Braidwood Station, Units 1 and 2; Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing, 13029-13032 [E8-4861]
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Federal Register / Vol. 73, No. 48 / Tuesday, March 11, 2008 / Notices
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Date of initial notice in Federal
Register: September 25, 2007 (72 FR
54479). The supplement dated January
24, 2008, provided additional
information that clarified the
application, did not expand the scope of
the application as originally noticed,
and did not change the staff’s original
proposed no significant hazards
consideration determination. The
Commission’s related evaluation of the
amendments is contained in a Safety
Evaluation dated February 20, 2008.
No significant hazards consideration
comments received: No.
PSEG Nuclear LLC, Docket Nos. 50–272
and 50–311, Salem Nuclear Generating
Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of application for amendments:
March 16, 2007, as supplemented on
August 30, September 14, and
November 20, 2007, and January 16,
2008.
Brief description of amendments: The
amendments revise the Updated Final
Safety Analysis Report (UFSAR) to
modify the Salem licensing basis with
respect to the response times associated
with a steam generator feedwater pump
(SGFP) trip and feedwater isolation
valve (FIV) closure. The amendments
also revise the Technical Specification
(TS) requirements for the containment
fan cooler unit (CFCU) cooling water
flow rate. These changes are associated
with a revised containment response
analysis that credits an SGFP trip and
FIV closure (on a feedwater regulator
valve failure) to reduce the mass/energy
release to the containment during a
main steam line break. The containment
analysis also credits a reduced heat
removal capability for the CFCUs,
allowing a reduction in the required
service water flow to the CFCUs.
Date of issuance: February 27, 2008.
Effective date: As of the date of
issuance, to be implemented prior to
restart from refueling outage 1R19 for
Salem Unit 1 and prior to restart from
refueling outage 2R16 for Salem Unit 2.
Amendment Nos.: 287 and 270.
Facility Operating License Nos. DPR–
70 and DPR–75: The amendments revise
the TSs, the license and the UFSAR.
Date of initial notice in Federal
Register: April 10, 2007 (72 FR 17951).
The letters dated August 30, September
14, and November 20, 2007, and January
16, 2008, provided clarifying
information that did not change the
initial proposed no significant hazards
consideration determination or expand
the application beyond the scope of the
original Federal Register notice.
The Commission’s related evaluation
of the amendments is contained in a
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Safety Evaluation dated February 27,
2008.
No significant hazards consideration
comments received: No.
Southern Nuclear Operating Company,
Inc., Georgia Power Company,
Oglethorpe Power Corporation,
Municipal Electric Authority of
Georgia, City of Dalton, Georgia, Docket
Nos. 50–321 and 50–366, Edwin I.
Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of application for amendments:
October 18, 2007.
Brief description of amendments: The
amendments revised the Technical
Specifications for unit staff
qualifications and also included a
revised position title for ‘‘Health
Physics Superintendent.’’
Date of issuance: February 21, 2008.
Effective date: As of the date of
issuance and shall be implemented
within 90 days from the date of
issuance.
Amendment Nos.: 255 and 199.
Renewed Facility Operating License
Nos. DPR–57 and NPF–5: Amendments
revised the licenses and the technical
specifications.
Date of initial notice in Federal
Register: November 20, 2007 (72 FR
65372).
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated February 21,
2008.
No significant hazards consideration
comments received: No.
Dated at Rockville, Maryland, this 29th day
of February 2008.
For the Nuclear Regulatory Commission.
John W. Lubinski,
Deputy Director, Division of Operating
Reactor Licensing, Office of Nuclear Reactor
Regulation.
[FR Doc. E8–4690 Filed 3–10–08; 8:45 am]
BILLING CODE 7590–01–P
NUCLEAR REGULATORY
COMMISSION
[Docket Nos. STN 50–456 and STN 50–457]
Braidwood Station, Units 1 and 2;
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, or the
NRC) is considering issuance of an
amendment to Facility Operating
License No. NPF–72 and Facility
Operating License No. NPF–77 to
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13029
Exelon Generation Company, LLC (the
licensee) for operation of the Braidwood
Station, Units 1 and 2 (Braidwood),
which is located in Will County,
Illinois.
