[Federal Register Volume 72, Number 249 (Monday, December 31, 2007)]
[Notices]
[Pages 74354-74365]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-25416]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from December 6, 2007 to December 19, 2007. The
last biweekly notice was published on December 18, 2007 (72 FR 71703).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed no Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received
[[Page 74355]]
within 30 days after the date of publication of this notice will be
considered in making any final determination. Within 60 days after the
date of publication of this notice, the licensee may file a request for
a hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, Division of Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the Commission's
Public Document Room (PDR), located at One White Flint North, Public
File Area O1F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice,
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
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 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 01F21, 11555 Rockville Pike
(first floor), Rockville, Maryland. Publicly available records will be
accessible from the Agencywide Documents Access and Management System's
(ADAMS) Public Electronic Reading Room on the Internet at the NRC Web
site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request
for a hearing or petition for leave to intervene is filed within 60
days, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for hearing or a petition for leave to intervene must be
filed in accordance with the NRC E-Filing rule, which the NRC
promulgated in 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
[email protected], or by
[[Page 74356]]
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 http://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 http://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 http://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 submitted 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 http://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 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
http://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 submission.
For further details with respect to this amendment action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina.
Date of amendment request: November 19, 2007.
Description of amendment request: The proposed amendment would make
administrative revisions to delete requirements that are obsolete or
redundant, or correct and clarify the typing and formatting of other
requirements. The proposed changes will not result in changes to the
plant design or the procedural controls for the operation,
surveillance, or maintenance of the plant.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the Proposed Changes Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated?
No. The proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed changes are administrative. The changes
delete obsolete or redundant requirements, clarify existing
requirements, and correct typing and formatting errors. There will
be no resulting changes to the plant design or procedural controls.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the Proposed Changes Create the Possibility of a New or
Different Kind of Accident From Any Previously Evaluated?
No. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated. There
are no physical changes being made to the plant or to the manner in
which the plant is operated. Therefore, the changes do not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Do the Proposed Changes Involve a Significant Reduction in
the Margin of Safety?
No. The proposed changes do not involve a significant reduction
in the margin of safety. There are no physical changes being made to
the plant or to the manner in which
[[Page 74357]]
the plant is operated. The proposed changes are administrative. The
changes delete obsolete or redundant requirements, clarify existing
requirements, and correct typing and formatting errors. Therefore,
the changes do not involve a significant reduction in any margin of
safety for HBRSEP [H.B. Robinson Steam Electric Plant], Unit No. 2.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal. Department, Progress Energy Service Company, LLC, Post
Office Box 1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Thomas H. Boyce.
Duke Power Company LLC, Docket No. 50-369, McGuire Nuclear Station,
Unit 1, Mecklenburg County, North Carolina.
Date of amendment request: February 21, 2007, as supplemented
August 9, 2007.
Description of amendment request: The proposed amendment would
allow, on a one-time basis, an extension of the interval governing the
conduct of the Integrated Leak Rate Test (ILRT) for McGuire Nuclear
Station, Unit 1. The proposed amendment would revise administrative
Technical Specification (TS) 5.5.2, ``Containment Leak Rate Testing
Program,'' from the currently approved 15-year interval (since the last
McGuire Nuclear Station, Unit 1, Type A test) to a frequency
encompassing the end of the McGuire Nuclear Station, Unit 1, End-of-
Cycle 19 refueling outage (approximately 6 months beyond the present TS
frequency).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. 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.
First Standard
The proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed extension to the Type A testing intervals
cannot increase the probability of an accident previously evaluated
since extension of the intervals is not a physical plant
modification that could alter the probability of accident
occurrence, nor is it an activity or modification by itself that
could lead to equipment failure or accident initiation. The proposed
extension to the Type A testing intervals does not result in a
significant increase in the consequences of an accident as
documented in NUREG-1493 [``Performance-Based Containment Leak-Test
Program'', NUREG-1493, September 1995]. The NUREG notes that very
few potential containment leakage paths are not identified by Type B
and Type C tests. It concludes that reducing the Type A testing
frequency to once per twenty years leads to an imperceptible
increase in risk. McGuire [Nuclear Station, Unit 1 (McGuire Unit 1)]
provides a high degree of assurance through testing and inspection
that the containment will not degrade in a manner detectable only by
Type A testing. Prior Type A tests for McGuire Unit 1 identified
containment leakage within acceptance criteria, indicating a very
leak tight containment. Inspections required by the ASME Code
[American Society of Mechanical Engineers (ASME), Boiler and
Pressure Vessel Code (Code)] are also performed in order to identify
indications of containment degradation that could affect leak
tightness. Separately, Type B and Type C testing, required by TS
[Technical Specification] identify any containment opening from
design penetrations, such as valves, that would otherwise be
detected by a Type A test. These factors establish that an extension
to the Type A test intervals will not represent a significant
increase in the consequences of an accident.
Second Standard
The proposed amendments will not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed revisions to the McGuire TS add a one-time
extension to the current interval for Type A testing. The current
test interval of fifteen years, based on past performance, would be
extended on a one-time basis to approximately fifteen and a half
years from the last Type A test. The proposed extension to the Type
A test interval does not create the possibility of a new or
different type of accident since there are no physical changes being
made to the plants and there are no changes to the operation of the
plants that could introduce a new failure mode.
