[Federal Register Volume 72, Number 156 (Tuesday, August 14, 2007)]
[Notices]
[Pages 45454-45466]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-15459]
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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 July 19, 2007, to August 1, 2007. The last
biweekly notice was published on July 31, 2007 (72 FR 41780).
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 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.
[[Page 45455]]
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, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
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 a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this 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].
[[Page 45456]]
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: June 29, 2007.
Description of amendments request: The amendment would modify
Technical Specification (TS) requirements related to control room
envelope (CRE) habitability in TS 3.7.8, ``Control Room Emergency
Ventilation System (CREVS),'' and TS 5.5, ``Programs and Manuals.'' The
changes are consistent with the Nuclear Regulatory Commission approved
Technical Specification Task Force (TSTF)-448, Revision 3, ``Control
Room Habitability.'' The availability of the TS improvement was
published in the Federal Register on January 17, 2007 (72 FR 2022) as
part of the consolidated item improvement process (CLIIP). In addition,
the amendment would remove a footnote currently contained in the
Completion Time of TS 3.7.8, Required Action D. The footnote was added
in Amendment Nos. 250/227 and was only applicable during the Unit 1
2002 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 proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
The removal of a footnote [to TS 3.7.8] that is no longer
applicable is an editorial change that does not affect accident
initiators or precursors, nor alter the design assumptions,
conditions or configuration of the facility. The proposed change
also does not affect the ability of SSCs to perform their intended
function to mitigate the consequences of an accident. Therefore, the
proposed editorial change does not increase the probability or
consequences of an 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 change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice.
The proposed change is the editorial removal of a footnote [to
TS 3.7.8] that no longer applies. The removal of a footnote that no
longer applies does not impact the accident analyses. Additionally,
it does not add or modify any existing plant equipment and does not
introduce any new operational methods. Therefore, the proposed
editorial change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Therefore, this change does 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 does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition.
The proposed editorial change [removal of a footnote to TS
3.7.8] does not affect safety analyses acceptance criteria or safety
system operation. Removal of a footnote that is no longer applicable
does not result in plant operation outside the design basis.
Therefore, the proposed editorial change does not involve a
reduction in 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: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Generation Group, LLC, 750 East Pratt Street,
17th floor, Baltimore, MD 21202.
NRC Branch Chief: Mark G. Kowal.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: February 20, 2007.
Description of amendment request: The proposed change would revise
Limerick Generating Station (LGS), Units 1 and 2, Technical
Specifications (TSs), Section 6.8.4.g, ``Primary Containment Leakage
Rate Testing Program,'' to allow a one-time extension of no more than 5
years for the Type A, Integrated Leakage Rate Test (ILRT) interval.
This revision is a one-time exception to the 10-year frequency of the
performance-based leakage rate testing program for Type A tests as
defined in Nuclear Energy Institute (NEI) document NEI 94-01, Revision
0, ``Industry Guideline For Implementing Performance-Based Option of 10
CFR Part 50, Appendix J,'' pursuant to Title 10 of the Code of Federal
Regulations (10 CFR) Part 50, Appendix J, Option B. The requested
exception is to allow the ILRT to be performed within 15 years from the
last ILRT as opposed to the current 10-year frequency. The most recent
containment Type A ILRTs for LGS Units 1 and 2 were performed on May
15, 1998, and May 21, 1999, respectively.
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 amendment involve a significant increase in
the probability or
[[Page 45457]]
consequences of an accident previously evaluated?
Response: No.
The proposed change will revise TS 6.8.4.g (``Primary
Containment Leakage Rate Testing Program'') of the LGS, Units 1 and
2 TS to reflect a one-time extension to the Type A Integrated Leak
Rate Test (ILRT) as currently specified in the Technical
Specifications. This change will extend the requirement to perform
the Type A ILRT from the current requirement of 10 to 15 years,
which is ``no later than May 15, 2013'' for LGS, Unit 1 and is ``no
later than May 21, 2014'' for Unit 2.
The function of the containment is to isolate and contain
fission products released from the reactor coolant system following
a design basis Loss of Coolant Accident (LOCA) and to confine the
postulated release of radioactive material to within limits. The
test interval associated with Type A ILRTs is not a precursor of any
accident previously evaluated. Type A ILRTs provide assurance that
the LGS, Units 1 and 2 containments will not exceed allowable
leakage rate values specified in the TS and will continue to perform
their design function following an accident. The risk assessment of
the proposed change has concluded that there is an insignificant
increase in Large Early Release Frequency, Person-Rem, and
Conditional Containment Failure Frequency. Additionally, containment
inspections have also been performed which demonstrate the continued
structural integrity of the primary containment and will be
performed in the future as required by the ASME Code.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change for a one-time extension of the Type A ILRTs
for LGS, Units 1 and 2 will not affect the control parameters
governing unit operation or the response of plant equipment to
transient and accident conditions. The proposed change does not
introduce any new equipment, modes of system operation or failure
mechanisms.
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 amendment involve a significant reduction
in a margin of safety?
Response: No.