The proposed amendment in the
licensee’s application dated February
25, 2008, would revise Technical
Specification (TS) 5.5.9, ‘‘Steam
Generator (SG) Program,’’ and TS 5.6.9,
‘‘Steam Generator Tube Inspection
Report.’’ For TS 5.5.9, the amendment
would replace the existing alternate
repair criteria (ARC) in the provisions
for SG tube repair criteria during
Braidwood, Unit 2, refueling outage 13
and the subsequent operating cycle. For
TS 5.6.9, three new reporting
requirements are proposed to be added
to the existing seven requirements. The
proposed changes would only affect
Braidwood, Unit 2; however this is
docketed for Braidwood, Units 1 and 2,
since the TS are common to Units 1 and
2.
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.
Of the various accidents previously
evaluated, the proposed changes only affect
the steam generator tube rupture (SGTR),
postulated steam line break (SLB), locked
rotor and control rod ejection accident
evaluations. Loss-of-coolant accident (LOCA)
conditions cause a compressive axial load to
act on the tube. Therefore, since the LOCA
tends to force the tube into the tubesheet
rather than pull it out, it is not a factor in
this amendment request. Another faulted
load consideration is a safe shutdown
earthquake (SSE); however, the seismic
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analysis of Model D5 steam generators has
shown that axial loading of the tubes is
negligible during an SSE.
At normal operating pressures, leakage
from primary water stress corrosion cracking
(PWSCC) below 17 inches from the top of the
tubesheet is limited by both the tube-totubesheet crevice and the limited crack
opening permitted by the tubesheet
constraint. Consequently, negligible normal
operating leakage is expected from cracks
within the tubesheet region.
For the SGTR event, the required structural
margins of the steam generator tubes is
maintained by limiting the allowable
ligament size for a circumferential crack to
remain in service to 214 degrees below 17
inches from the top of the tubesheet. Tube
rupture is precluded for cracks in the
hydraulic expansion region due to the
constraint provided by the tubesheet. The
potential for tube pullout is mitigated by
limiting the allowable crack size to 214
degrees, which takes into account eddy
current uncertainty and crack growth rate. It
has been shown that a circumferential crack
with an azimuthal extent of 214 degrees
meets the performance criteria of NEI
[Nuclear Energy Institute] 97–06, Rev. 2,
‘‘Steam Generator Program Guidelines’’ and
the Draft Regulatory Guide (RG) 1.121,
‘‘Bases for Plugging Degraded PWR Steam
Generator Tubes.’’ Likewise, a visual
inspection will be conducted as necessary to
confirm that a circumferential crack of
greater than 294 degrees does not remain in
service in the tube-to-tubesheet weld metal in
any tube thereby mitigating the potential for
tube pullout. Therefore, the margin against
tube burst/pullout is maintained during
normal and postulated accident conditions
and the proposed change does not result in
a significant increase in the probability or
consequence of a SGTR.
The probability of a SLB is unaffected by
the potential failure of a SG tube as the
failure of a tube is not an initiator for a SLB
event. SLB leakage is limited by leakage flow
restrictions resulting from the leakage path
above potential cracks through the tube-totubesheet crevice. The leak rate during
postulated accident conditions has been
shown to remain within the accident analysis
assumptions for all axial or circumferentially
oriented cracks occurring 17 inches below
the top of the tubesheet. Since normal
operating leakage is limited to 0.10 gallons
per minute (gpm) (or 150 gallons per day
(gpd)), the attendant accident condition leak
rate, assuming all leakage to be from
indications below 17 inches from the top of
the tubesheet would be bounded by 0.5 gpm.
This value is within the accident analysis
assumptions for the limiting design basis
accident for Braidwood 2, which is the
postulated SLB event.
Based on the above, the performance
criteria of NEI–97–06, Rev. 2 and RG 1.121
continue to be met and 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?
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Response: No.
The proposed change does not introduce
any changes or mechanisms that create the
possibility of a new or different kind of
accident. Tube bundle integrity is expected
to be maintained for all plant conditions
upon implementation of the interim alternate
repair criteria. The proposed change does not
introduce any new equipment or any change
to existing equipment. No new effects on
existing equipment are created nor are any
new malfunctions introduced.