Third Standard
The proposed amendment will not involve a significant reduction
in a margin of safety. The proposed revisions to the McGuire TS add
a one-time extension to the current interval for Type A testing. The
current test interval of fifteen years, based on past performance,
would be extended on a one-time basis to approximately fifteen and a
half years from the last Type A test. The proposed extension to Type
A test intervals will not significantly reduce the margin of safety.
The NUREG-1493 generic study of the effects of extending containment
leakage testing intervals found that a twenty-year interval resulted
in an imperceptible increase in risk to the public. NUREG-1493 found
that, generically, the design containment leakage rate contributes
about 0.1 percent of the overall risk and that decreasing the Type A
testing frequency would have a minimal effect on this risk, since 95
percent of the Type A detectable leakage paths would already be
detected by Type B and Type C testing. Similar proposed changes have
been previously reviewed and approved by the NRC, and they are
applicable to McGuire. Based upon the preceding discussion, Duke
Energy Corporation [Duke Power Company, LLC] has concluded that the
proposed amendments do not involve a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Evangelos C. Marinos.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana.
Date of amendment request: November 15, 2007.
Description of amendment request: The proposed change would
relocate Surveillance Requirement (SR) 3.8.3.6 from the Technical
Specifications (TS) to a licensee-controlled document. SR 3.8.3.6
requires the Emergency Diesel Generator (EDG) Fuel Oil Storage Tanks
(FOSTs) to be drained, sediment removed, and cleaned on a 10-year
interval.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The FOSTs provide the storage for the EDG fuel oil, assuring an
adequate volume is available for each EDG to operate for seven days
in the event of a loss of offsite power concurrent with a loss of
coolant accident. The relocation of the SR to drain and clean the
FOSTs will not impact any of the previously analyzed accidents.
Sediment in the tank, or failure to perform this SR, does not
necessarily result in an inoperable storage tank. Fuel oil quantity
and quality are assured by other TS SRs which remain unchanged.
These SRs help ensure tank sediment is minimized and ensure that any
[[Page 74358]]
degradation of the tank wall surface that results in a fuel oil
volume reduction is detected and corrected in a timely manner. As a
result, adequate controls exist to allow relocation of this
preventative maintenance cleaning requirement to licensee controlled
documents.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed TS changes do not involve the addition or
modification of any plant equipment. Also, the proposed change will
not alter the design configuration, or method of operation of plant
equipment beyond its normal functional capabilities. The proposed TS
change does not create any new credible failure mechanisms,
malfunctions or accident initiators.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not alter or exceed a design basis or
safety limit. Diesel generator fuel oil quantity and quality will
continue to be maintained within acceptable limits of the TS to
assure the ability of the EDG to perform its intended function.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hilt.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi.
Date of amendment request: December 5, 2007.
Description of amendment request: The proposed amendment would
change the Grand Gulf Nuclear Station, Unit 1 (GGNS), Technical
Specification (TS) 5.6.5, ``Core Operating Limits Report (COLR),'' to
add a reference to an analytical method that will be used to determine
core operating limits. The new reference, NEDC-33383P, ``GEXL97
Correlation Applicable to ATRIUM-10 Fuel,'' will allow Entergy
Operations, Inc. (Entergy) to use a Global Nuclear Fuel (GNF) method to
determine fuel assembly critical power of AREVA ATRIUM-10 fuel. GGNS
currently operates with a full core of ATRIUM-10 fuel. Entergy plans to
use the GEXL97 correlation for GGNS operating Cycle 17 currently
scheduled to begin in the fall 2008. Additionally, an administrative
change is proposed to an existing reference in TS 5.6.5.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Core operating limits are established each operating cycle in
accordance with TS 3.2, ``Power Distribution'' and TS 5.6.5, ``Core
Operating Limits Report (COLR)''. These core operating limits ensure
that the fuel design limits are not exceeded during any conditions
of normal operation or in the event of any Anticipated Operational
Occurrence (AOO). The methods used to determine the operating limits
are those previously found acceptable by the NRC and listed in TS
Section 5.6.5.b.
A change to TS 5.6.5.b is requested to include an additional
reference to the list of analytical methods. GGNS currently operates
with a full core of AREVA ATRIUM-10 fuel but is scheduled to load
GE14 fuel during the next refueling outage. GGNS plans to use the
analysis methods of the new fuel vendor, GNF for the analysis of the
mixed core. The GEXL97 correlation accurately models predicted core
behavior and appropriately determines the overall critical power
uncertainty of the method. In addition, the GEXL97 application range
covers the range of expected operation of the ATRIUM-10 fuel during
normal steady state and transient conditions in the GGNS reload
cores. Although a depressurization transient could result in vessel
pressures below the range of GEXL97, the transient would not
threaten fuel cladding integrity, since the margin to the MCPR
[minimum critical power ratio] safety limit increases with
decreasing reactor pressure.