The integrity of the containment penetrations and isolation
valves is verified through Type B and Type C local leak rate tests
(LLRTs) and the overall leak tight integrity of the containment is
verified by a Type A ILRT, as required by 10 CFR 50, Appendix J,
``Primary Reactor Containment Leakage Testing for Water-Cooled Power
Reactors.'' These tests are performed to verify the essentially leak
tight characteristics of the containment at the design basis
accident pressure. The proposed change for a one-time extension of
the Type A ILRT does not affect the method for Type A, B or C
testing or the test acceptance criteria.
EGC has conducted a risk assessment to determine the impact of a
change to the LGS, Units 1 and 2 Type A ILRT from 10 to 15 years.
This risk assessment measured the impact to the Large Early Release
Frequency, Person-Rem, and Conditional Containment Failure
Frequency. This assessment indicated that the proposed LGS, Units 1
and 2 Type A ILRT interval extension has a very small change in risk
to the public and is an acceptable plant change from a risk
perspective.
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: Mr. Bradley Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
Florida Power and Light Company, Docket Nos. 50-335 and 50-389, St.
Lucie Plant, Units 1 and 2, St. Lucie County, Florida
Date of amendment request: June 4, 2007.
Description of amendment request: The proposed amendment would
remove the Technical Specification (TS) requirements that reference
hydrogen recombiners and hydrogen monitors. The proposed amendment
suggests changes support implementation of the revisions to 10 CFR
50.44, ``Standards for Combustible Gas Control System in Light Water
Cooled Power Reactors,'' that became effective on September 16, 2003.
The changes would be consistent with Revision 1 of the NRC-approved
Technical Specification Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-447, ``Elimination of Hydrogen
Recombiners and Change to Hydrogen and Oxygen Monitors.'' The
particular TS improvement in question was announced in the Federal
Register Notice on September 25, 2003, as part of the consolidated line
item improvement process.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen [and oxygen] monitors are no longer required to mitigate
design-basis accidents and, therefore, the hydrogen monitors do not
meet the definition of a safety-related component as defined in 10
CFR 50.2. RG [Regulatory Guide] 1.97 Category 1, is intended for key
variables that most directly indicate the accomplishment of a safety
function for design-basis accident events. The hydrogen [and oxygen]
monitors no longer meet the definition of Category 1 in RG 1.97. As
part of the rulemaking to revise 10 CFR 50.44 the Commission found
that Category 3, as defined in RG 1.97, is an appropriate
categorization for the hydrogen monitors because the monitors are
required to diagnose the course of beyond design-basis accidents.
[Also, as part of the rulemaking to revise 10 CFR 50.44, the
Commission found that Category 2, as defined in RG 1.97, is an
appropriate categorization for the oxygen monitors, because the
monitors are required to verify the status of the inert
containment.]
The regulatory requirements for the hydrogen [and oxygen]
monitors can be relaxed without degrading the plant emergency
response. The emergency response, in this sense, refers to the
methodologies used in ascertaining the condition of the reactor
core, mitigating the consequences of an accident, assessing and
projecting offsite releases of radioactivity, and establishing
protective action recommendations to be communicated to offsite
authorities. Classification of the hydrogen monitors as Category 3,
[classification of the oxygen monitors as Category 2] and removal of
the hydrogen [and oxygen] monitors from TS will not prevent an
accident management strategy through the use of the SAMGs [severe
accident management guidelines], the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen [and oxygen] monitor
requirements, including removal of these
[[Page 45458]]
requirements from TS, does not involve a significant increase in the
probability or the consequences of any accident previously
evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen [and oxygen] monitor requirements,
including removal of these requirements from TS, will not result in
any failure mode not previously analyzed. The hydrogen recombiner
and hydrogen [and oxygen] monitor equipment was intended to mitigate
a design-basis hydrogen release. The hydrogen recombiner and
hydrogen [and oxygen] monitor equipment are not considered accident
precursors, nor does their existence or elimination have any adverse
impact on the pre-accident state of the reactor core or post
accident confinement of radionuclides within the containment
building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen [and oxygen] monitor requirements,
including removal of these requirements from TS, in light of
existing plant equipment, instrumentation, procedures, and programs
that provide effective mitigation of and recovery from reactor
accidents, results in a neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI [Three-Mile Island],
Unit 2 accident can be adequately met without reliance on safety-
related hydrogen monitors.
[Category 2 oxygen monitors are adequate to verify the status of
an inerted containment.]
Therefore, this change does not involve a significant reduction
in the margin of safety. The intent of the requirements established
as a result of the TMI, Unit 2 accident can be adequately met
without reliance on safety-related oxygen monitors.] Removal of
hydrogen [and oxygen] monitoring from TS will not result in a
significant reduction in their functionality, reliability, and
availability.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Thomas H. Boyce.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment request: June 13, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TS) 5.5.9, ``Ventilation Filter
Testing Program (VFTP),'' to impose lower (i.e., more restrictive)
limits on the maximum pressure drop across the combined high efficiency
particulate air filters and charcoal adsorbers in three safety-related
ventilation systems. These ventilation systems are the Control Room
Emergency Ventilation System, the Engineered Safety Features
Ventilation System, and the Fuel-Handling Area Exhaust Ventilation
System.