Therefore, based on the above evaluation,
the proposed 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 change maintains the
required structural margins of the steam
generator tubes for both normal and accident
conditions. NEI 97–06, Rev. 2 and RG 1.121
are used as the basis in the development of
the interim alternate repair criteria (IARC)
methodology for determining that steam
generator tube integrity considerations are
maintained within acceptable limits. RG
1.121 describes a method acceptable to the
NRC staff for meeting General Design Criteria
14, 15, 31, and 32 by reducing the probability
and consequences of an SGTR. RG 1.121
concludes that by determining the limiting
safe conditions of tube wall degradation
beyond which tubes with unacceptable
cracking, as established by inservice
inspection, should be removed from service
or repaired, the probability and consequences
of a SGTR are reduced. This RG uses safety
factors on loads for tube burst that are
consistent with the requirements of Section
III of the 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 in a tube
or the tube-to-tubeshe[e]t weld, the
Westinghouse analysis, provided in report
‘‘LTR–CDME–08–11 P-Attachment,’’ defines
a length of remaining tube ligament that
provides the necessary resistance to tube
pullout due to the pressure induced forces
(with applicable safety factors applied).
Additionally, it is shown that application of
the IARC will not result in unacceptable
primary-to-secondary leakage during all plant
conditions, including transients and
postulated accident conditions.
Based on the above, it is concluded that the
proposed changes do not result in any
reduction in a margin of safety.
Based on the above, EGC [the licensee]
concludes that the proposed amendment
presents no significant hazards consideration
under the standards set forth in 10 CFR
50.92(c) and, accordingly, a finding of ‘‘no
significant hazards consideration’’ is
justified.
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
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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, 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 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
O1F21, 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 person(s)
may file a request for a hearing with
respect to issuance of the amendment to
the subject facility operating license and
any person(s) whose interest may be
affected by this proceeding and who
wishes to participate as a party in the
proceeding must file a written request
via electronic submission through the
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NRC E-filing system 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 person(s) 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
requestor’s/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
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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.
A request for hearing or a petition for
leave to intervene must be filed in
accordance with the NRC E-Filing rule,
which the NRC promulgated on August
28, 2007 (72 FR 49139). The E-Filing
process requires participants to submit
and serve documents over the Internet
or in some cases to mail copies on
electronic storage media. Participants
may not submit paper copies of their
filings unless they seek a waiver in
accordance with the procedures
described below.
To comply with the procedural
requirements of E-Filing, at least five (5)
days prior to the filing deadline, the
petitioner/requestor must contact the
Office of the Secretary by e-mail at
HEARINGDOCKET@NRC.GOV, or by
calling (301) 415–1677, to request (1) a
digital ID certificate, which allows the
participant (or its counsel or
representative) to digitally sign
documents and access the E-Submittal
server for any proceeding in which it is
participating; and/or (2) creation of an
electronic docket for the proceeding
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13031
(even in instances in which the
petitioner/requestor (or its counsel or
representative) already holds an NRCissued digital ID certificate). Each
petitioner/requestor will need to
download the Workplace Forms
Viewer TM to access the Electronic
Information Exchange (EIE), a
component of the E-Filing system. The
Workplace Forms Viewer TM is free and
is available at https://www.nrc.gov/sitehelp/e-submittals/install-viewer.html.
Information about applying for a digital
ID certificate is available on NRC’s
public Web site at https://www.nrc.gov/
site-help/e-submittals/applycertificates.html.
Once a petitioner/requestor has
obtained a digital ID certificate, had a
docket created, and downloaded the EIE
viewer, it can then submit a request for
hearing or petition for leave to
intervene. Submissions should be in
Portable Document Format (PDF) in
accordance with NRC guidance
available on the NRC public Web site at
https://www.nrc.gov/site-help/esubmittals.html. A filing is considered
complete at the time the filer submits its
documents through EIE. To be timely,
an electronic filing must be submitted to
the EIE system no later than 11:59 p.m.
