[Federal Register Volume 71, Number 243 (Tuesday, December 19, 2006)]
[Notices]
[Pages 75987-76003]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-21346]
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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 November 22, 2006 to December 7, 2006. The
last
[[Page 75988]]
biweekly notice was published on December 5, 2006 (71 FR 70553).
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.
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
[[Page 75989]]
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].
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: September 15, 2006.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 6.8.5, ``Reactor Building
Leakage Rate Testing Program,'' to allow a one-time deferral of the
next Type A, containment integrated leak rate test (ILRT) from ``no
later than September 2008'' to ``prior to startup from T1R18 refueling
outage.'' The NRC has previously approved a one-time 5-year extension
to the Type A ILRT schedule for TMI-1 by issuance of Amendment No. 244,
dated August 14, 2003. Amendment No. 244 changed the TSs to state that
the Type A ILRT shall be performed no later than September 2008. The
proposed amendment would add approximately 15 months to the currently-
approved 15-year interval. This deferral would allow the Type A ILRT to
be performed during a steam generator replacement in the fall of 2009.
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:
Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change will revise TS 6.8.5 to reflect a one-time
extension to the Three Mile Island, Unit 1 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 ``no later than
September 2008'' to ``prior to startup from the T1R18 refueling
outage,'' which is currently scheduled for Fall 2009. The current
Type A ILRT interval of 15 years, based on past performance, would
be extended on a one-time basis by approximately 15 months.
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 TMI, Unit 1 containment will not exceed allowable leakage rate
values specified in the TS and will continue to perform its design
function following an accident. The risk assessment of the proposed
change has concluded that there is an insignificant increase in
postulated total population dose rate and an insignificant increase
in the postulated conditional containment failure probability.
Additionally, containment inspections have also been performed which
demonstrate the continued structural integrity of the primary
containment.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
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 ILRT
for TMI, Unit 1 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.
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 [Part] 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.
AmerGen has conducted a risk assessment to determine the impact
of a change to the TMI, Unit 1 Type A ILRT schedule from a baseline
ILRT frequency of three times in 10 years to once in 15 years plus
15 months for the risk measures of Large Early Release Frequency
(i.e., LERF), Population Dose, and Conditional Containment Failure
Probability (i.e., CCFP). This assessment indicated that the
proposed TMI, Unit 1 ILRT interval extension has a 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.
[[Page 75990]]
Attorney for licensee: Mr. Brad Fewell, Assistant General Counsel,
Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA
19348.
NRC Branch Chief: Harold K. Chernoff.
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: May 30, 2006, as supplemented by letter
dated November 20, 2006.
Description of amendment request: The proposed amendment would
modify the Technical Specification (TS) requirements related to steam
generator tube integrity. The amendment would adopt Nuclear Regulatory
Commission (NRC)-approved Revision 4 to Technical Specification Task
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
449, ``Steam Generator Tube Integrity.''
The NRC staff published a notice of opportunity for comment in the
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments
adopting TSTF-449, including a model safety evaluation and model no
significant hazards consideration (NSHC) determination, using the
consolidated line item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on May 6, 2005
(70 FR 24126). The licensee affirmed the applicability of the following
NSHC determination in its application dated May 30, 2006.
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 proposed change requires a SG [Steam Generator] Program that
includes performance criteria that will provide reasonable assurance
that the SG tubing will retain integrity over the full range of
operating conditions (including startup, operation in the power
range, hot standby, cooldown and all anticipated transients included
in the design specification). The SG performance criteria are based
on tube structural integrity, accident induced leakage, and
operational LEAKAGE.
A Steam Generator Tube Rupture (SGTR) event is one of the design
basis accidents that are analyzed as part of a plant's licensing
basis. In the analysis of a SGTR event, a bounding primary to
secondary LEAKAGE rate equal to the operational LEAKAGE rate limits
in the licensing basis plus the LEAKAGE rate associated with a
double-ended rupture of a single tube is assumed.
For other design basis accidents such as Main Steam Line Break
(MSLB), rod ejection, and reactor coolant pump locked rotor the
tubes are assumed to retain their structural integrity (i.e., they
are assumed not to rupture). These analyses typically assume that
primary to secondary LEAKAGE for all SGs is 1 gallon per minute or
increases to 1 gallon per minute as a result of accident induced
stresses. The accident induced leakage criterion introduced by the
proposed changes accounts for tubes that may leak during design
basis accidents. The accident induced leakage criterion limits this
leakage to no more than the value assumed in the accident analysis.
The SG performance criteria proposed change to the TSs
identifies the standards against which tube integrity is to be
measured. Meeting the performance criteria provides reasonable
assurance that the SG tubing will remain capable of fulfilling its
specific safety function of maintaining reactor coolant pressure
boundary integrity throughout each operating cycle and in the
unlikely event of a design basis accident. The performance criteria
are only a part of the SG Program required by the proposed change to
the TSs. The program, defined by NEI 97-06, Steam Generator Program
Guidelines, includes a framework that incorporates a balance of
prevention, inspection, evaluation, repair, and leakage monitoring.
The proposed changes do not, therefore, significantly increase the
probability of an accident previously evaluated.
The consequences of design-basis accidents are, in part,
functions of the DOSE EQUIVALENT I-131 in the primary coolant and
the primary to secondary LEAKAGE rates resulting from an accident.
Therefore, limits are included in the plant technical specifications
for operational leakage and for DOSE EQUIVALENT I-131 in primary
coolant to ensure the plant is operated within its analyzed
condition. The typical analysis of the limiting design basis
accident assumes that primary to secondary leak rate after the
accident is 1 gallon per minute with no more than 500 gallons per
day in any one SG, and that the reactor coolant activity levels of
DOSE EQUIVALENT I-131 are at the TS values before the accident.
The proposed change does not affect the design of the SGs, their
method of operation, or primary coolant chemistry controls. The
proposed approach updates the current TSs and enhances the
requirements for SG inspections. The proposed change does not
adversely impact any other previously evaluated design basis
accident and is an improvement over the current TSs.
Therefore, the proposed change does not affect the consequences
of a SGTR accident and the probability of such an accident is
reduced. In addition, the proposed change does not affect the
consequences of an MSLB, rod ejection, or a reactor coolant pump
locked rotor event, or other previously evaluated accident.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed performance based requirements are an improvement
over the requirements imposed by the current technical
specifications. Implementation of the proposed SG Program will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
result of the implementation of the SG Program will be an
enhancement of SG tube performance. Primary to secondary LEAKAGE
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the SG Program to manage SG tube
inspection, assessment, repair, and plugging. The requirements
established by the SG Program are consistent with those in the
applicable design codes and standards and are an improvement over
the requirements in the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TSs.
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.
[[Page 75991]]
NRC Branch Chief (Acting): Douglas V. Pickett.
Carolina Power & Light Company, Docket No. 50-400, Shearon Harris
Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina
Date of amendment request: May 23, 2006, as supplemented by letter
dated October 3, 2006.
Description of amendment request: The proposed amendment would
modify the Technical Specification (TS) requirements related to steam
generator tube integrity. The amendment would adopt Nuclear Regulatory
Commission (NRC)-approved Revision 4 to Technical Specification Task
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
449, ``Steam Generator Tube Integrity.''
The NRC staff published a notice of opportunity for comment in the
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments
adopting TSTF-449, including a model safety evaluation and model no
significant hazards consideration (NSHC) determination, using the
consolidated line item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on May 6, 2005
(70 FR 24126). The licensee affirmed the applicability of the following
NSHC determination in its application dated May 23, 2006.
