Dated at Rockville, Maryland, this 6th day of August 2007.

    For the Nuclear Regulatory Commission.
John F. Stang,
Senior Project Manager, Plant Licensing Branch II-1, Division of 
Operating Reactor Licensing, Office of Nuclear Reactor Regulation.
 [FR Doc. E7-15766 Filed 8-10-07; 8:45 am]
BILLING CODE 7590-01-P