Eastern Time on the due date. Upon
receipt of a transmission, the E-Filing
system time-stamps the document and
sends the submitter an e-mail notice
confirming receipt of the document. The
EIE system also distributes an e-mail
notice that provides access to the
document to the NRC Office of the
General Counsel and any others who
have advised the Office of the Secretary
that they wish to participate in the
proceeding, so that the filer need not
serve the documents on those
participants separately. Therefore,
applicants and other participants (or
their counsel or representative) must
apply for and receive a digital ID
certificate before a hearing request/
petition to intervene is filed so that they
can obtain access to the document via
the E-Filing system.
A person filing electronically may
seek assistance through the ‘‘Contact
Us’’ link located on the NRC Web site
at https://www.nrc.gov/site-help/esubmittals.html or by calling the NRC
technical help line, which is available
between 8:30 a.m. and 4:15 p.m.,
Eastern Time, Monday through Friday.
The help line number is (800) 397–4209
or locally, (301) 415–4737. Participants
who believe that they have a good cause
for not submitting documents
electronically must file a motion, in
accordance with 10 CFR 2.302(g), with
their initial paper filing requesting
authorization to continue to submit
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documents in paper format. Such filings
must be submitted 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; or
(2) courier, express mail, or expedited
delivery service to the Office of the
Secretary, Sixteenth Floor, One White
Flint North, 11555 Rockville Pike,
Rockville, Maryland 20852, Attention:
Rulemaking and Adjudications Staff.
Participants filing a document in this
manner are responsible for serving the
document on all other participants.
Filing is considered complete by firstclass mail as of the time of deposit in
the mail, or by courier, express mail, or
expedited delivery service upon
depositing the document with the
provider of the service.
Non-timely requests and/or petitions
and contentions will not be entertained
absent a determination by the
Commission, the presiding officer, or
the Atomic Safety and Licensing Board
that the petition and/or request should
be granted and/or the contentions
should be admitted, based on a
balancing of the factors specified in 10
CFR 2.309(c)(1)(i)–(viii). To be timely,
filings must be submitted no later than
11:59 p.m. Eastern Time on the due
date.
Documents submitted in adjudicatory
proceedings will appear in NRC’s
electronic hearing docket which is
available to the public at https://
ehd.nrc.gov/EHD_Proceeding/home.asp,
unless excluded pursuant to an order of
the Commission, an Atomic Safety and
Licensing Board, or a Presiding Officer.
Participants are requested not to include
personal privacy information, such as
social security numbers, home
addresses, or home phone numbers in
their filings. With respect to copyrighted
works, except for limited excerpts that
serve the purpose of the adjudicatory
filings and would constitute a Fair Use
application, Participants are requested
not to include copyrighted materials in
their submissions.
For further details with respect to this
license amendment application, see the
letter dated February 25, 2008, from the
Exelon Generation Company, LLC,
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
electronically 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/
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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 4th day
of March 2008.
For the Nuclear Regulatory Commission.
Meghan M. Thorpe-Kavanaugh,
Project Manager, Plant Licensing Branch III–
2, Division of Operating Reactor Licensing,
Office of Nuclear Reactor Regulation.
[FR Doc. E8–4861 Filed 3–10–08; 8:45 am]
BILLING CODE 7590–01–P
NUCLEAR REGULATORY
COMMISSION
[Docket No. 50–354]
PSEG Nuclear, LLC; Hope Creek
Generating Station Final
Environmental Assessment and
Finding of No Significant Impact;
Related to the Proposed License
Amendment To Increase the Maximum
Reactor Power Level
U.S. Nuclear Regulatory
Commission (NRC).
SUMMARY: As required by Title 10 of the
Code of Federal Regulations (10 CFR)
Part 51, the NRC has prepared a final
Environmental Assessment (EA) as its
evaluation of a request by the PSEG
Nuclear, LLC (PSEG) for a license
amendment to increase the maximum
thermal power at Hope Creek
Generating Station (HCGS) from 3,339
megawatts-thermal (MWt) to 3,840
MWt. The EA assesses environmental
impacts up to a maximum thermal
power level of 3,952 MWt, as the
applicant’s environmental report was
based on that power level. The NRC
staff did not identify any significant
impact from the information provided
in the licensee’s EPU application for
HCGS or from the NRC staff’s
independent review. The final EA and
Finding of No Significant Impact are
being published in the Federal Register.
