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, 41845-41848 [E6-11672]
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sroberts on PROD1PC70 with NOTICES
Federal Register / Vol. 71, No. 141 / Monday, July 24, 2006 / Notices
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, https://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
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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
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,
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
PO 00000
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41845
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 email to OGCMailCenter@nrc.gov. A copy
of the request for hearing and petition
for leave to intervene should also be
sent to the John O’Neill, 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 29, 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, https://
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 pdr@nrc.gov.
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. E6–11674 Filed 7–21–06; 8:45 am]
BILLING CODE 7590–01–P
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
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Federal Register / Vol. 71, No. 141 / Monday, July 24, 2006 / Notices
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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
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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 tubeto-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-tosecondary 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
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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
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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 60day 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
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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, https://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
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41847
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
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Federal Register / Vol. 71, No. 141 / Monday, July 24, 2006 / Notices
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,
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 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, https://
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 pdr@nrc.gov.
Dated at Rockville, Maryland, this 14th day
of July 2006.
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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
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 (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 (1)
delete the containment atmosphere
gaseous radioactivity monitor from
Technical Specification (TS) 3.4.15,
‘‘RCS [Reactor Coolant System] Leakage
Detection Instrumentation,’’ and (2)
revise existing conditions, required
actions, completion times, and
surveillance requirements in TS 3.4.15
to account for the monitor being
deleted. The licensee submitted this
amendment request in its application
dated June 26, 2006. This application
revised the licensee’s application dated
August 26, 2005, for which a notice of
consideration of issuance of an
amendment to facility operating license
and opportunity for a hearing was
published in the Federal Register on
October 25, 2005 (70 FR 61663).
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)
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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) The proposed change does not involve
a significant increase in the probability or
consequences of an accident previously
evaluated.
Response: No.
The proposed change has been evaluated
and determined to not increase the
probability or consequences of an accident
previously evaluated. The proposed change
does not make hardware changes and does
not alter the configuration of any plant
system, structure, or component (SSC). The
proposed change only removes the
containment atmosphere gaseous
radioactivity monitor as an option for
meeting the OPERABILITY requirements for
TS 3.4.15. The TS will continue to require
diverse means of leakage detection
equipment, thus ensuring that [RCS] leakage
due to cracks would continue to be identified
prior to propagating to the point of a pipe
break and the plant shutdown accordingly.
Therefore, the consequences of an accident
[previously evaluated] are not increased.
(2) The proposed change does not create
the possibility of a new or different kind of
accident from any accident previously
evaluated.
Response: No.
The proposed change does not involve the
use or installation of new equipment and the
currently installed equipment will not be
operated in a new or different manner. No
new or different system interactions are
created and no new processes are introduced.
The proposed changes will not introduce any
new failure mechanisms, malfunctions, or
accident initiators not already considered in
the design and licensing bases [for WCGS].
The proposed change does not affect any SSC
associated with an accident initiator. 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 change does not involve
a significant reduction in a margin of safety.
Response: No.
The proposed change does not alter any
Reactor Coolant System (RCS) leakage
detection components. The proposed change
only removes the containment atmosphere
gaseous radioactivity monitor as an option
for meeting the OPERABILITY requirements
for TS 3.4.15. This change is required since
the level of radioactivity in the WCGS reactor
coolant has become much lower than what
was assumed in the USAR [(Updated Safety
Analysis Report) when the plant was
licensed] and the gaseous channel [(monitor)]
can no longer promptly detect a small RCS
leak under normal [operating] conditions.
The proposed amendment continues to
require diverse means of [RCS] leakage
detection equipment with [the] capability to
promptly detect RCS leakage. Although not
E:\FR\FM\24JYN1.SGM
24JYN1
]]>Agencies
[Federal Register Volume 71, Number 141 (Monday, July 24, 2006)]
[Notices]
[Pages 41845-41848]
From the Federal Register Online via the Government Printing 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, https://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, 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 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, https://
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 pdr@nrc.gov.
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