Additionally, Entergy proposes an administrative change to the
GESTAR-Il reference in TS 5.6.5.b. The administrative change does
not alter any method of analysis as described in the NRC approved
versions of GESTAR-II. The requested TS changes concern the use of
analytical methods and do not involve any plant modifications or
operational changes that could affect any postulated accident
precursors or accident mitigation systems and do not introduce any
new accident initiation mechanisms. The proposed changes have no
effect on the type or amount of radiation released, and have no
effect on predicted offsite doses in the event of an accident. Thus,
the proposed change does not affect the probability of an accident
previously evaluated nor does it increase the radiological
consequences of any accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed TS changes will not change the design function,
reliability, performance, or operation of any plant systems,
components, or structures. It does not create the possibility of a
new failure mechanism, malfunction, or accident initiators not
considered in the design and licensing bases. Plant operation will
continue to be within the core operating limits that are established
using NRC approved methods that are applicable to the GGNS design
and the GGNS fuel.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change adds GEXL97 to the list of analytical
methods in TS 5.6.5.b that can be used to determine core operating
limits. Use of the GEXL97 correlation analytical method provides an
equivalent level of protection as that currently provided. The
administrative change does not alter any method of analysis as
described in the NRC approved versions of GESTAR-II. The proposed
change does not modify the safety limits or set points at which
protective actions are initiated, and does not change the
requirements governing operation or availability of safety equipment
assumed to operate to preserve the margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hiltz.
[[Page 74359]]
Exelon Generation Company, LLC, Docket No. 50-353, Limerick Generating
Station, Unit 2, Montgomery County, Pennsylvania.
Date of amendment request: November 16, 2007.
Description of amendment request: The proposed changes revise
technical specification (TS) action requirements associated with
inoperable reactor coolant system (RCS) leakage detection systems. A
new TS action requirement is proposed that will address the
inoperability of the drywell unit cooler condensate flow rate
monitoring system concurrent with one other RCS leakage detection
system, other than the primary containment atmosphere gaseous
radioactivity monitoring system. This would relax the allowed out-of-
service time for the specified combination of systems and is related to
the current inoperability of the drywell unit cooler condensate flow
rate monitoring system. The proposed changes would be effective for the
remainder of the current operating cycle (Cycle 10), which is currently
scheduled to end in the spring of 2009, or until the next shutdown of
sufficient duration to allow for drywell unit cooler condensate flow
rate monitoring system repairs, whichever comes first.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes continue to maintain an acceptable level of
reactor coolant system (RCS) leakage detection instrumentation
required to support plant operations. The level of RCS leakage
detection capability inherent with the proposed changes will
continue to provide acceptable early warning detection of potential
RCS pressure boundary degradation. The proposed changes do not
impact the physical configuration or design function of plant
structures, systems, or components (SSCs) or the manner in which
SSCs are operated, modified, tested, or inspected [with the
exception of an increase in allowed out-of-service time for a
concurrent inoperability of the drywell unit cooler condensate flow
rate monitoring system and another specified RCS leakage detection
system]. The proposed changes do not impact the initiators or
assumptions of analyzed events, nor do they impact mitigation of
accidents or transient events. Therefore, the proposed changes do
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes only affect systems associated with the
detection of leakage resulting from the degradation of the RCS
pressure boundary. The proposed changes do not alter plant
configuration or require that new plant equipment be installed. The
RCS leakage detection systems will continue to function as designed
in all modes of operation. No new accident type is created as a
result of the proposed changes. No new failure mode for any
equipment is created. The proposed changes do not alter assumptions
made about accidents previously evaluated. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not involve any physical changes to
plant SSCs or the manner in which SSCs are operated, modified,
tested, or inspected. The proposed changes do not involve a change
to any safety limits, limiting safety system settings, limiting
conditions of operation, or design parameters for any SSC. The
proposed changes do not impact any safety analysis assumptions and
do not involve a change in initial conditions, system response
times, or other parameters affecting an accident analysis.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, with changes as noted above, it appears that the three
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: J. Bradley Fewell, Esquire, Associate
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois.
Date of amendment request: October 9, 2007.
Description of amendment request: A change is proposed to the
technical specifications (TS) of Quad Cities Nuclear Power Station
(QCNPS), Units 1 and 2, consistent with Technical Specifications Task
Force (TSTF) Change Traveler TSTF-423 to the standard TSs for boiling
water reactor plants, to allow, for some systems, entry into hot
shutdown rather than cold shutdown to repair equipment, if risk is
assessed and managed consistent with the program in place for complying
with the requirements of Title 10 of the Code of Federal Regulations
(10 CFR) Section 50.65(a)(4). Changes proposed herein will be made to
the QCNPS, Units 1 and 2, TSs for selected required action end states
providing this allowance.
The licensee reviewed the proposed no significant hazards
consideration (NSHC) determination published in the Federal Register on
March 23, 2007 (71 FR 14726) and concluded that it is applicable to
QCNPS, Units 1 and 2. The licensee incorporated the proposed
determination by reference to satisfy the requirements of 10 CFR
50.91(a).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC is presented below:
1. The Proposed Change Does Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously
Evaluated.