The NRC published a draft EA and
finding of no significant impact on the
proposed action for public comment in
the Federal Register on October 22,
2007 (72 FR 59563). Two sets of
comments were received on the draft
EA: (1) From PSEG Nuclear, LLC by
letter dated November 21, 2007
(Agencywide Documents Access and
Management System (ADAMS)
Accession No. ML073600851); and (2)
from the State of New Jersey Department
of Environmental Protection (NJDEP) by
letter dated November 21, 2007
AGENCY:
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(ADAMS Accession No. ML073600859).
These comments are addressed below.
Disposition of Public Comments on the
Draft Environmental Assessment E
PSEG Comment Number 1: Modify
the Cooling Tower Impacts section to
more clearly reflect that NJDEP has
issued the Title V Air Operating Permit
authorizing emissions at 42 lbs/hr upon
approval of the [United States
Environmental Protection Agency]
USEPA.
NRC Response Number 1: This
comment is a clarification and editorial
correction to the draft Environmental
Assessment. Based on this comment, the
NRC staff revised the appropriate
section of the final EA.
PSEG Comment Number 2: Modify
the Discharge Impacts section to reflect
that the [total dissolved solids] TDS
limits are indirectly in the Title V Air
Operating Permit and not in the [New
Jersey Pollutant Discharge Elimination
System] NJPDES Permit.
NRC Response Number 2: This
comment is a clarification and editorial
correction to the draft Environmental
Assessment. Based on this comment, the
NRC staff revised the appropriate
section of the final EA.
PSEG Comment Number 3: Modify
the Discharge Impacts section to reflect
that total suspended solids and [total
organic carbon] TOC are not routinely
monitored and acute and chronic
biological toxicity tests are performed
during each NJPDES Permit renewal.
NRC Response Number 3: This
comment is a clarification and editorial
correction to the draft Environmental
Assessment. Based on this comment, the
NRC staff revised the appropriate
section of the final EA.
PSEG Comment Number 4: Modify
the Impacts on Aquatic Biota section,
Table 1, to reflect that Atlantic Croaker
are considered to be a single Atlantic
coast stock.
NRC Response Number 4: Upon
further review, the NRC agrees with the
comment. Based on this comment, the
NRC staff revised the appropriate
section of the final EA.
PSEG Comment Number 5: Modify
the Impacts on Aquatic Biota section to
identify inland silversides instead of
tidewater silversides.
NRC Response Number 5: Upon
further review, the NRC agrees with the
comment. Based on this comment, the
NRC staff revised the appropriate
section of the final EA.
PSEG Comment Number 6: Modify
the Impacts on Aquatic Biota section to
reflect the extensive biological
monitoring program at the adjacent
Salem Generating Station, reflect the
E:\FR\FM\11MRN1.SGM
11MRN1
]]>Agencies
[Federal Register Volume 73, Number 48 (Tuesday, March 11, 2008)]
[Notices]
[Pages 13029-13032]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E8-4861]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
[Docket Nos. STN 50-456 and STN 50-457]
Braidwood Station, Units 1 and 2; 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, or the NRC)
is considering issuance of an amendment to Facility Operating License
No. NPF-72 and Facility Operating License No. NPF-77 to Exelon
Generation Company, LLC (the licensee) for operation of the Braidwood
Station, Units 1 and 2 (Braidwood), which is located in Will County,
Illinois.
The proposed amendment in the licensee's application dated February
25, 2008, would revise Technical Specification (TS) 5.5.9, ``Steam
Generator (SG) Program,'' and TS 5.6.9, ``Steam Generator Tube
Inspection Report.'' For TS 5.5.9, the amendment would replace the
existing alternate repair criteria (ARC) in the provisions for SG tube
repair criteria during Braidwood, Unit 2, refueling outage 13 and the
subsequent operating cycle. For TS 5.6.9, three new reporting
requirements are proposed to be added to the existing seven
requirements. The proposed changes would only affect Braidwood, Unit 2;
however this is docketed for Braidwood, Units 1 and 2, since the TS are
common to Units 1 and 2.
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.
Of the various accidents previously evaluated, the proposed
changes only affect the steam generator tube rupture (SGTR),
postulated steam line break (SLB), locked rotor and control rod
ejection accident evaluations. Loss-of-coolant accident (LOCA)
conditions cause a compressive axial load to act on the tube.