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 of occurrence or consequences of an accident
previously evaluated?
Response: No.
The proposed change consists of establishing more restrictive
criteria in the Technical Specification (TS) for the maximum
pressure drop across high efficiency particulate air filters (HEPA)
and charcoal adsorbers in safety-related ventilation systems. These
TS criteria are used to determine the acceptability of periodic test
results. These criteria are not accident initiators. Therefore,
there will be no effect on the probability of an accident. The
safety-related ventilation systems involved in the proposed change
function to mitigate the consequences of accidents. The proposed
change will provide increased assurance that the HEPA filters and
charcoal adsorbers in these systems will be capable of performing
their safety function of reducing the release of radioactive
material resulting from evaluated accidents. Therefore, there will
be no increase in the consequences of those accidents.
Therefore, the proposed change will 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 consists of establishing more restrictive
acceptance criteria for existing TS[-]required tests. The proposed
change does not affect the manner in which the tests are performed.
The proposed change will not result in any new or different methods
or modes of operation of existing structures, systems, or
components. The proposed change will not introduce any new
structures, system, or components.
Therefore, the proposed change will 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 margin of safety associated with the proposed change is the
capability of the applicable safety-related ventilation systems to
prevent radiation exposures from exceeding acceptable limits due to
the release of radioactive material caused by an evaluated accident.
The proposed change will provide increased assurance that the HEPA
filters and charcoal adsorbers in these systems will be capable of
performing this function.
Therefore, the proposed change will not involve a significant
reduction in the margin of safety.
The Nuclear Regulatory Commission (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: Kimberly Harshaw, Esquire, One Cook Place,
Bridgman, MI 49106.
NRC Acting Branch Chief: Travis Tate.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment request: June 27, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TS) Surveillance Requirements 3.8.1.2,
8, 12, 13, 16, and 19, changing the steady state frequency of all
diesel generators (DGs) from the current allowed frequency range of
59.4-61.2 Hz, to 59.4-60.5 Hz (i.e., a decrease of the upper limit,
resulting in narrowing of
[[Page 45459]]
the current range). The licensee stated that the current frequency
range is nonconservative and could result in undesirable effects such
as centrifugal charging pump motor brake horsepower exceeding its
nameplate maximum horsepower, and overloading the DGs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided
its analysis of the issue of no significant hazards consideration. The
NRC staff has performed its own analysis, which is presented below:
(1) Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
Response: No.
The more restrictive steady state frequency range ensures that
the diesel generators and equipment being powered by the diesel
generators will function as designed to mitigate an accident as
described in the Update Final Safety Analysis Report (UFSAR). The
DGs and the equipment they power are part of the systems required to
mitigate accidents; no accident analyzed in the UFSAR is initiated
by mitigation equipment. Therefore, the proposed change to the
allowed frequency range of the DGs will not have any impact on the
probability of an accident previously evaluated. Furthermore, other
than narrowing the allowed frequency range of the DGs, there is no
other design or operational change. Therefore, the proposed change
does not increase the probability of malfunction of the DGs or the
equipment they power.
Narrowing of the DG maximum steady state frequency limit will
ensure that the DGs and equipment powered by the DGs will perform as
originally designed and analyzed to mitigate the consequences of any
accident described in the UFSAR. Therefore, the proposed change does
not involve a significant increase in the consequences of an
accident previously evaluated in the UFSAR.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
There is no design change associated with the proposed
amendment. Making an existing DG requirement more restrictive alone
will not alter plant configuration because no new or different type
of equipment will be installed, and because no methods governing
plant operation will be changed. The proposed change to allowed
frequency range will not have any effect on the assumptions of
accident scenarios previously made in the UFSAR. 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.
Despite the proposed change to the DG maximum steady state
frequency limit, the DGs and equipment powered by the DGs will
continue to perform as originally designed, and originally analyzed
in the UFSAR. There is no associated change to the methods and
assumptions used to analyze DG performance. The proposed change will
maintain the required function of the DGs and the equipment powered
by the DGs to ensure that operation of structures, systems, or
components is as currently set forth in the UFSAR. Therefore, the
proposed change does not involve a significant reduction in the
margin of safety.
The Nuclear Regulatory Commission (NRC) staff has reviewed the
licensee's analysis and, based on its own analysis, it appears that the
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC
staff proposes to determine that the proposed amendment involves no
significant hazards consideration.
Attorney for licensee: Kimberly Harshaw, Esquire, One Cook Place,
Bridgman, MI 49106.
NRC Acting Branch Chief: Travis L. Tate.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: July 9, 2007.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS), an NRC-controlled document,
by moving the Table of Contents (TOC) out of the TS and making the TOC
into a licensee-controlled document.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR) Part 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration (NSHC)
which is reproduced below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
No.
The proposed change is administrative and affects control of a
document, the TOC, listing the specifications in the plant TS.
