[Federal Register Volume 71, Number 141 (Monday, July 24, 2006)]
[Notices]
[Pages 41845-41848]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-11672]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-482]


Wolf Creek Nuclear Operating Corporation; 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 (NRC or the Commission) is 
considering issuance of an amendment to Facility Operating License No. 
NPF-42, issued to Wolf Creek Nuclear Operating Corporation (the 
licensee), for operation of the Wolf Creek Generating Station (WCGS), 
located in Coffey County, Kansas.
    The proposed amendment would revise Technical Specification 5.5.9, 
``Steam Generator (SG) Program,'' by

[[Page 41846]]

changing the ``Refueling Outage 14'' to ``Refueling Outage 15'' in two 
places. This change would extend the provisions for SG tube repair 
criteria and inspections that were approved for Refueling Outage 14, 
and the subsequent operating cycle, in Amendment No. 162 issued April 
28, 2005, to Refueling Outage 15, and the subsequent operating cycle. 
This was proposed in the licensee's application dated June 30, 2006.
    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), 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:

    (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 do[es] 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 changes 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 
tubesheet and from the differential pressure between the primary and 
secondary side. Based on this design, the structural margins against 
burst, discussed in [Nuclear Energy Institute] NEI 97-06, Revision 
2, and 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 changes 
result in no significant increase in the probability of the 
occurrence of a[n] SGTR accident.
    At normal operating pressures, leakage from primary water stress 
corrosion cracking (PWSCC) below the proposed limited 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 
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[n] SLB is unaffected by the potential 
failure of a steam generator tube as this failure is not an 
initiator for a[n] SLB.
    The consequences of a[n] SLB are also not significantly affected 
by the proposed change. During a[n] SLB accident, the reduction in 
pressure above the tubesheet on the shell side of the steam 
generator 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[n] 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 during postulated SLB accident conditions would 
be expected to be less than that during normal operation for 
indications near the bottom of the tubesheet (i.e., including 
indications in the tube-end welds). This conclusion is based on the 
observation that while the driving pressure causing leakage 
increases by approximately a factor of two, the flow resistance 
associated with an increase in the tube-to-tubesheet contact 
pressure, during a[n] SLB, increases by approximately a factor of 6. 
While such a leakage decrease is logically expected, the postulated 
accident leak rate could be conservatively bounded by twice the 
normal operating leak rate if the increase in contact pressure is 
ignored. Since normal operating leakage is limited to less than 
0.104 gpm (150 gpd) per TS 3.4.13, ``RCS Operational LEAKAGE,'' the 
associated accident condition leak rate, assuming all leakage to be 
from lower tubesheet indications, would be bounded by 0.208 gpm, 
twice the normal operational leakage. This value is well within the 
assumed accident leakage rate of 1.0 gpm discussed in WCGS Updated 
Safety Analysis Report, Table 15.1-3, ``Parameters Used in 
Evaluating the Radiological Consequences of a Main Steam Line 
Break.'' Hence it is reasonable to omit any consideration of 
inspection of the tube, tube-end weld, bulges/overexpansions or 
other anomalies below 17 inches from the top of the hot leg 
tubesheet. Therefore, the consequences of a[n] 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 
[any] 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. Therefore, the proposed changes do 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 changes maintain 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 
limited hot leg tubesheet inspection depth methodology for 
determining that steam generator tube integrity considerations 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 
consequences 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

[[Page 41847]]

circumferentially oriented cracking, Westinghouse letter LTR-CDME-
05-82-P, ``Limited Inspection of the Steam Generator Tube Portion 
Within the Tubesheet at Wolf Creek Generating Station,'' 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 hot leg 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 hot leg tubesheet inspection depth criteria.
    Therefore, the proposed changes do not involve a significant 
reduction in any margin to 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, 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 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 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

[[Page 41848]]

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 the Jay Silberg, Esq., 
Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street, NW., Washington, DC 
20037, attorney for the licensee.
    For further details with respect to this action, see the 
application for amendment dated June 30, 2006, which is available for 
public inspection at the Commission's PDR, located at One White Flint 
North, File Public 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 14th day of July 2006.

    For the Nuclear Regulatory Commission.
Jack Donohew,
Senior Project Manager, Plant Licensing Branch IV, Division of 
Operating Reactor Licensing, Office of Nuclear Reactor Regulation.
 [FR Doc. E6-11672 Filed 7-21-06; 8:45 am]
BILLING CODE 7590-01-P