[Federal Register Volume 72, Number 58 (Tuesday, March 27, 2007)]
[Notices]
[Pages 14303-14312]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-5342]
-----------------------------------------------------------------------
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 March 2, 2007 to March 15, 2007. The last
biweekly notice was published on March 13, 2007 (72 FR 11383).
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
[[Page 14304]]
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
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: February 8, 2007.
Description of amendment request: The proposed amendment would
modify Grand Gulf Nuclear Station, Unit 1 (GGNS) technical
specification (TS) requirements for MODE change limitations in limiting
condition for operation (LCO) 3.0.4 and surveillance requirement (SR)
3.0.4. The proposed TS changes are consistent with Revision 9 of
Nuclear Regulatory Commission (NRC) approved Industry TS Task Force
(TSTF) Standard TS Change Traveler, TSTF-359, ``Increase Flexibility in
MODE Restraints.'' In addition, the proposed amendment would also
change TS Section 1.4, Frequency, Example 1.4-1, ``Surveillance
Requirements,'' to accurately reflect the changes made by TSTF-359,
which is consistent with NRC-approved TSTF-485, Revision 0, ``Correct
Example 1.4-1.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 2, 2002 (67 FR 50475), as part of the
Consolidated Line Item Improvement Process (CLIIP), on possible
amendments to revise the
[[Page 14305]]
plant-specific TS to modify requirements for MODE change limitations in
LCO 3.0.4 and SR 3.0.4.
The NRC staff subsequently issued a notice of availability of the
models for Safety Evaluation and No Significant Hazards Consideration
Determination for referencing in license amendment applications in the
Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed
the applicability of the CLIIP, including the model No Significant
Hazards Consideration Determination, in its application dated February
8, 2007.
The proposed TS changes are consistent with NRC-approved Industry
TSTF Standard TS change, TSTF-359, Revision 8, as modified by 68 FR
16579. TSTF-359, Revision 8, was subsequently revised to incorporate
the modifications discussed in the April 4, 2003, Federal Register
notice and other minor changes. TSTF-359, Revision 9, was subsequently
submitted to the NRC on April 28, 2003, and was approved by the NRC on
May 9, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the NRC staff's analysis
of the issue of no significant hazards consideration is presented
below:
Criterion 1--The Proposed Changes Do Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously Evaluated
The proposed changes in TS Section 1.4, Frequency, Example 1.4-1,
would accurately reflect the changes made by TSTF-359 in LCO 3.0.4 and
SR 3.0.4, which are consistent with NRC-approved TSTF-485, Revision 0.
These changes are considered administrative in that they modify the
example to demonstrate the proper application of LCO 3.0.4 and SR
3.0.4. The requirements of LCO 3.0.4 and SR 3.0.4 are clear and are
clearly explained in the associated Bases. As a result, modifying the
example will not result in a change in usage of the TS.
The proposed changes in LCO 3.0.4 and SR 3.0.4 allow entry into a
mode or other specified condition in the applicability of a TS, while
in a TS condition statement and the associated required actions of the
TS. The proposed changes do not adversely affect accident initiators or
precursors, the ability of structures, systems, and components to
perform their intended function to mitigate the consequences of an
initiating event within the assumed acceptance limits, or radiological
release assumptions used in evaluating the radiological consequences of
an accident previously evaluated. Being in a TS condition and the
associated required actions are not an initiator of any accident
previously evaluated. Therefore, the probability of an accident
previously evaluated is not significantly increased. The consequences
of an accident while relying on required actions as allowed by proposed
LCO 3.0.4, are no different than the consequences of an accident while
entering and relying on the required actions while starting in a
condition of applicability of the TS. Therefore, the consequences of an
accident previously evaluated are not significantly affected by these
changes. The addition of a requirement to assess and manage the risk
introduced by these changes will further minimize possible concerns.