Transferring control from the NRC to NMC (the licensee) does not
affect the operation, physical configuration, or function of plant
equipment or systems. It does not impact the initiators or
assumptions of analyzed events, nor does it impact the mitigation of
accidents or transient events. The change has no impact on, and
hence cannot increase, the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
No.
The proposed change is administrative and does not alter the
plant configuration, require installation of new equipment, alter
assumptions about previously analyzed accidents, or impact the
operation or function of plant equipment or systems. Therefore, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
No.
The proposed change is administrative. The TOC is not required
by regulation to be in the TS. [Its] removal does not impact any
safety assumptions or have the potential to reduce a margin of
safety as described in the TS Bases. The change involves a transfer
of control of the TOC from the NRC to NMC. No change in the
technical content of the TS [ ] is involved. Consequently, transfer
from the NRC to NMC has no impact on the margin of safety, and hence
cannot involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this analysis, it appears that the three standards of 10 CFR 50.92(c)
are satisfied. Therefore, the NRC staff proposes to determine that the
proposed amendment involves no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Acting Branch Chief: Travis L. Tate.
PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric
Station, Unit 2
(SSES 2), Luzerne County, Pennsylvania.
Date of amendment request: March 2, 2007.
Description of amendment request: The proposed amendment would add
an ACTIONS Note 3 to the SSES 2 Technical Specification 3.8.1, ``AC
Sources--Operating,'' to allow a Unit 1 4160 volt subsystem to be de-
energized and removed from service to perform bus maintenance.
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.
This change does not involve any physical change to structures,
systems, or components (SSCs) and does not alter the method of
operation of any SSCs. The current assumptions in the safety
analysis regarding accident initiators and mitigation of accidents
are unaffected by these changes. No
[[Page 45460]]
SSC failure modes or mechanisms are being introduced, and the
likelihood of previously analyzed failures remains unchanged.
Operation in accordance with the proposed new ACTIONS Note 3 in
Unit 2 Technical Specification 3.8.1 ensures that the AC
[alternating current] distribution system and supported equipment
remain capable of performing their functions as described in the
Final Safety Analysis Report (FSAR). There are no changes to any
accident initiators or to the mitigating capability of safety-
related equipment supported by the Class 1E Electrical AC system.
The protection provided by these safety-related systems will
continue to be provided as assumed by the safety analysis.
Therefore, this proposed change does 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 does not involve a physical alteration of
any plant equipment. No new equipment is being introduced, and
installed equipment is not being operated in a new or different
manner. There are no setpoints, at which protective or mitigative
actions are initiated, affected by this change. This change does not
alter the manner in which equipment operation is initiated, nor will
the function demands on credited equipment be changed. No
alterations in the procedures that ensure the plant remains within
analyzed limits are being proposed, and no changes are being made to
the procedures relied upon to respond to an off-normal event as
described in the FSAR [final safety analysis report]. As such, no
new failure modes are being introduced. The change does not alter
assumptions made in the safety analysis and licensing basis.
Therefore, the proposed change does 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 margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed change is acceptable because the new
ACTIONS Note 3 has been established to be consistent with the
existing completion times for declaring required equipment
inoperable that has no offsite power or DG [diesel generator] power
available. Therefore, the plant response to analyzed events is not
affected by this change and will continue to provide the margin of
safety assumed by the safety analysis.
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: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief: Mark G. Kowal.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: July 26, 2007
Description of amendment request: The proposed amendment would
remove values for turbine first stage pressure equivalent to
Pbypass from the Technical Specifications.
Pbypass is the reactor power level below which the turbine
stop valve closure and the turbine control valve fast closure reactor
protection system trip functions and the end-of-cycle recirculation
pump trip are bypassed automatically.
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 amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed removal of values for turbine first stage pressure
associated with Pbypass from the Technical Specifications
does not alter the requirements for component operability or
surveillance currently in the Technical Specifications. The proposed
change will have no impact on any safety related structures, systems
or components.
The probability of occurrence of a previously evaluated accident
is not increased because this change does not introduce any new
potential accident initiating conditions. The consequences of
accidents previously evaluated in the UFSAR [Updated Final Safety
Analysis Report] are not affected because the ability of the
components to perform their required function is not affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change is administrative in nature, and does not
result in physical alterations or changes in the method by which any
safety related system performs its intended function. The proposed
change does not affect any safety analysis assumptions. The proposed
change does not create any new accident initiators or involve an
activity that could be an initiator of an accident of a different
type.
All components will continue to be tested to the same
requirements as defined in the Technical Specification Surveillance
Requirements. The proposed revision does not make changes in any
method of testing or how any safety related system performs its
safety functions.
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 amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change to remove values for turbine first stage
pressure associated with Pbypass from the Technical
Specifications does not alter the Technical Specification
requirements for reactor protection system operability. The turbine
first stage pressure setpoint will be controlled in accordance with
plant procedures and will be verified during post-installation
testing.
The proposed change will not affect the current Technical
Specification requirements or the components to which they apply.
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: Mr. Jeffrie J. Keenan, Esquire, PSEG
Nuclear--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
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: June 26, 2007.