The proposed change allows a change to certain required end
states when the TS Completion Times for remaining in power operation
will be exceeded. Most of the requested technical specification (TS)
changes are to permit an end state of hot shutdown (Mode 3) rather
than an end state of cold shutdown (Mode 4) contained in the current
TS. The request was limited to: (1) Those end states where entry
into the shutdown mode is for a short interval, (2) entry is
initiated by inoperability of a single train of equipment or a
restriction on a plant operational parameter, unless otherwise
stated in the applicable technical specification, and (3) the
primary purpose is to correct the initiating condition and return to
power operation as soon as is practical. Risk insights from both the
qualitative and quantitative risk assessments were used in specific
TS assessments. Such assessments are documented in Section 6 of GE
NEDC-32988, Revision 2, ``Technical Justification to Support Risk
Informed Modification to Selected Required Action End States for BWR
Plants.'' They provide an integrated discussion of deterministic and
probabilistic issues, focusing on specific technical specifications,
which are used to support the proposed TS end state and associated
restrictions. The staff finds that the risk insights support the
conclusions of the specific TS assessments. Therefore, the
probability of an accident previously evaluated is not significantly
increased, if at all. The consequences of an accident after adopting
proposed TSTF-423, are no different than the consequences of an
accident prior to adopting TSTF-423. Therefore, the consequences of
an accident previously evaluated are not significantly affected by
this change. The addition of a requirement to assess and manage the
risk introduced by this change will further minimize possible
concerns. Therefore, this
[[Page 74360]]
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The Proposed Change Does Not Create the Possibility of a New
or Different Kind of Accident From Any Previously Evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
If risk is assessed and managed, allowing a change to certain
required end states when the TS Completion Times for remaining in
power operation are exceeded, i.e., entry into hot shutdown rather
than cold shutdown to repair equipment, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change and the commitment by the licensee to adhere to the guidance
in TSTF-IG-05-02, Implementation Guidance for TSTF-423, Revision 0,
``Technical Specifications End States, NEDC-32988-A,'' will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
3. The Proposed Change Does Not Involve a Significant Reduction
in the Margin of Safety.
The proposed change allows, for some systems, entry into hot
shutdown rather than cold shutdown to repair equipment, if risk is
assessed and managed. The BWROG's [Boiling Water Reactor Owners
Group's] risk assessment approach is comprehensive and follows staff
guidance as documented in RGs [Regulatory Guides] 1.174 and 1.177.
In addition, the analyses show that the criteria of the three-tiered
approach for allowing TS changes are met. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in RG 1.177. A risk assessment was performed to justify
the proposed TS changes. The net change to the margin of safety is
insignificant. Therefore, this change does not involve a significant
reduction in a margin of safety.
Therefore, the NRC staff proposes to determine that the requested
amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Steam Electric Station, Units 1 and 2, Somervell County,
Texas
Date of amendment request: November 29, 2007.
Brief description of amendments: Revision to Technical
Specification (TS) 3.6.7, (``Spray Additive System,'' to allow
modifications to the facility potentially required to comply with U.S.
Nuclear Regulatory Commission (NRC) Generic Letter 2004-02, ``Potential
Impact of Debris Blockage on Emergency Recirculation during Design
Basis Accident at Pressurized Water Reactors.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No
The proposed change[s] [do] not impact the initiation or
probability of occurrence of any accident.
The accidents evaluated in the Final Safety Analysis report
(FSAR) that could be affected by this proposed change are those
involving the pressurization of the containment and those involving
recirculation of fluid within the Emergency Core Cooling System
(ECCS) or the Containment Spray System (e.g., loss of coolant
accidents (LOCAs)).
The change to a minimum pH [potential of Hydrogen] of 7.1 will
not result in a significant increase in the radiological
consequences of a LOCA as-described below.
The equilibrium spray pH during the recirculation phase
resulting from this change will be greater than or equal to 7.1. The
pH range for the spray will be bounded by the water spray solution
which is borated water with a maximum of 2600 parts per million
(ppm) boron buffered to a final spray solution pH much less than the
10.5 as described in the current FSAR Section 3.11(B) for the
postulated spray solution environment. The maximum pH is the
limiting parameter for equipment qualification. Since the resulting
pH level will be closer to neutral using the lower limit of 7.1,
post-LOCA corrosion of containment components will not be increased.
Post-LOCA hydrogen generation will be reduced. There will not be an
adverse radiation dose effect on any safety-related equipment. Thus,
the potential for failures of the ECCS or safety-related equipment
following a LOCA will not be increased as a result of the proposed
change.
This modification affects the Containment Spray System which is
intended to respond to and mitigate the effects of a LOCA. The
Containment Spray System will continue to function in a manner
consistent with the plant design basis. There will be no degradation
in the performance of nor an increase in the number of challenges to
equipment assumed to function during an accident situation.
Therefore, these Technical Specification (TS) revisions do not
affect the probability of any event initiators. There will be no
adverse changes to normal plant operating parameters, Engineered
Safety Features (ESF) actuation setpoints, or accident mitigation
capabilities.
The proposed change allows the Spray Additive System currently
used to mitigate the consequences of an accident to maintain the
equilibrium sump pH at greater than or equal to 7.1 to minimize
chloride-induced stress corrosion cracking in austenitic stainless
components important to safety located inside containment.
Therefore, the proposed changes will not increase the probability of
an accident or malfunction of equipment important to safety
previously evaluated in the FSAR.