Therefore, since the LOCA tends to force the tube into the tubesheet
rather than pull it out, it is not a factor in this amendment
request. Another faulted load consideration is a safe shutdown
earthquake (SSE); however, the seismic
[[Page 13030]]
analysis of Model D5 steam generators has shown that axial loading
of the tubes is negligible during an SSE.
At normal operating pressures, leakage from primary water stress
corrosion cracking (PWSCC) below 17 inches from the top of the
tubesheet 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.
For the SGTR event, the required structural margins of the steam
generator tubes is maintained by limiting the allowable ligament
size for a circumferential crack to remain in service to 214 degrees
below 17 inches from the top of the tubesheet. Tube rupture is
precluded for cracks in the hydraulic expansion region due to the
constraint provided by the tubesheet. The potential for tube pullout
is mitigated by limiting the allowable crack size to 214 degrees,
which takes into account eddy current uncertainty and crack growth
rate. It has been shown that a circumferential crack with an
azimuthal extent of 214 degrees meets the performance criteria of
NEI [Nuclear Energy Institute] 97-06, Rev. 2, ``Steam Generator
Program Guidelines'' and the Draft Regulatory Guide (RG) 1.121,
``Bases for Plugging Degraded PWR Steam Generator Tubes.'' Likewise,
a visual inspection will be conducted as necessary to confirm that a
circumferential crack of greater than 294 degrees does not remain in
service in the tube-to-tubesheet weld metal in any tube thereby
mitigating the potential for tube pullout. Therefore, the margin
against tube burst/pullout is maintained during normal and
postulated accident conditions and the proposed change does not
result in a significant increase in the probability or consequence
of a SGTR.
The probability of a SLB is unaffected by the potential failure
of a SG tube as the failure of a tube is not an initiator for a SLB
event. SLB leakage is limited by leakage flow restrictions resulting
from the leakage path above potential cracks through the tube-to-
tubesheet crevice. The leak rate during postulated accident
conditions has been shown to remain within the accident analysis
assumptions for all axial or circumferentially oriented cracks
occurring 17 inches below the top of the tubesheet. Since normal
operating leakage is limited to 0.10 gallons per minute (gpm) (or
150 gallons per day (gpd)), the attendant accident condition leak
rate, assuming all leakage to be from indications below 17 inches
from the top of the tubesheet would be bounded by 0.5 gpm. This
value is within the accident analysis assumptions for the limiting
design basis accident for Braidwood 2, which is the postulated SLB
event.
Based on the above, the performance criteria of NEI-97-06, Rev.
2 and RG 1.121 continue to be met and 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 introduce any changes or mechanisms
that create the possibility of a new or different kind of accident.
Tube bundle integrity is expected to be maintained for all plant
conditions upon implementation of the interim alternate repair
criteria. The proposed change does not introduce any new equipment
or any change to existing equipment. No new effects on existing
equipment are created nor are any new malfunctions introduced.
Therefore, based on the above evaluation, the proposed 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 change maintains the required structural margins of
the steam generator tubes for both normal and accident conditions.
NEI 97-06, Rev. 2 and RG 1.121 are used as the basis in the
development of the interim alternate repair criteria (IARC)
methodology for determining that steam generator tube integrity
considerations are maintained within acceptable limits. RG 1.121
describes a method acceptable to the NRC staff for meeting General
Design Criteria 14, 15, 31, and 32 by reducing the probability and
consequences of an SGTR. RG 1.121 concludes that by determining the
limiting safe conditions of tube wall degradation beyond which tubes
with unacceptable cracking, as established by inservice inspection,
should be removed from service or repaired, the probability and
consequences of a SGTR are reduced. This RG uses safety factors on
loads for tube burst that are consistent with the requirements of
Section III of the 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 in a tube or the tube-to-
tubeshe[e]t weld, the Westinghouse analysis, provided in report
``LTR-CDME-08-11 P-Attachment,'' defines a length of remaining tube
ligament that provides the necessary resistance to tube pullout due
to the pressure induced forces (with applicable safety factors
applied). Additionally, it is shown that application of the IARC
will not result in unacceptable primary-to-secondary leakage during
all plant conditions, including transients and postulated accident
conditions.