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TSs) to establish more effective
and appropriate action, surveillance, and administrative requirements
related to ensuring the habitability of the control room envelope (CRE)
in accordance with Nuclear Regulatory Commission (NRC)-approved TS Task
Force (TSTF) Standard Technical Specification change traveler TSTF-448,
Revision 3, ``Control Room Habitability.'' Specifically, the proposed
amendment would modify TS 3.7.7, Control Room Makeup and Cleanup
Filtration System (CRMCFS) and TS Section 6.8, ``Administrative
Controls-Procedures, Programs, and Manuals.'' The NRC staff issued a
``Notice of Availability of Technical Specification Improvement to
[[Page 45461]]
Modify Requirements Regarding Control Room Envelope Habitability Using
the Consolidated Line Item Improvement Process'' associated with TSTF-
448, Revision 3, in the Federal Register on January 17, 2007 (72 FR
2022). The notice included a model safety evaluation, a model no
significant hazards consideration (NSHC) determination, and a model
license amendment request. In its application dated June 26, 2007, the
licensee affirmed the applicability of the model NSHC determination
which is presented below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC adopted by the licensee is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident from any Accident Previously
Evaluated
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice. Therefore, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition. Therefore, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the analysis adopted by the licensee
and, based on this review, it appears that the standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the request for amendments involves NSHC.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Branch Chief: Thomas G. Hiltz.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas.
Date of amendment request: April 10, 2007.
Brief description of amendments: The proposed amendments would
revise Technical Specifications (TS) 3.1, ``Reactivity Control
Systems,'' TS 3.2, ``Power Distribution Limits,'' TS 3.3,
``Instrumentation,'' and TS 5.6.5b, ``Core Operating Limits Report
(COLR).'' The requested change proposes to incorporate standard
Westinghouse-developed and NRC-approved analytical methods into the
lists of methodologies used to establish the core operating limits.
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.
No physical plant changes or changes in manner in which the
plant will be operated as a result of the methodology changes. The
proposed changes do not impact the condition or performance of any
plant structure, system or component. The core operating limits are
established to support Technical Specifications 3.1, 3.2, 3.3, and
3.4. The core operating limits ensure that 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 establish the core operating limits for each operating cycle
are based on methods previously found acceptable by the NRC and
listed in Technical Specifications section 5.6.5.b. Application of
these NRC-approved methods will continue to ensure that acceptable
operating limits are established to protect the fuel cladding
integrity during normal operation and AOOs. The requested Technical
Specification changes, including those changes proposed to conform
with the NRC-approved analysis methodologies, do not involve any
plant modifications or operational changes that could affect system
reliability, performance, or possibility of operator error. The
requested changes do not affect any postulated accident precursors,
does not affect any accident mitigation systems, and does not
introduce any new accident initiation mechanisms.
As a result, the proposed changes to the CPSES [Comanche Peak
Steam Electric Station] Technical Specifications do not involve any
increase in the probability or the consequences of any accident or
malfunction of equipment important to safety previously evaluated
since neither accident probabilities nor consequences are being
affected by this proposed change.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
There are no physical changes being made to the plant. No new
modes of plant operation are being introduced. The parameters
assumed in the analyses are within the design limits of the existing
plant equipment. All plant systems will perform as designed during
the response to a potential accident.
Therefore, the proposed change to the CPSES Technical
Specifications does not create the possibility of a new or different
kind of accident or malfunction of equipment important to safety
from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The NRC-approved accident analysis methodologies include
restrictions on the
[[Page 45462]]
choice of inputs, the degree of conservatism inherent in the
calculations, and specified event acceptance criteria. Analyses
performed in accordance with these methodologies will not result in
adverse effects on the regulated margin of safety. Similarly, the
use of axial power distribution controls based on the relaxed axial
offset control strategy is a time-proven and NRC-approved method.
The method is consistent with the accident analyses assumptions as
described in the list of NRC-approved methodologies proposed to be
used to establish the core operating limits. Finally, the proposed
changes to allow operation with the BEACON [Best Estimate Analyzer
for Core Operation Nuclear] power distribution monitoring tool
provide additional information to the reactor operators on the state
of the reactor core. Again, the use of the BEACON tool and the
methodology used to develop the inputs to the tool are consistent
with and controlled by the NRC-approved methodologies used to
establish the core operating limits. As such, the margin of safety
assumed in the plant safety analysis is not adversely affected by
the proposed changes.
Based on the above evaluations, TXU Power concludes that the
proposed amendment(s) present 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.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: Thomas G. Hiltz.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: May 22, 2007.
Brief description of amendments: The proposed amendment would
revise the Technical Requirements Surveillance (TRS) 13.3.33.2, Cycling
Frequency for the Turbine Stop and Control Valves. The proposed change
would increase the frequency interval for the turbine stop and control
valves testing from 12 to 26 weeks.