The offsite and control room doses will continue to meet the
requirements of [Title 10 of the Code of Federal Regulations (10
CFR) part 100] 10 CFR 100, 10 CFR 50 Appendix A [General Design
Criterion] GDC 19, [Standard Review Plan] SRP 15.6.5.11, and SRP
6.4.11. The proposed new pH limit will provide satisfactory
retention of iodine in the sump water, as well as provide adequate
pH control to minimize the potential of chloride-induced stress
corrosion cracking of austenitic stainless steel components.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change to the revised Surveillance for the
Containment Spray Additive System provides for a required minimum
equilibrium pH in containment post accident. There are no electrical
or mechanical components being added whose failure could prevent the
system from functioning.
No new accident scenarios, transient precursors, or limiting
single failures are introduced as a result of the proposed changes.
There will be no adverse effect or challenges imposed on any safety-
related system as a result of this proposed change. The amount of
sodium hydroxide (NaOH) will provide a minimum equilibrium sump pH
of 7.1 following mixing. Therefore, the possibility of a new or
different type of accident is not created.
There are no changes which would cause the malfunction of
safety-related equipment, assumed to be operable in the accident
analyses, as a result of the proposed Technical Specification
changes. The possibility of a malfunction of safety-related
equipment with a different result is not created.
Therefore, the proposed change[s] [do] not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No
The only function of the chemical additive system is to provide
pH control of the post-accident containment recirculation sump
water, since the borated water from the Refueling Water Storage Tank
(RWST) used as the containment spray pump suction source during
injection is sufficient to remove iodine from the containment
atmosphere following a LOCA. The net effect on the pH control
function of reducing the
[[Page 74361]]
amount of buffer is that the equilibrium sump pH will be lowered to
a minimum of 7.1. There will be no change to the current Technical
Specification acceptance limits on RWST volume and boron
concentration. The resulting equilibrium sump pH level from this
change will be closer to neutral; therefore, the post-LOCA corrosion
of containment components will not be increased (i.e., would be
reduced).
Because the long term pH will be maintained greater than or
equal to 7.1, margin to minimize the potential for stress corrosion
cracking is maintained.
The radiological analysis, as discussed in the technical
analysis above, is shown not to be impacted. There will be no change
to the [departure from nucleate boiling ratio] DNBR Correlation
Limit, the design DNBR limits, or the safety analysis DNBR limits
discussed in Bases Section 2.1.1. There will be no effect on the
manner in which Safety Limits or Limiting Safety System Settings are
determined nor will there be any effect on those plant systems
necessary to assure the accomplishment of protection functions.
There will be no adverse impact on Departure of Nucleate Boiling
Ratio limits, [heat flux hot channel factor] FQ, [nuclear
enthalpy rise hot channel factor] F-delta-H, LOCA peak cladding
temperature, peak local power density, or any other margin of
safety.
Therefore the proposed change[s] [do] not involve a reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Timothy P. Matthews, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: Thomas G. Hiltz.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California.
Date of amendment requests: October 2, 2007.
Description of amendment requests: The proposed amendments would
revise Technical Specification (TS) 3.5.4, ``Refueling Water Storage
Tank (RWST),'' Surveillance Requirement (SR) 3.5.4.2, to increase the
minimum required borated water volume from ``>= [greater than or equal
to] 400,000 gallons (81.5% indicated level)'' to ``>= 455,300 gallons
(93.6% level),'' to reflect the new sump design required to comply with
U.S. Nuclear Regulatory Commission (NRC) Generic Letter 2004-02,
``Potential Impact of Debris Blockage on Emergency Recirculation during
Design-Basis Accident at Pressurized Water Reactors.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. [Do] the proposed change[s] involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change[s] [revise] the minimum RWST borated water
volume. The RWST borated water volume is not an initiator of any
accident previously evaluated. As a result, the probability of an
accident previously evaluated is not affected. The proposed
change[s] [do] not alter or prevent the ability of structures,
systems, and components from performing their intended function to
mitigate the consequences of an initiating event within the assumed
acceptance limits. The effect on containment flood level, equipment
qualification, and containment sump pH remain within the limits
assumed in the design and accident analyses. The proposed change[s]
[do] not affect the source term, containment isolation, or
radiological release assumptions used in evaluating the radiological
consequences of an accident previously evaluated. Further, the
proposed change[s] [do] not increase the types or amounts of
radioactive effluent that may be released offsite, nor significantly
increase individual or cumulative occupational/public radiation
exposures. The proposed change[s] are consistent with the safety
analysis assumptions and resultant consequences.
Therefore, the proposed change[s] [do] not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. [Do] the proposed change[s] create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The change[s] [do] not involve a physical alteration of the
plant (i.e., no new or different components or physical changes are
involved with this change) or a change in the methods governing
normal plant operation. The change[s] [do] not alter any assumptions
made in the safety analysis.
Therefore, the proposed change[s] will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. [Do] the proposed change[s] involve a significant reduction
in a margin of safety?
Response: No.
The proposed change[s] to revise the required RWST minimum
borated water volume [do] not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not affected by [these] change[s]. The proposed change[s] will
not result in plant operation in a configuration outside of the
design basis.
Therefore, the proposed change[s] [do] not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c)
are satisfied. Therefore, the NRC staff proposes to determine that
the amendment requests involve no significant hazards consideration.
Attorney for licensee: Jennifer Post, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Branch Chief: Thomas G. Hiltz.
Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power
Plant (HBPP), Unit 3 Humboldt County, California.
Date of amendment request: November 5, 2007.
Description of amendment request: The licensee has proposed
amending the technical specifications (TS) to delete many
operational and administrative requirements upon transfer of spent
nuclear fuel assemblies and fuel fragment containers from the Spent
Fuel Pool (SFP) to the Humboldt Bay Independent Spent Fuel Storage
Installation (ISFSI). Some TS requirements will be relocated to the
HBPP Quality Assurance Plan.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes reflect the transfer of spent fuel from the
Spent Fuel Pool to the Humboldt Bay (HB) Independent Spent Fuel
Storage Installation. Design basis accidents related to the SFP are
discussed in the Humboldt Bay Power Plant Unit 3 Defueled Safety
Analysis Report (DSAR). These postulated accidents are predicated on
spent fuel being stored in the SFP. With the removal of the spent
fuel from the SFP, there are no important-to-safety systems,
structures or components required to function or to be monitored. In
addition, there are no remaining credible accidents involving spent
fuel or the SFP that require actions of a Certified Fuel Handler or
Noncertified Fuel Handler to prevent occurrence or to mitigate
consequences. The proposed change to the Design Features section of
the Technical Specifications (TS) clarifies that the spent fuel is
being stored in dry casks within an ISFSI. The probability or
consequences of accidents at the ISFSI are evaluated in the HB ISFSI
Final Safety Analysis Report (FSAR) and are independent of the
accidents evaluated in the HBPP Unit 3 DSAR. Therefore, the proposed
changes will not involve a significant increase in the probability
or consequences of an accident previously evaluated.
(2) Does the change create the possibility of a new or different
kind of accident from any accident evaluated?
Response: No.
The proposed changes reflect the reduced operational risks as a
result of the spent fuel being transferred to dry casks within an
ISFSI. The proposed changes do not modify any systems, structures or
components. The
[[Page 74362]]
plant conditions for which the HBPP Unit 3 DSAR design basis
accidents relating to spent fuel and the SFP have been evaluated are
no longer applicable. The aforementioned proposed changes do not
affect any of the parameters or conditions that could contribute to
the initiation of an accident. Design basis accidents associated
with the dry cask storage of spent fuel are already considered in
the HB ISFSI FSAR. No new accident scenarios are created as a result
of deleting nonapplicable operational and administrative
requirements. Therefore, the proposed changes will not create the
possibility of a new or different kind of accident from those
previously evaluated.
(3) Does the change involve a significant reduction in a margin
of safety?
Response: No.
The proposed changes reflect the reduced operational risks as a
result of the spent fuel being transferred to dry casks within an
ISFSI. The design basis and accident assumptions within the HBPP
Unit 3 DSAR and the TS relating to spent fuel are no longer
applicable. The proposed changes do not affect remaining plant
operations, nor structures, systems, or components supporting
decommissioning activities. In addition, the proposed changes do not
result in a change in initial conditions, system response time, or
in any other parameter affecting the course of a decommissioning
activity accident analysis. Therefore, the proposed changes will not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Jennifer K. Post, Pacific Gas and
Electric Company, 77 Beale Street, B30A, San Francisco, CA.
NRC Branch Chief: Andrew Persinko.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri.
Date of amendment request: October 31, 2007.
Description of amendment request: The amendment would revise
Technical Specification (TS) 3.8.1, ``Essential Service Water System
(ESW),'' and TS 3.8.1, ``AC [Alternating Current] Sources--Operating.''
A note would be added to Condition A, one ESW train inoperable, of TS
3.8.1, and Condition B, one diesel generator (DG) inoperable, of TS
3.8.1 would be revised. The revisions are to allow a one-time
completion time extension from 72 hours to 14 days to support a planned
replacement of ESW piping prior to December 31, 2008, in the licensee's
fall 2008 refueling outage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
[The only change to the plant is that existing ESW piping will
be replaced in the fall 2008 refueling outage. There are no other
changes to the plant and no hardware or equipment will be added to
the plant. This replacement is to address localized degradation of
the ESW piping due to microbiologically induced corrosion.]
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since no
hardware changes are proposed to the protection systems. The same
reactor trip system (RTS) and engineered safety feature actuation
system (ESFAS) instrumentation will continue to be used. The
protection systems will continue to function in a manner consistent
with the plant design basis. The use of polyethylene (PE) piping
[(i.e., replacing existing ESW piping by PE piping)] in the ESW
system in accordance with ASME [American Society of Mechanical
Engineers Boiler and Pressure Vessel Code] Code Case N-755, with
justified materials and design exceptions as noted in [the
licensee's letter dated August 30, 2007 (ULNRC-05434), which
requested relief from the ASME Code to replace the ESW piping by the
PE piping], will [have the PE piping that replaces the ESW piping]
provide an acceptable level of quality and safety. There will be no
changes to the essential service water (ESW) system or [the]
ultimate heat sink (UHS) surveillance and operating limits. [The
licensee's letter dated August 30, 2007,] demonstrates the
acceptability of using the PE piping in this buried ASME Class 3
application [(i.e., replacing existing ESW piping)].