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 proposed change requires a SG [Steam Generator] Program that
includes performance criteria that will provide reasonable assurance
that the SG tubing will retain integrity over the full range of
operating conditions (including startup, operation in the power
range, hot standby, cooldown and all anticipated transients included
in the design specification). The SG performance criteria are based
on tube structural integrity, accident induced leakage, and
operational LEAKAGE.
A Steam Generator Tube Rupture (SGTR) event is one of the design
basis accidents that are analyzed as part of a plant's licensing
basis. In the analysis of a SGTR event, a bounding primary to
secondary LEAKAGE rate equal to the operational LEAKAGE rate limits
in the licensing basis plus the LEAKAGE rate associated with a
double-ended rupture of a single tube is assumed.
For other design basis accidents such as Main Steam Line Break
(MSLB), rod ejection, and reactor coolant pump locked rotor the
tubes are assumed to retain their structural integrity (i.e., they
are assumed not to rupture). These analyses typically assume that
primary to secondary LEAKAGE for all SGs is 1 gallon per minute or
increases to 1 gallon per minute as a result of accident induced
stresses. The accident induced leakage criterion introduced by the
proposed changes accounts for tubes that may leak during design
basis accidents. The accident induced leakage criterion limits this
leakage to no more than the value assumed in the accident analysis.
The SG performance criteria proposed change to the TSs
identifies the standards against which tube integrity is to be
measured. Meeting the performance criteria provides reasonable
assurance that the SG tubing will remain capable of fulfilling its
specific safety function of maintaining reactor coolant pressure
boundary integrity throughout each operating cycle and in the
unlikely event of a design basis accident. The performance criteria
are only a part of the SG Program required by the proposed change to
the TSs. The program, defined by NEI 97-06, Steam Generator Program
Guidelines, includes a framework that incorporates a balance of
prevention, inspection, evaluation, repair, and leakage monitoring.
The proposed changes do not, therefore, significantly increase the
probability of an accident previously evaluated.
The consequences of design-basis accidents are, in part,
functions of the DOSE EQUIVALENT I-131 in the primary coolant and
the primary to secondary LEAKAGE rates resulting from an accident.
Therefore, limits are included in the plant technical specifications
for operational leakage and for DOSE EQUIVALENT I-131 in primary
coolant to ensure the plant is operated within its analyzed
condition. The typical analysis of the limiting design basis
accident assumes that primary to secondary leak rate after the
accident is 1 gallon per minute with no more than 500 gallons per
day in any one SG, and that the reactor coolant activity levels of
DOSE EQUIVALENT I-131 are at the TS values before the accident.
The proposed change does not affect the design of the SGs, their
method of operation, or primary coolant chemistry controls. The
proposed approach updates the current TSs and enhances the
requirements for SG inspections. The proposed change does not
adversely impact any other previously evaluated design basis
accident and is an improvement over the current TSs.
Therefore, the proposed change does not affect the consequences
of a SGTR accident and the probability of such an accident is
reduced. In addition, the proposed change does not affect the
consequences of an MSLB, rod ejection, or a reactor coolant pump
locked rotor event, or other previously evaluated accident.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed performance based requirements are an improvement
over the requirements imposed by the current technical
specifications. Implementation of the proposed SG Program will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
result of the implementation of the SG Program will be an
enhancement of SG tube performance. Primary to secondary LEAKAGE
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the SG Program to manage SG tube
inspection, assessment, repair, and plugging. The requirements
established by the SG Program are consistent with those in the
applicable design codes and standards and are an improvement over
the requirements in the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TSs.
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 (Acting): Douglas V. Pickett.
[[Page 75992]]
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, (HBRSEP) Unit No. 2, Darlington County, South Carolina
Date of amendment request: June 1, 2006, as supplemented by letter
dated November 20, 2006.
Description of amendment request: The proposed amendment would
revise the surveillance requirements (SR) for the emergency core
cooling system suction inlet in the containment as specified in
Technical Specification SR 3.5.2.6.
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 change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed surveillance change will continue to ensure
that the emergency core cooling system (ECCS) containment sump inlet
is inspected in a manner that will verify operability. Performance
of the required system surveillances, in conjunction with the
applicable operational and design requirements for the ECCS, provide
assurance that the system will be capable of performing the required
design functions for accident mitigation and that the system will
perform in accordance with the functional requirements for the
system as described in the Updated Final Safety Analysis Report for
HBRSEP, Unit No. 2. The proposed rewording of the surveillance
requirement will continue to ensure that the ECCS containment sump
suction inlet is not restricted by debris and suction inlet
strainers show no evidence of structural distress or abnormal
corrosion for HBRSEP, Unit No. 2. This ensures that the rate of
occurrence and consequences of analyzed accidents will not change.
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 previously evaluated?
No. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated. HBRSEP,
Unit No. 2, is replacing the existing ECCS containment sump inlet
trash racks and screens with new strainers in accordance with the
response to Generic Letter 2004-02. The strainer is a passive
component in the ECCS, which is a standby safety system used for
accident mitigation. As such, the strainer cannot be an accident
initiator. A change to Technical Specifications Surveillance
Requirement 3.5.2.6 is needed to accommodate the change to the ECCS
containment sump inlet design. This change does not alter the nature
of events postulated in the HBRSEP, Unit No. 2, Updated Final Safety
Analysis Report, nor does it introduce any unique precursor
mechanisms. Therefore, this change does 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 change does not involve a significant reduction
in the margin of safety. The proposed change to the ECCS containment
sump inlet surveillance requirement provides appropriate and
applicable surveillance for this system. The proposed change to this
surveillance requirement for the ECCS system will continue to ensure
system operability. The proposed change does not adversely affect
any plant safety limits, setpoints, or design parameters. The change
also does not adversely affect the fuel, fuel cladding, Reactor
Coolant System (RCS), or containment integrity. Therefore, this
change does not affect any margin of safety for HBRSEP, 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 (Acting): Douglas Pickett.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of amendment request: May 31, 2006.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) requirements related to steam
generator (SG) tube integrity. In particular, Dominion Nuclear
Connecticut, Inc. (DNC) is proposing to replace the existing SG tube
surveillance program with the NRC-approved Technical Specifications
Task Force (TSTF) 449, Revision 4. The proposed changes are consistent
with the Consolidated Line Item Improvement Process (CLIIP) provided in
the May 6, 2005, Federal Register notice (70 FR 24126). In addition,
the Millstone Power Station, Unit No. 2 (MPS2) TSs are revised beyond
the scope of the CLIIP to provide consistent terminology and format.
Basis for proposed no significant hazards consideration determination:
DNC proposed minor variations and/or deviations from the TS changes
described in the CLIIP beyond the scope of the no significant hazards
consideration determination published on March 2, 2005. DNC has
evaluated the proposed beyond-scope TS changes and determined it does
not represent a significant hazards consideration. As required by 10
CFR 50.91(a), DNC has provided its analysis of the issue of no
significant hazards consideration. The NRC staff has reviewed the
licensee's analysis against the standards of 10 CFR 50.92(c). The NRC
staff's review is presented below.
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes do not affect initiators of previously
analyzed events or assumed mitigation of accident or transient events.