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 will increase the frequency interval for
testing the high pressure (HP) and low pressure (LP) turbine stop
and control valves to 26 weeks. This test requires the movement of
the HP and LP turbine valves through one complete cycle once every
26 weeks. The test verifies freedom of movement of the valve
components and is beneficial in early detection of problems with
valve operation. [The test ensures that all turbine steam inlet
valves are capable of closing to protect the turbine from excessive
overspeed, which could generate potentially damaging missiles.]
Siemens, the turbine manufacturer for Comanche Peak Steam
Electric Station (CPSES), has evaluated the change in the
probability of generating external/high-trajectory turbine missiles
resulting from a hypothetical LP turbine disk failure which could
adversely affect safety-related SSCs [structures, systems, and
components] due to the change in the surveillance interval weeks
using a previously approved missile probability analysis
methodology. The results of the analysis show the new valve test
interval of 26 weeks with a turbine inspection interval of 100,000
hours is safe and acceptable as the probability of occurrence of a
turbine missile per turbine year is less than the Nuclear Regulatory
Commission (NRC) limit of 1E-4 per 8760 hours (turbine year) or
11.42E-4 at 100,000 hours (Reference 7.4 [of the licensee's May 22,
2007, application]). Therefore, the risk of the loss of an essential
system from a single event is acceptable. Since the probability of
generating external, high-trajectory turbine missiles resulting from
a hypothetical LP turbine disc failure which could adversely affect
safety related SSCs due to the increased valve test interval from 12
to 26 weeks is less than the NRC limit, it is acceptable to increase
the turbine test interval in TRS 13.3.33.2. The test interval change
would increase overall plant capacity factor and result in a net
improvement in plant safety by reducing the likelihood of plant
trips and stress and wear on plant components. In addition, the
increased test intervals would reduce the likely cause of a plant
transient and unnecessary burden on personnel resources which is
consistent with Generic Letter 93-005 (Reference 7.7 [of the
licensee's May 22, 2007, application]) and NUREG-1366 (Reference 7.2
[of the licensee's May 22, 2007, application]). Based upon Siemens'
analysis and the updated stop and control valves failure
probability, it is concluded that the implementation of this change
in testing frequency will not increase the probability or
consequences of an accident previously evaluated in the UFSAR.
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility or the manner in which
the plant is operated and maintained. The proposed change does not
alter or prevent the ability of structures, systems, and components
(SSCs) from performing their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change does not affect the source term,
containment isolation, or radiological release assumptions used in
evaluating the radiological consequences of an accident previously
evaluated. The proposed change is consistent with safety analysis
assumptions and resultant consequences.
Therefore, the proposed change does 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 will reduce the frequency for testing the
high pressure (HP) and low pressure (LP) turbine stop and control
valves. Turbine overspeed is limited by rapid closure of the turbine
stop and control valves. Turbine overspeed can result in the
occurrence of turbine missiles from a burst type failure of the low
pressure blades or disks. The damage from turbine missiles has been
previously evaluated in the UFSAR [updated final safety analysis
report] (Reference 7.3 [of the licensee's May 22, 2007,
application]). The proposed activity does not introduce the
possibility of a new accident because no new failure modes are
introduced.
Turbine overspeed with the resulting turbine missiles is the
only accident potentially affected by failure of the turbine stop
and control valves. The turbine missile analysis is not altered by
reducing the frequency of high and low pressure stop and control
valve testing. Reducing the frequency of turbine valve testing from
every 12 weeks to every 26 weeks does not result in a significant
change in the failure rate, nor does it affect the failure modes for
the turbine valves.
There are no hardware changes nor are there any changes in the
method by which any safety-related plant system performs its safety
function. This amendment will not affect the normal method of plant
operation or change any operating parameters. No performance
requirements or response time limits will be affected. No new
accident scenarios, transient precursors, failure mechanisms, or
limiting single failures are 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.
Therefore, the proposed change does 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 proposed change does not involve a significant reduction in
a margin of safety since the conclusions of the safety analyses in
the CPSES FSAR [final safety analysis report] (Reference 7.3 [of the
licensee's May 22, 2007, application]) are essentially unchanged and
NRC safety limits are not exceeded.
Therefore the proposed change does not involve a reduction in a
margin of safety.
[[Page 45463]]
Based on the above evaluations, TXU Power concludes that the
proposed amendment(s) present no significant hazards 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.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: Thomas G. Hiltz.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: July 13, 2007.