Based on the above, it is concluded that the proposed changes do
not result in any reduction in a margin of safety.
Based on the above, EGC [the licensee] concludes that the
proposed amendment presents no significant hazards consideration
under the standards set forth in 10 CFR 50.92(c) and, accordingly, a
finding of ``no significant hazards consideration'' is justified.
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 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
O1F21, 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
person(s) may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any
person(s) whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request via electronic submission through the
[[Page 13031]]
NRC E-filing system 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
person(s) 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 requestor's/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.
A request for hearing or a petition for leave to intervene must be
filed in accordance with the NRC E-Filing rule, which the NRC
promulgated on August 28, 2007 (72 FR 49139). The E-Filing process
requires participants to submit and serve documents over the Internet
or in some cases to mail copies on electronic storage media.
Participants may not submit paper copies of their filings unless they
seek a waiver in accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
five (5) days prior to the filing deadline, the petitioner/requestor
must contact the Office of the Secretary by e-mail at
HEARINGDOCKET@NRC.GOV, or by calling (301) 415-1677, to request (1) a
digital ID certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and/or (2)
creation of an electronic docket for the proceeding (even in instances
in which the petitioner/requestor (or its counsel or representative)
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms Viewer
TM to access the Electronic Information Exchange (EIE), a
component of the E-Filing system. The Workplace Forms Viewer
TM is free and is available at https://www.nrc.gov/site-help/
e-submittals/install-viewer.html. Information about applying for a
digital ID certificate is available on NRC's public Web site at https://
www.nrc.gov/site-help/e-submittals/apply-certificates.html.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it can then submit
a request for hearing or petition for leave to intervene. Submissions
should be in Portable Document Format (PDF) in accordance with NRC
guidance available on the NRC public Web site at https://www.nrc.gov/
site-help/e-submittals.html. A filing is considered complete at the
time the filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE system no later than
11:59 p.m. Eastern Time on the due date. Upon receipt of a
transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
EIE system also distributes an e-mail notice that provides access to
the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically may seek assistance through the
``Contact Us'' link located on the NRC Web site at https://www.nrc.gov/
site-help/e-submittals.html or by calling the NRC technical help line,
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time,
Monday through Friday. The help line number is (800) 397-4209 or
locally, (301) 415-4737. Participants who believe that they have a good
cause for not submitting documents electronically must file a motion,
in accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit
[[Page 13032]]
documents in paper format. Such filings must be submitted 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; or (2) courier, express
mail, or expedited delivery service to the Office of the Secretary,
Sixteenth Floor, One White Flint North, 11555 Rockville Pike,
Rockville, Maryland 20852, Attention: Rulemaking and Adjudications
Staff. Participants filing a document in this manner are responsible
for serving the document on all other participants. Filing is
considered complete by first-class mail as of the time of deposit in
the mail, or by courier, express mail, or expedited delivery service
upon depositing the document with the provider of the service.
Non-timely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission, the presiding
officer, or the Atomic Safety and Licensing Board that the petition
and/or request should be granted and/or the contentions should be
admitted, based on a balancing of the factors specified in 10 CFR
2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later
than 11:59 p.m. Eastern Time on the due date.
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
https://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant
to an order of the Commission, an Atomic Safety and Licensing Board, or
a Presiding Officer. Participants are requested not to include personal
privacy information, such as social security numbers, home addresses,
or home phone numbers in their filings. With respect to copyrighted
works, except for limited excerpts that serve the purpose of the
adjudicatory filings and would constitute a Fair Use application,
Participants are requested not to include copyrighted materials in
their submissions.
For further details with respect to this license amendment
application, see the letter dated February 25, 2008, from the Exelon
Generation Company, LLC, 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 electronically 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 4th day of March 2008.
For the Nuclear Regulatory Commission.
Meghan M. Thorpe-Kavanaugh,
Project Manager, Plant Licensing Branch III-2, Division of Operating
Reactor Licensing, Office of Nuclear Reactor Regulation.
[FR Doc. E8-4861 Filed 3-10-08; 8:45 am]
BILLING CODE 7590-01-P