Therefore, these changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes involve adding a new definition and rewording
the existing TS to be consistent with NUREG-1432, Revision 3. In
addition, the requested change for MPS2 incorporates a more
conservative leakage limit of 75 gallons per day per steam generator as
opposed to the CLIIP specified limit of 150 gallons per day per steam
generator. The changes will not impose any requirements or eliminate
any existing requirements that will create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety. Since the
proposed changes do not have an impact on any safety analysis
assumptions and accidents previously evaluated, there are no margin of
safety issues involved.
Therefore, the changes do not involve a significant reduction in a
margin of safety.
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, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Branch Chief: Harold K. Chernoff.
[[Page 75993]]
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit No. 3, New London County, Connecticut
Date of amendment request: May 31, 2006.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) requirements related to steam
generator (SG) tube integrity. In particular, Dominion Nuclear
Connecticut, Inc. (DNC) is proposing to replace the existing SG tube
surveillance program with the NRC-approved Technical Specifications
Task Force (TSTF) 449, Revision 4. The proposed changes are consistent
with the Consolidated Line Item Improvement Process (CLIIP) provided in
the May 6, 2005, Federal Register notice (70 FR 24126). In addition,
the Millstone Power Station, Unit No. 3 (MPS3) TSs are revised beyond
the scope of the CLIIP to provide consistent terminology and format.
Basis for proposed no significant hazards consideration
determination: DNC proposed minor variations and/or deviations from the
TS changes described in the CLIIP beyond the scope of the no
significant hazards consideration determination published on March 2,
2005. DNC has evaluated the proposed beyond-scope TS changes and
determined it does not represent a significant hazards consideration.
As required by 10 CFR 50.91(a), DNC has provided its analysis of the
issue of no significant hazards consideration to support this
conclusion. The NRC staff has reviewed the licensee's analysis against
the standards of 10 CFR 50.92(c). The NRC staff's review is presented
below.
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes involve rewording the existing technical
specifications to be consistent with NUREG-1431, Revision 3. These
proposed changes do not affect initiators of previously analyzed events
or assumed mitigation of accident or transient events.
Therefore, these changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
These proposed changes do not involve physical alteration of the
plant (no new or different type of equipment will be installed). The
changes will not impose any requirements or eliminate any existing
requirements that will create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
Since the proposed changes do not have an impact on any safety
analysis assumptions and accidents previously evaluated, there are no
margin of safety issues involved.
Therefore, the changes do not involve a significant reduction in a
margin of safety.
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, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Branch Chief: Harold K. Chernoff.
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: November 13, 2006.
Description of amendment request: The proposed amendment would
revise Grand Gulf Nuclear Station, Unit 1, Technical Specification (TS)
Limiting Condition for Operation (LCO) 3.10.1, and the associated TS
Bases, to expand its scope to include provisions for temperature
excursions greater than 200 [deg]F as a consequence of inservice leak
and hydrostatic testing, and as a consequence of scram time testing
initiated in conjunction with an inservice leak or hydrostatic test,
while considering operational conditions to be in MODE 4.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 21, 2006 (71 FR 48561), on possible
amendments to revise the plant-specific TS, to expand the scope of TS
LCO 3.10.1, to include provisions for temperature excursions greater
than 200 [deg]F as a consequence of inservice leak and hydrostatic
testing, and as a consequence of scram time testing initiated in
conjunction with an inservice leak or hydrostatic test, while
considering operational conditions to be in MODE 4, including a model
safety evaluation and model No Significant Hazards Determination
(NSHC), using the consolidated line item improvement process. The NRC
staff subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on October 27, 2006 (71 FR 63050). The licensee affirmed the
applicability of the model NSHC determination in its application dated
November 13, 2006.
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
Technical Specifications currently allow for operation at
greater than [200] [deg]F while imposing MODE 4 requirements in
addition to the secondary containment requirements required to be
met. Extending the activities that can apply this allowance will not
adversely impact 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.
Criterion 2: The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Accident Previously
Evaluated
Technical Specifications currently allow for operation at
greater than [200] [deg]F while imposing MODE 4 requirements in
addition to the secondary containment requirements required to be
met. No new operational conditions beyond those currently allowed by
LCO 3.10.1 are introduced. The changes do not involve a physical
alteration of the plant (i.e., no new or different type of equipment
will be installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements or eliminate any existing requirements. The
changes do not alter assumptions made in the safety analysis. The
proposed changes are consistent with the safety analysis assumptions
and current plant operating practice. Therefore, the proposed 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 a Margin of Safety
Technical Specifications currently allow for operation at
greater than [200] [deg]F while imposing MODE 4 requirements in
addition to the secondary containment requirements required to be
met. Extending the activities that can apply this allowance will not
adversely impact any margin of safety. Allowing completion of
inspections and testing and supporting completion of scram time
testing initiated in conjunction with an inservice leak or
hydrostatic test prior to power operation results in enhanced safe
operations by eliminating unnecessary maneuvers to control reactor
temperature and pressure. Therefore, the proposed change
[[Page 75994]]
does not involve a significant reduction in a margin of safety.
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: David Terao.
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 application for amendments: July 14, 2006.
Description of amendment request: The proposed changes would modify
Technical Specification (TS) requirements related to required end
states for TS action statements. The changes are generally consistent
with the NRC-approved Revision 0 to Technical Specification Task Force
(TSTF) Change Traveler, TSTF-423, ``Risk Informed Modification to
Selected Required Action End States for BWR Plants.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on December 14, 2005 (70 FR 74037), on possible
amendments adopting TSTF-423, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on March 23, 2006 (71 FR 14726).
The licensee affirmed the applicability of the following TSTF-423
model NSHC determination in its application dated July 14, 2006.
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 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
[General Electric] NEDC-32988, Revision 2, ``Technical Justification
to Support Risk Informed Modification to Selected Required Action
End States for BWR [boiling-water reactor] 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 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 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.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a 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 Owner's
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.
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: Mr. Brad Fewell, Assistant General Counsel,
Exelon Generation Company LLC, 200 Exelon Way, Kennett Square, PA
19348.
NRC Branch Chief: Harold K. Chernoff.
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 application for amendments: September 15, 2006.
Description of amendment request: The proposed amendment would
revise the required frequency for control rod scram time testing, as
described in Technical Specification (TS) Surveillance Requirement
3.1.4.2, from ``120 days cumulative operation in MODE 1'' to ``200 days
cumulative operation in MODE 1.'' The proposed TS change is based on
the NRC-approved Revision 0 to Technical Specification Task Force
(TSTF) Change Traveler, TSTF-460, ``Control Rod Scram Time Testing
Frequency.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on May 27, 2004 (69 FR 30339), on possible amendments
adopting TSTF-460, including a model safety evaluation and model no
significant hazards consideration (NSHC) determination, using the
consolidated line item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on August 23,
2004 (69 FR 51864).
The licensee affirmed the applicability of the following TSTF-460
[[Page 75995]]
model NSHC determination in its application dated September 15, 2006.
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. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change extends the frequency for testing control
rod scram time testing from every 120 days of cumulative Mode 1
operation to 200 days of cumulative Mode 1 operation. The frequency
of surveillance testing is not an initiator of any accident
previously evaluated. The frequency of surveillance testing does not
affect the ability to mitigate any accident previously evaluated, as
the tested component is still required to be operable. Therefore,
the proposed change does 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 previously evaluated?
Response: No.