The proposed changes will not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, and configurations of the facility or the manner in
which the plant is operated and maintained. The proposed changes
will not alter or prevent the ability of structures, systems, and
components (SSCs) from performing their intended [safety] functions
to mitigate the consequences of an initiating event within the
assumed acceptance limits.
The proposed changes do not affect the way in which safety-
related systems perform their [safety] functions.
All accident analysis acceptance criteria will continue to be
met with the proposed changes. The proposed changes will not affect
the source term, containment isolation, or radiological release
assumptions used in evaluating the radiological consequences of an
accident previously evaluated. The proposed changes will not alter
any assumptions or change any mitigation actions in the radiological
consequence evaluations in the FSAR [Final Safety Analysis Report
for the Callaway Plant].
The applicable radiological dose acceptance criteria [is
unchanged] and will continue to be met.
Therefore, the proposed changes do 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.
There are no proposed changes in the method by which any safety-
related plant SSC performs its safety function. [The proposed
changes will not affect the performance of the ESW piping in terms
of providing mitigation of design basis accidents per the FSAR
accident analyses.] The proposed changes will not affect the normal
method of plant operation or change any operating parameters. No
equipment performance requirements will be affected. The proposed
changes will not alter any assumptions made in the safety analyses.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures will be introduced as a
result of this amendment. There will be no adverse effect or
challenges imposed on any safety-related system as a result of this
amendment.
The proposed amendment will not alter the design or performance
of the 7300 Process Protection System, Nuclear Instrumentation
System, or Solid State Protection System used in the plant
protection systems.
Therefore, 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.
There will be no effect on those plant systems necessary to
assure the accomplishment of protection functions. There will be no
impact on the overpower limit, departure from nucleate boiling ratio
(DNBR) limits, heat flux hot channel factor (FQ), nuclear
enthalpy rise hot channel factor (F [delta] H), loss of coolant
accident peak cladding temperature (LOCA PCT), peak local power
density, or any other margin of safety. The applicable radiological
dose consequence acceptance criteria will continue to be met. [The
proposed changes will not affect the performance of the ESW piping
in terms of providing mitigation of design basis accidents per the
FSAR accident analyses.]
The proposed changes do not eliminate any surveillances or alter
the frequency of [any] surveillances required by the Technical
Specifications. None of the acceptance criteria for any accident
analyses will be changed.
The proposed changes will have no impact on the radiological
consequences of a design basis accident.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this
[[Page 74363]]
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Thomas G. Hiltz.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed no Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida.
Date of application for amendment: November 12, 2007.
Brief description of amendment: Use of alternate method of
monitoring rod position for a control rod or shutdown rod with an
inoperable rod position indicator.
Date of publication of individual notice in the Federal Register:
November 28, 2007 (72 FR 67323).
Expiration date of individual notice: December 28, 2007 (Public
comments) and January 28, 2008 (Hearing requests).
Notice of Issuance of amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina.
Date of application of amendments: January 31, 2007.
Brief description of amendments: The amendments revised the
Technical Specifications to remove requirements that are no longer
applicable due to the completion of the control room intake/booster fan
modifications.
Date of Issuance: December 11, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 358, 360, and 359.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: October 9, 2007 (72 FR
57353) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 11, 2007.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington.
Date of application for amendment: July 30, 2007, as supplemented
by letter dated November 6, 2007.
Brief description of amendment: The changes revise Technical
Specification (TS) 1.4, ``Frequency,'' TS 3.1.5, ``Control Rod Scram
Accumulators,'' TS 3.4.1, ``Recirculation Loops Operating,'' TS 3.5.1,
``ECCS [Emergency Core Cooling System]--Operating,'' TS 3.5.2, ``ECCS--
Shutdown,'' TS 3.7.1, ``Standby Service Water (SW) System and Ultimate
Heat Sink (UHS),'' TS 3.8.1, ``AC [Alternating Current] Sources--
Operating,'' TS 3.8.2, ``AC Sources--Shutdown,'' and TS 5.5.6, ``In-
service Testing Program.'' The changes include updates to adopt
approved TS Task Force (TSTF) Travelers 284, Revision 3, ``Add `Met'
vs. `Perform' to Specification 1.4, Frequency,'' TSTF-479, Revision 0,
``Changes to Reflect Revision of 10 CFR 50.55a,'' and TSTF-485,
Revision 0, ``Correct Example 1.4-1,'' and TSTF-497, Revision 0,
``Limit Inservice Testing Program SR [Surveillance Requirement] 3.0.2
Application to Frequencies of 2 Years or Less.''
Date of issuance: December 13, 2007.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 205.
Facility Operating License No. NPF-21: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 28, 2007 (72 FR
49572). The supplement dated November 6, 2007, 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 as
initially published in the Federal Register. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
December 13, 2007.
No significant hazards consideration comments received: No.
[[Page 74364]]
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana.
Date of amendment request: July 2, 2007.
Brief description of amendment: The amendment modified River Bend
Station, Unit 1, technical specifications (TSs) requirements for MODE
change limitations in Limiting Condition for Operation 3.0.4 and
Surveillance Requirement 3.0.4. The TS changes are consistent with
Revision 9 of NRC-approved Industry TS Task Force (TSTF) Standard TS
Change Traveler, TSTF-359, ``Increase Flexibility in MODE Restraints.''