Description of amendment request: The proposed amendment revises
the Technical Specifications (TSs) requirements related to main control
room and emergency switchgear room envelope habitability. These changes
are consistent with the Nuclear Regulatory Commission (NRC)-approved
Revision 3 of Technical Specification Task Force (TSTF) Standard
Technical Specifications (STS) Change Traveler TSTF-448, ``Control Room
Habitability.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes consist of TS wording, format and
conforming changes to facilitate incorporation of TSTF-448 [72 FR
2022] into the Surry custom TS and for consistency with NUREG-1431,
Revision 3, to the extent practical. The proposed changes are
administrative in nature and, as such, do not impact the condition
or performance of any plant structure, system or component. The
proposed changes do not affect the initiators of any previously
analyzed event or the assumed mitigation of accident or transient
events. As a result, the proposed administrative changes to the
Surry TS do not involve any increase in the probability or the
consequences of any accident or malfunction of equipment important
to safety previously evaluated since neither accident probabilities
or consequences are being affected by the proposed changes.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes are administrative in nature, and therefore
do not involve any changes in station operation or physical
modifications to the plant. In addition, no changes are being made
in the methods used to respond to plant transients that have been
previously analyzed. No changes are being made to plant parameters
within which the plant is normally operated or in the setpoints,
which initiate protective or mitigative actions, and no new failure
modes are being introduced. Therefore, the proposed changes to the
Surry Technical Specifications do not create the possibility of a
new or different kind of accident or malfunction of equipment
important to safety from any previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes consist of TS wording, format and
conforming changes to facilitate incorporation of TSTF-448 into the
Surry custom TS and for consistency with NUREG-1431, Revision 3. The
proposed changes are administrative in nature, and do not impact
station operation or any plant structure, system or component that
is relied upon for accident mitigation. Furthermore, the margin of
safety assumed in the plant safety analysis is not affected in any
way by the proposed changes. Therefore, the proposed administrative
changes to the Surry Technical Specifications do not involve a
reduction in a margin of safety.
The NRC staff has reviewed the 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: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Branch Chief: Evangelos C. Marinos.
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.
Exelon Generation Company, LLC, and PSEG Nuclear LLC,
Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station,
Units 2 and 3, York and Lancaster Counties, Pennsylvania
Date of amendment request: March 6, 2007.
Brief description of amendment request: The proposed amendment
would modify the main steam isolation valve (MSIV) leakage Technical
Specification (TS) Surveillance Requirement (SR) 3.6.1.3.14 to
establish a total leakage rate limit for the sum of the four main steam
lines.
Date of publication of individual notice in Federal Register: July
24, 2007.
Expiration date of individual notice: September 22, 2007.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of application for amendments: June 25, as supplemented July
3, 2007.
Description of amendments request: The proposed amendment would
allow deletion of License Condition 2.(G)2 regarding the performance of
power uprate large transient testing.
Date of publication of individual notice in the Federal Register:
July 13, 2007 (72 FR 38627).
Expiration date of individual notice: August 14, 2007 (Public
comments) and September 11, 2007 (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
[[Page 45464]]
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 51.12(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].
AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of application for amendment: September 28, 2006.
Brief description of amendment: The amendment revises the Oyster
Creek Technical Specification (TS) definition of Channel Calibration,
Channel Check, and Channel Test consistent with NUREG-1433, Revision
3.0, ``Standard Technical Specifications General Electric Plants, BWR/4
Specifications,'' dated June 2004. These definitions apply to all
instrument functions in the TSs, including Reactor Protection System
instruments.
Date of Issuance: July 27, 2007.
Effective date: As of the date of Issuance to be implemented within
60 days.
Amendment No.: 263.
Facility Operating License No. DPR-16: The amendment revised the
TSs.
Date of initial notice in Federal Register: November 21, 2006 (71
FR 67392). The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated July 27, 2007.
No significant hazards consideration comments received: No.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: July 20, 2006, as supplemented
by letter dated May 3, 2007.
Brief description of amendments: The amendments revised Technical
Specifications (TS) 3.1.6, ``Shutdown Control Element Assembly (CEA)
Insertion Limits,'' to modify the TS Limiting Condition for Operation
(LCO) 3.1.6 and Surveillance Requirements (SRs) 3.1.6.1 to require
shutdown CEAs to be withdrawn to >=147.75 inches, instead of the
current limit of >=144.75 inches.
Date of issuance: July 25, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: Unit 1-168, Unit 2-168, Unit 3-168.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Operating License and Technical Specifications.
Date of initial notice in Federal Register: September 26, 2006 (71
FR 56191). The supplement dated May 3, 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 amendment is contained in a Safety Evaluation dated July 25,
2007.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: February 2, 2007.
Brief description of amendment: This amendment deletes the
technical specification (TS) requirements related to containment
hydrogen monitors and supports implementation of the revisions of 10
CFR 50.44, Combustible Gas Control for Nuclear Power Reactors, that
became effective on October 16, 2003. This is a Consolidated Line Item
Improvement Program modification, which adopts TS Task Force (TSTF)
Standard TS Change Traveler, TSTF-447, Elimination of Hydrogen
Recombiners and Change to Hydrogen and Oxygen Monitors.
Date of issuance: July 16, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 216.
Renewed Facility Operating License No. DPR-23: Amendment revises
the technical specifications.
Date of initial notice in Federal Register: April 24, 2007 (72 FR
20378). The Commission's related evaluation of the amendment is
contained in a safety evaluation dated March 21, 2007.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: December 20, 2006.
Brief description of amendment: This amendment revises Technical
Specification (TS) 6.12, ``High Radiation Area.'' The amendment aligns
the requirements contained in the TS with the revised Regulatory Guide
8.38, Revision 1, ``Control of Access to High and Very High Radiation
Areas in Nuclear Power Plants.'' Specifically, the changes include
differentiating dose rates associated with high and very high radiation
areas, adding requirements for groups entering high radiation areas,
and clarifying the communication requirements for workers in high
radiation areas.