The proposed change extends the frequency for testing control
rod scram time testing from every 120 days of cumulative Mode 1
operation to 200 days of cumulative Mode 1 operation. The proposed
change does not result in any new or different modes of plant
operation. 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 extends the frequency for testing control
rod scram time testing from every 120 days of cumulative Mode 1
operation to 200 days of cumulative Mode 1 operation. The proposed
change continues to test the control rod scram time to ensure the
assumptions in the safety analysis are protected. Therefore, the
proposed change does not involve a significant 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 NSHC.
Attorney for Licensee: Mr. Brad Fewell, Assistant General Counsel,
Exelon Generation Company LLC, 200 Exelon Way, Kennett Square, PA
19348.
NRC Branch Chief: Harold K. Chernoff.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: November 14, 2006.
Description of amendment request: The proposed amendment would
revise Specification 3.3.5.1-1 of the Technical Specifications (TSs) to
permit a one-time extension of the quarterly surveillance interval
(i.e., from 92 days to 140 days) for three low pressure coolant
injection (LPCI) loop select logic functions.
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 presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This amendment requests a one-time extension to the performance
interval for a limited number of TS surveillance requirements. The
performance of these surveillances, or the failure to perform, is
not a precursor and does not affect the probability of an accident.
Therefore, the delay in performance proposed in this amendment
request for these surveillance requirements does not increase the
probability of an accident previously evaluated.
A delay in performing these surveillances does not result in a
system being unable to perform its required function. In the case of
this one-time extension, the relatively short period of additional
time period for the systems and components to be in service prior to
the next performance of the surveillance will not affect the ability
of those systems to operate as designed. Therefore, the systems
required to mitigate accidents will remain capable of performing
their required function. No new failure modes have been introduced
because of this action and the consequences remain consistent with
previously evaluated accidents. Therefore, the proposed delay in
performance of the surveillance requirements in this amendment
request does not involve a significant increase in the consequences
of an accident.
Therefore, operation of the facility in accordance with the
proposed license amendment would 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 or accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not involve a physical alteration of
any system, structure, or component (SSC) or a change in the way any
SSC is operated. The proposed amendment does not involve operation
of any SSCs in a manner or configuration different from those
previously recognized or evaluated. No new failure mechanisms will
be introduced by the one-time surveillance requirement deferrals
being requested.
Thus, the proposed amendment 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?
Response: No.
The proposed amendment is a one-time extension of the
performance interval of a limited number of TS surveillance
requirements. Extending these surveillance requirements does not
involve a modification of any TS Limiting Condition for Operation.
Extending these surveillance requirements does not involve a change
to any limit on accident consequences specified in the license or
regulations. Extending these surveillance requirements does not
involve a change to how accidents are mitigated or a significant
increase in the consequences of an accident. Extending these
surveillance requirements does not involve a change in a methodology
used to evaluate consequences of an accident. Extending these
surveillance requirements does not involve a change in any operating
procedure or process.
The instrumentation and components involved in this request have
exhibited reliable operation based on the results of their
performance during past periodic ECCS [emergency core cooling
system] functional testing.
Based on the limited additional period of time that the systems
and components will be in service before the surveillances are next
performed, as well as the operating experience that these
surveillances are typically successful when performed, it is
reasonable to conclude that the margins of safety associated with
these surveillance requirements will not be affected by the
requested extension.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The U.S. 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 request 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 Branch Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: November 6, 2006.
[[Page 75996]]
Description of amendment request: The proposed amendment would add
the realistic large break loss-of-coolant accident (RLBLOCA)
methodology to the analytical methods referenced in Technical
Specification (TS) 5.6.5.b.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (CFR) Section 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 license amendment adds approved analytical methods
used to determine the core operating limits per Technical
Specification 5.6.5.b. Accidents previously evaluated will be
unaffected because they will continue to be analyzed using
applicable methodologies approved by the Nuclear Regulatory
Commission to ensure all required safety limits are met. The
proposed amendment does not affect the acceptance criteria for any
Final Safety Analysis Report (FSAR) safety analysis analyzed
accidents and anticipated operational occurrences. As such, the
proposed amendment does not increase the probability or consequences
of an accident. The proposed amendment does not involve operation of
the required structures, systems or components (SSCs) in a manner or
configuration different from those previously recognized or
evaluated.
Therefore, operation of the facility in accordance with the
proposed amendment would 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 amendment does not involve a physical alteration of
any SSC or a change in the way any SSC is operated. The proposed
amendment does not involve operation of any required SSCs in a
manner or configuration different from those previously recognized
or evaluated. No new failure mechanisms will be introduced by the
changes being requested.
Therefore, the proposed amendment 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?
Response: No.
The proposed amendment does not, by itself, introduce a failure
mechanism. The proposed amendment does not involve any physical
changes to the plant or manner in which the plant is operated. The
proposed changes do not affect the acceptance criteria for any FSAR
safety analysis analyzed accidents or anticipated operational
occurrences. All required safety limits would continue to be
analyzed using methodologies approved by the Nuclear Regulatory
Commission.
Therefore, the proposed amendment would 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: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Branch Chief: L. Raghavan.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: November 13, 2006.
Description of amendment request: The proposed amendment would
relocate the requirements of Technical Specification (TS) 2.22, ``Toxic
Gas Monitors,'' and TS Table 3-3, Item 29 to the Fort Calhoun Station,
Unit No. 1, Updated Safety Analysis Report (USAR).
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 relocates requirements for toxic gas
monitors that do not meet the criteria for inclusion in the TS set
forth in 10 CFR 50.36(c)(2)(ii). The requirements for toxic gas
monitors are being relocated from [the] TS to the USAR, which will
be maintained pursuant to 10 CFR 50.59, thereby reducing the level
of regulatory control. The level of regulatory control has no impact
on the probability or consequences of an accident previously
evaluated. Therefore, the change does not involve a significant
increase in the probability or consequences of any 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 relocates requirements for toxic gas
monitors that do not meet the criteria for inclusion in [the] TS set
forth in 10 CFR 50.36(c)(2)(ii). The change does not involve a
physical alteration of the plant (no new or different type of
equipment will be installed) or make changes in the methods
governing normal plant operation. The change will not impose
different requirements, and adequate control of information will be
maintained. This change will not alter assumptions made in the
safety analysis and licensing basis. Therefore, the change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Does th[e] [proposed] change involve a significant reduction
in a margin of safety?
Response: No.
The proposed change relocates requirements for toxic gas
monitors that do not meet the criteria for inclusion in [the] TS set
forth in 10 CFR 50.36(c)(2)(ii). The change will not reduce a margin
of safety since the location of a requirement has no impact on any
safety analysis assumptions. In addition, the relocated requirements
for toxic gas monitors remain the same as the existing TS. Since any
future changes to these requirements or the surveillance procedures
will be evaluated per the requirements of 10 CFR 50.59, there will
be no reduction in a margin of safety. [Therefore, the TS change
does 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: James R. Curtiss, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: May 31, 2006.
Description of amendment request: The proposed amendments would
correct administrative errors in the SSES 1 and 2 Technical
Specifications (TSs) by adding a logical connector in Condition B of
Limiting Condition for Operation (LCO) 3.8.1 (SSES 1 TS only) and
correct the routing of Interstate 80 (I-80) on Figure 4.1-2 in the SSES
1 and 2 TSs Section 4.0.
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?
[[Page 75997]]
Response: No.