In addition, the amendment also changed TS Section 1.4, ``Frequency,''
Example 1.4-1, ``Surveillance Requirements,'' to accurately reflect the
changes made by TSTF-359, which is consistent with NRC-approved TSTF-
485, Revision 0, ``Correct Example 1.4-1.''
Date of issuance: December 6, 2007.
Effective date: As of the date of issuance and shall be implemented
120 days from the date of issuance.
Amendment No.: 156.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: September 11, 2007 (72
FR 51856).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 6, 2007.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315, Donald C. Cook
Nuclear Plant, Units 1 and 2 (DCCNP-1 and DCCNP-2), Berrien County,
Michigan.
Date of application for amendments: September 15, 2006, as
supplemented on July 25 and October 9, 2007.
Brief description of amendments: The amendments revised the DCCNP-1
and DCCNP-2 Technical Specifications (TS) to allow certain functions in
the reactor protection system and engineered safety feature actuation
system instrumentation which have installed bypass test capability to
be tested in bypass. The licensee's request to correct the
administrative error will be reviewed and resolved by separate
correspondence.
Date of issuance: December 17, 2007.
Effective date: As of the date of issuance, and shall be
implemented within 45 days.
Amendment No.: 300, 283.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revise the License Page and Technical Specifications.
Date of initial notice in Federal Register: November 21, 2006 (71
FR 67396)
The supplemental letters contained clarifying information, did not
change the initial no significant hazards consideration determination,
and did not expand the scope of the original Federal Register notice.
The Commission's related evaluation of the amendment is contained in a
safety evaluation dated December 17, 2007.
No significant hazards consideration comments received: No.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell
County, Texas.
Date of amendment request: December 19, 2006.
Brief description of amendments: The amendments revised Technical
Specification 5.5.16, ``Containment Leakage Rate Testing Program,'' for
consistency with the requirements of paragraph 50.55a(g)(4) of Title 10
of the Code of Federal Regulations for components classified as Code
Class CC.
Date of issuance: December 13, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: Unit 1-141; Unit 2-141.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: May 8, 2007 (72 FR
26179). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 13, 2007.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: August 16, 2007, as supplemented by
letter date November 5, 2007.
Brief description of amendment: The amendment revised Technical
Specification 5.5.6, ``Inservice Testing Program,'' to allow a one-time
extension of the 5-year frequency requirement for setpoint testing of
safety valve MS-RV-70ARV.
Date of issuance: December 4, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 228.
Facility Operating License No. DPR-46: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: September 25, 2007 (72
FR 54476). The supplement dated November 5, 2007, 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 as
initially published in the Federal Register. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
December 4, 2007.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: February 28, 2007, as supplemented by
letter dated May 22, 2007.
Brief description of amendments: The amendments revise the language
in the Technical Specifications to conform to the licensing basis as
established by Amendment Nos. 87 and 74, for Units 1 and 2,
respectively, dated May 27, 1997.
Date of issuance: December 6, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: Unit 1-181; Unit 2-168.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: May 22, 2007 (72 FR
28723). The supplement dated May 22, 2007, 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 as
published in the Federal Register. The Commission's related evaluation
of the amendments is contained in a Safety Evaluation dated December 6,
2007.
No significant hazards consideration comments received: No.
[[Page 74365]]
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant (PINGP), Units 1 and 2, Goodhue County,
Minnesota.
Date of application for amendments: December 14, 2006, supplemented
by letter dated November 13, 2007.
Brief description of amendments: The amendments revise the sump
debris interceptor nomenclature in PINGP Unit 1 and Unit 2 Technical
Specifications (TS) 3.5.2 to more clearly reflect the configuration of
the new Emergency Core Cooling System sump strainers that were
installed to address Generic Letter 2004-02, ``Potential Impact of
Debris Blockage on Emergency Recirculation During Design Basis
Accidents at Pressurized-Water Reactors.'' The amendments also revise
the required Refueling Water Storage Tank (RWST) water level in TS
3.5.4 to reflect the administratively controlled water inventory in the
RWST.
Date of issuance: December 14, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 182/172.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Facility Operations License and Technical Specifications.
Date of initial notice in Federal Register: February 27, 2007 (72
FR 8804)
The supplemental letter contained clarifying information and did
not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice. The Commission's related evaluation of the amendments
is contained in a Safety Evaluation dated December 14, 2007.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska.
Date of amendment request: July 31, 2007.
Brief description of amendment: The amendment revises Technical
Specification (TS) 2.7(1), (Electrical Systems--Minimum Requirements,''
TS 2.7(2), (``Electrical Systems--Modification of Minimum
Requirements,'' and TS 3.7(5), ``Emergency Power System Periodic
Tests--Required Safety Related Inverters.'' The licensee is adding two
safety-related swing inverters to the 120 Volt alternating current
instrument buses. The TS changes reflect modifications made to the
plant and are needed to take advantage of the additional operational
flexibility the swing inverters will provide.
Date of issuance: December 17, 2007.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 251.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: August 28, 2007 (72 FR
49582). The Commission's related evaluation of the amendment is
contained in a safety evaluation dated December 17, 2007.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 21st day of December, 2007.
For The Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E7-25416 Filed 12-28-07; 8:45 am]
BILLING CODE 7590-01-P