Date of issuance: July 23, 2007.
Effective date: This amendment is effective as of the date of
issuance and shall be implemented within 60 days of issuance.
Amendment No.: 125.
Facility Operating License No. NPF-63: Amendment revises the TSs.
Date of initial notice in Federal Register: February 27, 2007 (72
FR 8802). The Commission's related evaluation of the amendment is
contained in a safety evaluation dated July 23, 2007.
No significant hazards consideration comments received: No.
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: July 31, 2006 as supplemented
May 24, 2007.
Brief description of amendments: The amendments revised TS 3.6.3,
``Containment Isolation Valves,'' by removing the allowance to open the
upper containment purge isolation
[[Page 45465]]
valves in the applicable modes of operation when containment integrity
is required by the TSs. In addition, the amendments deleted TS 3.3.6,
``Containment Purge and Exhaust Isolation Instrumentation''. The change
made the TSs requirements consistent for both the upper and lower
containment purge isolation valves.
Date of issuance: July 26, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 243, 224.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: December 5, 2006 (71 FR
70558) The supplement dated May 24, 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.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated July 26, 2007.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: February 10, 2006, as
supplemented by letter dated March 8, 2007.
Brief description of amendment: The changes would clarify technical
specifications (TSs) for the Perry Nuclear Power Plant (PNPP) by
revising the TS action requirements that must be followed when one or
more annulus gas treatment system initiation channels are inoperable.
The clarifying changes will make the PNPP TSs consistent with Nuclear
Regulatory Commission (NRC) staff precedents for containment filtering
safety systems that operate continuously in the protection mode of
operation.
Date of issuance: July 30, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 147.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: May 23, 2006 (71 FR
29678). The March 8, 2007, supplement contained clarifying information
and did not change the NRC staff's initial proposed finding of no
significant hazards consideration. The Commission's related evaluation
of the amendment is contained in a Safety Evaluation dated July 30,
2007.
No significant hazards consideration comments received: No.
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 amendments: January 27, 2006, as
supplemented November 28, 2006, April 30, 2007, and July 17, 2007.
Brief description of amendments: These amendments revise Technical
Specifications (TS) Section 3/4 9.1, ``Boron Concentration,'' Section
3/4 9.14, ``Spent Fuel Storage,'' and Section 3/4 5.5.1, ``Fuel Storage
Criticality'' to allow use of Metamic rack inserts, and administrative
controls that require mixing higher reactivity fuel with lower-
reactivity fuel.
Date of issuance: July 17, 2007.
Effective date: As of the date of issuance and shall be implemented
prior to the end of Unit 4 Cycle 24.
Amendment Nos: 234 and 229.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the TS.
Date of initial notice in Federal Register: May 9, 2006 (71 FR
26999). The supplements dated November 28, 2006, April 30, 2007, and
July 17, 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 on May 9, 2006.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 17, 2007.
No significant hazards consideration comments received: No.
Nuclear Management Company, Docket No. 50-263, Monticello Nuclear
Generating Plant (MNGP), Wright County, Minnesota
Date of application for amendment: January 29, 2007, as
supplemented on June 5, 2007.
Brief description of amendment: The amendment revised Table
3.3.5.1-1 of the Technical Specifications for three low-pressure
coolant injection loop select logic functions. The surveillance of
these three functions was previously required to be performed every 92
days. The amended requirement requires a channel calibration and logic
system functional test, respectively, every 24 months. In addition, the
allowable values associated with these three functions are changed to
match the extended surveillance interval.
Date of issuance: July 20, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No: 151.
Renewed Facility Operating License No. DPR-22: Amendment revised
the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: March 13, 2007 (72 FR
11391).
The supplemental letter dated June 5, 2007, 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 July 20, 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, 2006, as supplemented by
letters dated April 6 and May 31, 2007, and electronic mail dated July
18, 2007.
Brief description of amendments: The amendments revised TSs 3/
4.8.2.1, ``DC [Direct Current] Sources--Operating,'' and 3/4.8.2.2,
``DC Sources--Shutdown,'' and add a new TS 3/4.8.2.3, ``Battery
Parameters.'' The amendments revised allowed outage times for battery
chargers as well as battery charger testing criteria, and relocate a
number of battery surveillance requirements to a licensee-controlled
Battery Monitoring and Maintenance Program. The changes are consistent
with Standard TS Change Traveler TSTF-360, Revision 1, ``DC Electrical
Rewrite.''
Date of issuance: July 20, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment Nos.: Unit 1-180; Unit 2-167.
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: September 12, 2006 (71
FR 53721). The supplemental letters dated April 6 and May 31, 2007, and
electronic mail dated July 18, 2007, provided additional information
that
[[Page 45466]]
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 July 20, 2007.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 2nd day of August 2007.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E7-15459 Filed 8-13-07; 8:45 am]
BILLING CODE 7590-01-P