Change to Technical Specification 3.8.1
The proposed change is administrative in nature and does not
impact any accident initiators or analyzed events or assumed
mitigation of accident or transient events. They do not involve the
addition or removal of any equipment, or any design changes to the
facility. Therefore, this proposed change does not represent a
significant increase in the probability or consequences of an
accident previously evaluated.
Change to Technical Specification Figure 4.1-2
The proposed change is administrative in nature and does not
impact any accident initiators or analyzed events or assumed
mitigation of accident or transient events. It does not involve the
addition or removal of any equipment or any design changes to the
facility.
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.
Change to Technical Specification 3.8.1
The proposed change is an administrative change and does not
involve a modification to the physical configuration of the plant
(i.e., no new equipment will be installed) or change in the methods
governing normal plant operation. The proposed change will not
impose any new or different requirements or introduce a new accident
initiator, accident precursor, or malfunction mechanism.
Additionally, there is no change in the types or increases in the
amounts of any effluent that may be released off-site, and there is
no increase in individual or cumulative occupational exposure.
Therefore, this proposed change does not create the possibility of
an accident of a different kind than previously evaluated.
Change to Technical Specification Figure 4.1-2
The proposed change is an administrative change and will not
impose any new or different requirements or introduce a new accident
initiator, accident precursor, or malfunction mechanism.
Additionally, there is no change in the types or increases in the
amounts of any effluent that may be released off-site, and there is
no increase in individual or cumulative occupational exposure.
Therefore, this proposed 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.
Change to Technical Specification 3.8.1
The proposed change revises Condition B in LCO 3.8.1 to be
consistent with Technical Specification 1.2, ``Logical Connectors.''
This change is administrative in nature. Therefore, this proposed
change does not involve a significant reduction in a margin of
safety.
Change to Technical Specification Figure 4.1-2
The proposed change is administrative in nature and does not
affect any plant systems.
Therefore, this 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: Richard J. Laufer.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: September 7, 2006.
Description of amendment request: The proposed amendments would
revise the SSES 1 and 2 Technical Specification (TSs) Section 5.5.6,
``Inservice Testing Program,'' and TS 5.5.12, ``Primary Containment
Leakage Rate Testing Program,'' to be consistent with the requirements
of Title 10 of the Code of Federal Regulations (10 CFR) Section
50.55a(f)(4) and 10 CFR 50.55a(g)(4), respectively. The proposed
amendments would implement TS Task Force (TSTF) 343, Revision 1 and
TSTF 479, Revision 0.
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.
Change to Technical Specification 5.5.6
The proposed change revises the Inservice Testing Program for
consistency with the requirements of 10 CFR 50.55a(f)(4) for pumps
and valves which are classified as American Society of Mechanical
Engineers (ASME) Code Class 1, Class 2 and Class 3.
The proposed change does not impact any accident initiators or
analyzed events or assumed mitigation of accident or transient
events. It does not involve the addition or removal of any
equipment, or any design changes to the facility. Therefore, this
proposed change does not represent a significant increase in the
probability or consequences of an accident previously evaluated.
Change to Technical Specification 5.5.12
The proposed change revises the TS administrative controls
programs for consistency with the requirements of 10 CFR [Part] 50,
paragraph 55a (g)(4) for components classified as Code Class CC.
The proposed change affects the frequency of visual examinations
that will be performed for the concrete surfaces of the containment
for the purpose of the Primary Containment Leakage Rate Testing
Program. In addition, the proposed change allows those examinations
to be performed during power operation as opposed to during a
refueling outage. The frequency of visual examinations of the
concrete surfaces of the containment and the mode of operation
during which those examinations are performed has no relationship to
or adverse impact on the probability of any of the initiating events
assumed in the accident analyses. The proposed change would allow
visual examinations that are performed pursuant to NRC approved ASME
Section XI Code requirement (except where relief has been granted by
the NRC) to meet the intent of visual examinations required by
Regulatory Guide 1.163, without requiring additional visual
examinations pursuant to the Regulatory Guide. The intent of early
detection of deterioration will continue to be met by the more
rigorous requirements of the Code required visual examinations. As
such, the safety function of the containment as a fission product
barrier is maintained.
The proposed change does not impact any accident initiators or
analyzed events or assumed mitigation of accident or transient
events. It does not involve the addition or removal of any
equipment, or any design changes to the facility.
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.
Change to Technical Specification 5.5.6
The proposed change revises the Inservice Testing Program for
consistency with the requirements of 10 CFR 50.55a(f)(4) for pumps
and valves which are classified as American Society of Mechanical
Engineers (ASME) Code Class 1, Class 2 and Class 3.
The proposed change does not involve a modification to the
physical configuration of the plant (i.e., no new equipment will be
installed) or change in the methods governing normal plant
operation. The proposed change will not impose any new or different
requirements or introduce a new accident initiator, accident
precursor, or malfunction mechanism. Additionally, there is no
change in the types or increases in the amounts of any effluent that
may be released off-site and there is no increase in individual or
cumulative occupational exposure. Therefore, this proposed change
does not
[[Page 75998]]
create the possibility of an accident of a different kind than
previously evaluated.
Change to Technical Specification 5.5.12
The proposed change revises the TS administrative controls
programs for consistency with the requirements of 10 CFR [Part] 50,
paragraph 55a (g)(4) for components classified as Code Class CC.
The change affects the frequency of visual examinations that
will be performed for the concrete surfaces containments. In
addition, the proposed change allows those examinations to be
performed during power operation as opposed to during a refueling
outage. The proposed change does not involve a modification to the
physical configuration of the plant (i.e., no new equipment will be
installed) or change in the methods governing normal plant
operation. The proposed change will not impose any new or different
requirements or introduce a new accident initiator, accident
precursor, or malfunction mechanism. Additionally, there is no
change in the types or increases in the amounts of any effluent that
may be released off-site and there is no increase in individual or
cumulative occupational exposure.
Therefore, this proposed 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.
Change to Technical Specification 5.5.6
The proposed change revises the Inservice Testing Program for
consistency with the requirements of 10 CFR 50.55a(f)(4) for pumps
and valves which are classified as American Society of Mechanical
Engineers (ASME) Code Class 1, Class 2 and Class 3. The safety
function of the affected pumps and valves will be maintained.
Therefore, this proposed change does not involve a significant
reduction in a margin of safety.
Change to Technical Specification 5.5.12
The proposed change revises the Improved Standard Technical
Specification Administrative Controls program requirements for
consistency with the requirements of 10 CFR [Part] 50, paragraph 55a
(g)(4) for components classified as Code Class CC.
The change affects the frequency of visual examinations that
will be performed for the concrete surfaces of containments. In
addition, the proposed change allows those examinations to be
performed during power operation as opposed to during a refueling
outage. The safety function of the containment as a fission product
barrier will be maintained.
[Therefore, this 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: Richard J. Laufer.
PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric
Station, Unit 2 (SSES 2), Luzerne County, Pennsylvania
Date of amendment request: November 16, 2006.
Description of amendment request: The proposed amendment would
revise the SSES 2 Technical Specification (TS) Section 2.1.1.2 to
reflect the Unit 2 Cycle 14 (U2C14) Minimum Critical Power Ratio (MCPR)
Safety Limits for two-loop and single-loop operation. Additionally, TS
Section 5.6.5.b would be revised to reflect the NRC-approved
methodology used in the MCPR Safety Limit Analysis.
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:
2. Does the proposed change involve a significant increase in
the probability [* * *] or consequences of an accident previously
evaluated?
Response: No.
The proposed change to the two-loop and single-loop MCPR Safety
Limits do not directly or indirectly affect any plant system,
equipment, component, or change the processes used to operate the
plant. Further, the proposed U2C14 MCPR Safety Limits were generated
using NRC-approved methodology and meet the applicable acceptance
criteria. Thus, this proposed amendment does not involve a
significant increase in the probability of occurrence or
consequences of an accident previously evaluated.
Prior to the startup of U2C14, licensing analyses are performed
(using NRC-approved methodology referenced in Technical
Specification Section 5.6.5.b) to determine changes in the critical
power ratio as a result of anticipated operational occurrences.
These results are added to the MCPR Safety Limit values to generate
the MCPR operating limits in the U2C14 COLR [Core Operating Limits
Report]. These limits could be different from those specified for
the previous Unit 2 COLR. The COLR operating limits thus assure that
the MCPR Safety Limit will not be exceeded during normal operation
or anticipated operational occurrences. Postulated accidents are
also analyzed prior to the startup of U2C14 and the results shown to
be within the NRC-approved criteria.
The changes to the references in Section 5.6.5.b were made to
properly reflect the NRC-approved methodology used to generate the
U2C14 core operating limits. The use of this approved methodology
does not increase the probability [* * *] or consequences of an
accident previously evaluated.
Therefore, the proposed amendment 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 changes to the two-loop and single-loop MCPR Safety Limits
do not directly or indirectly affect any plant system, equipment, or
component and therefore does not affect the failure modes of any of
these items. Thus, the proposed change does not create the
possibility of a previously unevaluated operator error or a new
single failure.
The changes to the references in Section 5.6.5.b were made to
properly reflect the NRC-approved methodology used to generate the
U2C14 core operating limits. The use of this approved methodology
does not create the possibility of a new or different kind of
accident.
Therefore, this proposed amendment 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.
Since the proposed changes do not alter any plant system,
equipment, component, or the processes used to operate the plant,
the proposed change will not jeopardize or degrade the function or
operation of any plant system or component governed by Technical
Specifications. The proposed two-loop and single-loop MCPR Safety
Limits do not involve a significant reduction in the margin of
safety as currently defined in the Bases of the applicable Technical
Specification sections because the MCPR Safety Limits calculated for
U2C14 preserve the required margin of safety.
The changes to the references in Section 5.6.5.b were made to
properly reflect the NRC-approved methodology used to generate the
U2C14 core operating limits. This approved methodology is used to
demonstrate that all applicable criteria are met, thus,
demonstrating that there is no reduction in the margin of safety.
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, 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: Richard J. Laufer.
[[Page 75999]]
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: August 4, 2006.
Description of amendment request: The amendments would revise the
Technical Specifications (TSs) to allow the movement of irradiated fuel
inside containment to commence at 24 hours after shutdown or at the
decay time calculated using the licensee's spent fuel pool integrated
decay heat management program, whichever is later.
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 of occurrence or consequences of an accident previously
evaluated?
Response: No.
The proposed license amendment would allow fuel assemblies to be
removed from the reactor core and be stored in the Spent Fuel Pool
in less time after subcriticality (but more accurately calculated),
than currently allowed by the TSs. Decreasing the decay time of the
fuel affects the radionuclide make-up of the fuel to be offloaded as
well as the amount of decay heat that is present from the fuel at
the time of offload. The proposed changes do not involve a
significant increase in the probability of occurrence of an accident
previously evaluated. The accident previously evaluated that is
associated with the proposed license amendment is the fuel handling
accident. Allowing the fuel to be offloaded based on the IDHM
[integrated decay heat management program] calculated time after
subcriticality does not impact the manner in which the fuel is
offloaded. The accident initiator is the dropping of the fuel
assembly. Since earlier offload does not affect fuel handling, there
is no increase in the probability of occurrence of a fuel handling
accident. The time frame in which the fuel assemblies are moved has
been evaluated against the 10 CFR 50.67 dose limits for members of
the public, licensee personnel and control room. Additionally, the
guidance provided in Reg. Guide 1.183 was used for the selective
application of Alternative Source Term. All dose limits are met with
the reduced core offload times; and significant margin is
maintained, as the minimum decay time prior to movement of fuel for
the FHA [fuel handling accident] analysis is 24 hours.
Therefore, the proposed license amendment does not increase the
probability of occurrence or the consequences of accidents
previously evaluated are not increased.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
Response: No.
The proposed license amendment would allow core offload to occur
in less time after subcriticality (but more accurately calculated),
which affects the radionuclide make-up of the fuel to be offloaded
as well as the amount of decay heat that is present from the fuel at
the time of offload. The radionuclide makeup of the fuel assemblies
and the amount of decay heat produced by the fuel assemblies do not
currently initiate any accident. A change in the radionuclide makeup
of the fuel at the time of core offload or an increase in the decay
heat produced by the fuel being offloaded will not cause the
initiation of any accident. The accident previously evaluated that
is associated with fuel movement is the fuel handling accident.
There is no change to the manner in which fuel is being handled or
in the equipment used to offload or store the fuel. The effects of
the additional decay heat load have been analyzed. The analysis
demonstrated that the existing Spent Fuel Pool cooling system and
associated systems under worst-case circumstances would maintain the
integrity of the Spent Fuel Pool. The proposed method of offload
does not create a new or different kind of accident from any
accident previously evaluated.
Therefore, the proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
Response: No.
The margin of safety pertinent to the proposed changes is the
dose consequences resulting from a fuel handling accident. The
shorter decay time prior to fuel movement has been evaluated against
10 CFR part 50.67 and all limits continue to be met. All dose limits
are met with the reduced core offload times; and significant margin
is maintained, as the minimum decay time prior to movement of fuel
for the FHA analysis is 24 hours. Decay heatup calculations
performed prior to each refueling outage as part of the IDHM program
ensure that planned spent fuel transfer to the SFP [spent fuel pool]
will not result in maximum SFP temperature exceeding the design
basis limit of 149 [deg]F (with both heat exchangers available) or
180 [deg]F (with one heat exchanger alternating between the two
pools). As stated above, the changes in radionuclide makeup and
additional heat load do not impact any safety settings and do not
cause any safety limit to not be met. In addition, the integrity of
the Spent Fuel Pool is maintained.
The time frame in which the fuel assemblies are moved has been
evaluated against the 10 CFR 50.67 dose limits for members of the
public, licensee personnel and control room. Additionally, the
guidance provided in Reg. Guide 1.183 was used. Calculations
performed conclude that expected dose limits following a Fuel
handling Accident are met with the proposed decay time prior to
commencing fuel movement.
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, 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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: November 7, 2006.
Description of amendment requests: The amendments request to revise
Main Steam Safety Valve Requirements and Actions (Technical
Specification 3.7.1).
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.
Based on a detailed plant transient analysis, the Limiting
Conditions for Operation (LCOs) and Action statements will continue
to restrict operation to within the regions that provide acceptable
results. The safety analysis was performed in accordance with the
Nuclear Regulatory Commission (NRC) approved San Onofre Units 2 and
3 reload analysis methodology, and considered the concerns
identified in NRC Information Notice 94-60.
The increase in Completion Time for Required Action 3.7.1.A.2
from 12 hours to 36 hours is consistent with NUREG-1432 Revision 3,
``Standard Technical Specifications for Combustion Engineering
Plants.''
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 add any new equipment, modify any
interfaces with any existing equipment, alter the equipment's
function, or change the method of operating the equipment. The
proposed change does not alter plant conditions in a manner that
could affect other plant components. The proposed change does not
[[Page 76000]]
cause any existing equipment to become an accident initiator.
The increase in Completion Time for Required Action 3.7.1.A.2
from 12 hours to 36 hours is consistent with NUREG-1432 Revision 3,
``Standard Technical Specifications for Combustion Engineering
Plants.''
Therefore, the proposed 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 Limiting Conditions for Operation (LCOs) and Action
statements will continue to restrict operation such that the
American Society of Mechanical Engineers (ASME) code requirements
continue to be met. The analyses were performed using the NRC
approved San Onofre Units 2 and 3 reload analysis methodology.
Therefore, the proposed change will have no impact on the margins as
defined in the Technical Specification bases.
The increase in Completion Time for Required Action 3.7.1.A.2
from 12 hours to 36 hours is consistent with NUREG-1432 Revision 3,
``Standard Technical Specifications for Combustion Engineering
Plants.''
Therefore, this 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 requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Branch Chief: David Terao.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, Limestone County,
Alabama
Date of amendment request: November 15, 2006 (TS-459).
Description of amendment request: The proposed amendment requests
revision to the Fire Protection License Condition for Units 1, 2, and
3, condition number (13), (14), and (7), respectively, to accommodate
operation of Units 1, 2, and 3.
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.
No. The proposed change revises the license condition to reflect
a combined Units 1, 2 and 3 Fire Protection Report. Compliance with
the applicable Appendix R requirements is ensured through
implementation of the Fire Protection Program and the Appendix R
Safe Shutdown Program including Regulatory Issue Summary 2006-10,
``Regulatory Expectations with Appendix R Paragraph III.G.2 Post-
Fire Manual Actions.'' The change does not affect any design bases
accident or the ability of any safe shutdown equipment to perform
its function. Also, although modifications were required to bring
BFN in compliance with 10 CFR 50 Appendix R, there are no physical
modifications required to implement this license amendment.
Therefore, this proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
3. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed change revises the license condition to reflect
a combined Units 1, 2 and 3 Fire Protection Report. Compliance with
the applicable Appendix R requirements is ensured through
implementation of the Fire Protection Program and Appendix R Safe
Shutdown Program including Regulatory Issue Summary 2006-10,
``Regulatory Expectations with Appendix R Paragraph III.G.2 Post-
Fire Manual Actions.'' This change does not affect any design basis
accident or the ability of any safe shutdown equipment to perform
its function. Also, there are no physical modifications required to
implement this license amendment. Therefore, this proposed change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
4. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed change revises the license condition to reflect
a combined Units 1, 2 and 3 Fire Protection Report. Compliance with
the applicable Appendix R requirements is ensured through the
implementation of the Fire Protection Program and Appendix R Safe
Shutdown Program (Units 1, 2, and 3 Fire Protection Report)
including Regulatory Issue Summary 2006-10, ``Regulatory
Expectations with Appendix R Paragraph III.G.2 Post-Fire Manual
Actions.'' The proposed change does not affect any design basis
accident and does not reduce or adversely affect the capability to
achieve and maintain safe shutdown in the event of a fire.
Furthermore, no reductions to the requirements for equipment
operability, surveillance requirements or setpoints are being made
which could result in reduction in the margin of safety. Therefore,
this proposed change will not result in a 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: L. Raghavan.
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 51.12(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
[[Page 76001]]
located in ADAMS, contact the PDR Reference staff at 1 (800) 397-4209,
(301) 415-4737 or by e-mail to [email protected].
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: December 23, 2005, as
supplemented by letters dated May 4 and August 3, 2006.
Brief description of amendments: These amendments revise Technical
Specification (TS) 3.8.1, ``AC Sources--Operating,'' to extend the
allowed out of service time for one inoperable emergency diesel
generator from 72 hours to 10 days. TS 3.8.3, ``Diesel Fuel Oil, Lube
Oil, and Starting Air,'' is revised by the addition of a clarifying
note to Condition F of this specification. Additionally, TS 3.4.9,
``Pressurizer,'' is revised to delete the words contained in the
limiting condition for operation which require that the two groups of
pressurizer heaters are capable of being powered from an emergency
power supply.
Date of issuance: December 5, 2006.
Effective date: As of the date of issuance to be implemented within
90 days from the date of issuance.
Amendment Nos.: Unit 1--164, Unit 2--164, Unit 3--164
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Operating Licenses for all three units.
Date of initial notice in Federal Register: January 31, 2006 (71 FR
5080). The May 4 and August 3, 2006, supplemental letters provided
additional information that clarified the application as originally
noticed, and did not change the staff's original no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 5, 2006.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: September 30, 2004, as
supplemented by letters dated March 16, September 29, 2005, and March
21, August 7, August 24, and September 11, 2006.
Brief description of amendment: The license amendment request
revised the technical specifications and the final safety analysis
report to amend the Columbia Generating Station's licensing and design
bases to reflect the application of the alternative source term
methodology with an exception. That exception is the Technical
Information Document (TID)-14844, ``Calculation of Distance Factors for
Power and Test Reactor Sites,'' which will continue to be used as the
radiation dose basis for equipment qualification, and radiation zone
maps/shielding calculations.
Date of issuance: November 27, 2006.
Effective date: As of its date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment No.: 199.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications and Final Safety Analysis Report.
Date of initial notice in Federal Register: October 26, 2004 (69 FR
62472). The March16, September 29, 2005, and March 21, August 7, August
24, and September 11, 2006, supplemental letters 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 no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 27, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois
Date of application for amendment: October 3, 2005.
Brief description of amendment: The amendment revised the Technical
Specification (TS) Section 1.1, ``Definitions,'' description of the
Pressure and Temperature Limits Report (PTLR), by deleting reference to
specifications containing limits in the PTLR; (2) revised
administrative controls TS 5.6.6, ``Reactor Coolant System (RCS)
Pressure and Temperature Limits Report (PTLR),'' by requiring the NRC
approval documents to be identified by date and topical reports to be
identified by number and title; and (3) added Westinghouse Electric
Company, LLC report, WCAP-16143, ``Reactor Vessel Closure Head/Vessel
Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2,'' to
the list of analytical methods provided in TS 5.6.6. The amendment also
revises the title of the NRC letter dated August 8, 2001 to clarify the
regulation being referenced.
Date of issuance: November 27, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 148, 148, 142, 142.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications and License.
Date of initial notice in Federal Register: March 14, 2006 (71 FR
13175).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 27, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendment: June 2, 2006, as supplemented by
letters dated August 18 and October 5, 2006.
Brief description of amendment: The amendments revised Technical
Specification Surveillance Requirement 3.1.7.10, ``Standby Liquid
Control System Sodium Pentaborate Isotopic Enrichment'' such that the
required enrichment increases from >= 30.0 atom percent boron-10 to >=
45.0 atom percent boron-10.
Date of issuance: November 16, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 222/214.
Renewed Facility Operating License Nos. DPR-19 and DPR-25: The
amendments revised the Technical Specification Surveillance Requirement
and Licenses.
Date of initial notice in Federal Register: (71 FR 46931; August
15, 2006).
The August 18 and October 5, 2006, supplements provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards consideration
determination as published in the Federal Register on August 15, 2006
(71 FR 46931). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 16, 2006.
No significant hazards consideration comments received: No.
[[Page 76002]]
FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of application for amendment: November 14, 2005, as
supplemented by letter dated September 1, 2006.
Brief description of amendment: The amendment revises Technical
Specification (TS) Table 3.3.6.1-1, ``Primary Containment Isolation
Instrumentation,'' to eliminate the Main Steamline Radiation Monitor
trip function.
Date of issuance: November 15, 2006.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 261.
Facility Operating License No. DPR-49: The amendment revises the
TSs.
Date of initial notice in Federal Register: (71 FR 43533) August 1,
2006.
The supplement provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the NRC staff's original proposed no
significant hazards consideration determination, as published in the
Federal Register on August 1, 2006 (71 FR 43533).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 15, 2006.
No significant hazards consideration comments received: No.
FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of application for amendment: December 22, 2005.
Brief description of amendment: The amendment revises the Duane
Arnold Energy Center licensing basis, as described in the Updated Final
Safety Analysis Report (UFSAR), to replace the current plant-specific
reactor pressure vessel material surveillance program with the Boiling
Water Reactor Vessel and Internals Project Integrated Surveillance
Program as the basis for demonstrating compliance with the requirements
of Appendix H to Part 50 of Title 10 of the Code of Federal
Regulations, ``Reactor Vessel Material Surveillance Program
Requirements.''
Date of issuance: November 27, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 262.
Facility Operating License No. DPR-49: The amendment authorizes
changes to the UFSAR.
Date of initial notice in Federal Register: (71 FR 43533) August 1,
2006.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 27, 2006.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: November 9, 2005, supplemented
by letter dated May 15, 2006.
Brief description of amendments: The amendments modify the
Technical Specifications (TS) for Prairie Island Nuclear Generating
Plant Units 1 and 2, to clarify which TS Surveillance Requirements
shall be met for the TS systems which include more components
(installed spare components) than are required to satisfy the TS
Limiting Conditions for Operation. These amendments revise TS 3.7.8,
``Cooling Water (CL) System,'' TS 3.8.1, ``AC Sources-Operating,'' and
TS 3.9.3, ``Nuclear Instrumentation.'' The amendments also make minor
corrections for some of these TSs.
Date of issuance: November 14, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 175 and 165.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 14, 2006 (71
FR 7809).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 14, 2006. The supplemental
information provided in letter May 15, 2006, did not impact the
conclusions of the Determination of No Significant Hazards
Consideration and Environmental Assessment presented in the November 9,
2005 submittal.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket No. 50-275, Diablo Canyon
Nuclear Power Plant, Unit No. 1, San Luis Obispo County, California
Date of application for amendment: December 16, 2005, as
supplemented by a letter dated September 27, 2006.
Brief description of amendment: The amendment revised Technical
Specification (TS) 5.6.5, ``Core Operating Limits Report (COLR).''
Specifically, the change added Westinghouse Topical Report WCAP-12945-
P-A, Addendum 1-A, Revision 0, ``Method for Satisfying 10 CFR 50.46
Reanalysis Requirements for Best Estimate LOCA [Loss-of-Coolant
Accident] Evaluation Models,'' dated December 2004, to the list of
approved analytical methods in TS 5.6.5.b.
Date of issuance: November 21, 2006.
Effective date: As of its date of issuance, and shall be
implemented within 90 days of issuance.
Amendment No.: 191.
Facility Operating License No. DPR-80: The amendment revised the
Technical Specifications and Facility Operating License.
Date of initial notice in Federal Register: February 14, 2006 (71
FR 7810).
The September 27, 2006, supplemental letter provided additional
information that clarified the application, and did not expand the
scope of the application as originally noticed.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 21, 2006.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: December 6, 2005, as
supplemented by letters dated March 14 and November 30, 2006.
Brief description of amendments: The amendment deleted Technical
Specification (TS) Limiting Condition for Operation (LCO) 3.3.10,
``Fuel Handling Isolation Signal (FHIS),'' and TS LCO 3.7.14, ``Fuel
Handling Building Post-Accident Cleanup Filter System,'' and their
associated surveillance requirements. The amendment also deleted the
Fuel Handling Building Post-Accident Cleanup Filter Systems from the
Ventilation Filter Testing Program in administrative TS 5.5.2.12.
Date of issuance: December 4, 2006.
Effective date: As of its date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: Unit 2--208; Unit 3--200.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: January 3, 2006 (71 FR
155). The March 14 and November 30, 2006, supplemental letters provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed,
[[Page 76003]]
and did not change the staff's original no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 4, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259 Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of application for amendment: October 12, 2004, as
supplemented by letters dated September 7 and November 1, 2006.
Brief description of amendment: To remove License Condition 2.C(4).
Date of issuance: November 28, 2006.
Effective date: As of date of issuance, to be implemented within 30
days.
Amendment No.: 265
Renewed Facility Operating License Nos. DPR-33: Amendment revised
the Renewed Operating License.
Date of initial notice in Federal Register: August 15, 2006 (71 FR
46937). The supplements dated September 7 and November 1, 2006,
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 November 28, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: May 25, 2006, as supplemented
by letter dated September 1, 2006.
Brief description of amendments: The requested changes provide a
revision to the design and licensing basis for the containment sump
debris transport analysis as described in the Sequoyah Nuclear Plant
(SQN) Updated Final Safety Analysis Report (UFSAR). The current
transport analysis for SQN is a two-dimensional physical transport
model, and Tennessee Valley Authority (TVA) is requesting to update the
analysis to a three-dimensional computational fluid dynamics transport
model. The results of the reanalysis will be used to size the flow area
of the advanced design containment sump strainers which will replace
the original sump intake structure.
Date of issuance: November 7, 2006.
Effective date: Implementation of the amendment is the
incorporation into the next UFSAR update made in accordance with 10 CFR
50.71(e), of the changes to the description of the facility as
described in TVA's application dated May 25, 2006, as supplemented by
letter dated September 1, 2006, and evaluated in the staff's Safety
Evaluation attached to this amendment.
Amendment Nos. 313 and 302.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: June 20, 2006 (71 FR
35460). The supplemental letter dated September 1, 2006, provided
clarifying information that was within the scope of the initial notice
and did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 7, 2006.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: September 20, 2006, as
supplemented by letter dated November 20, 2006.
Brief description of amendment: The amendment revised (1) the
definition of the Pressure and Temperature Limits Report (PTLR) in
Technical Specification (TS) 1.1, ``Definitions,'' and (2) TS 5.6.6,
``Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT
(PTLR).''
Date of issuance: December 5, 2006.
Effective date: As of its date of issuance, and shall be
implemented within 90 days of the date of issuance.
Amendment No.: 177.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: October 6, 2006 (71 FR
59136).
The supplemental letter dated November 20, 2006, provided
additional clarifying information, 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
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 5, 2006.
No significant hazards consideration comments received: No.
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.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of application for amendments: November 9, 2006 (TS-458).
Description of amendments request: The proposed amendment would
delete the Technical Specification Surveillance Requirement to verify
the position of a low pressure coolant injection crosstie valve.
Date of publication of individual notice in the Federal Register:
November 20, 2006 (71 FR 67166).
Expiration date of individual notice: December 20, 2006 (Public
comments) and January 19, 2007 (Hearing requests).
Dated at Rockville, Maryland, this 11th day of December 2006.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E6-21346 Filed 12-18-06; 8:45 am]
BILLING CODE 7590-01-P