Therefore, these changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Changes Do Not Create the Possibility of a
New or Different Kind of Accident from any Previously Evaluated
No new or different accidents result from utilizing the proposed
changes. The proposed changes do not involve a physical alteration of
the plant (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 proposed changes do not alter
assumptions made in the safety analysis and are consistent with the
safety analysis assumptions and current plant operating practice.
Entering into a mode or other specified condition in the applicability
of a TS, while in a TS condition statement and the associated required
actions of the TS, 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 these changes will further minimize
possible concerns. Thus, these changes do not create the possibility of
a new or different kind of accident from an accident previously
evaluated.
Criterion 3--The Proposed Changes Do Not Involve a Significant
Reduction in the Margin of Safety
The proposed changes in TS Section 1.4, Example 1.4-1, are
considered administrative and will have no effect on the application of
the TS requirements. Therefore, the margin of safety provided by the TS
requirements is unchanged. The proposed changes in TS LCO 3.0.4 and SR
3.0.4 allow entry into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS. The GGNS TS allows operation of
the plant without the full complement of equipment through the TS
conditions for not meeting the TS LCO. The risk associated with this
allowance is managed by the imposition of required actions that must be
performed within the prescribed completion times. The net effect of
being in a TS LCO condition on the margin of safety is not considered
significant. The proposed changes do not alter the required actions or
completion times of the TS. The proposed changes allow TS conditions to
be entered, and the associated required actions and completion times to
be used in new circumstances. This use is predicated upon the
licensee's performance of a risk assessment and the management of plant
risk. The changes also eliminate current allowances for utilizing
required actions and completion times in similar circumstances, without
assessing and managing risk. The net change to the margin of safety is
insignificant. Therefore, these changes do 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, 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, Units 1
and 2, Will County, Illinois.
Date of amendment request: January 8, 2007.
Description of amendment request: The proposed amendment would
revise the technical specification (TS) requirements for selected
reactor trip system (RTS) instrumentation, engineered safety feature
actuation system (ESFAS) instrumentation, and containment ventilation
isolation instrumentation to adopt completion times, test bypass time,
and surveillance test interval changes. The changes are based on
Westinghouse Electric Company, LLC, topical reports WCAP-14333-P-A,
Revision 1, ``Probabilistic
[[Page 14306]]
Risk Analysis of the [Reactor Protection System] RPS and ESFAS Test
Times and Completion Times,'' and WCAP-15376-P-A, Revision 1, ``Risk-
Informed Assessment of the RTS and ESFAS Surveillance Test Intervals
and Reactor Trip Breaker Test and Completion Times.''
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. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since no
hardware changes are proposed. The same RTS and ESFAS
instrumentation will continue to be used. The protection systems
will continue to function in a manner consistent with the plant
design basis. These changes to the TS do not result in a condition
where the design, material, and construction standards that were
applicable prior to the change are altered.
The proposed changes will not modify any system interface. The
proposed changes will not affect the probability of any event
initiators. There will be no degradation in the performance of or an
increase in the number of challenges imposed on safety-related
equipment assumed to function during an accident situation. There
will be no change to normal plant operating parameters or accident
mitigation performance. The proposed changes will not alter any
assumptions or change any mitigation actions in the radiological
consequence evaluations in the Updated Final Safety Analysis Report.
The determination that the results of the proposed changes are
acceptable was established in the NRC Safety Evaluations prepared
for WCAP-14333-P-A, (issued by letter dated July 15, 1998) and for
WCAP-15376-P-A, (issued by letter dated December 20, 2002).
Implementation of the proposed changes will result in an
insignificant risk impact.
Applicability of these conclusions has been verified through
plant-specific reviews and implementation of the generic analysis
results in accordance with the respective NRC Safety Evaluation
conditions.
The proposed changes to the CTs [completion times], test bypass
times, and Surveillance Frequencies reduce the potential for
inadvertent reactor trips and spurious engineered safeguard features
actuations, and therefore do not increase the probability of any
accident previously evaluated. The proposed changes do not change
the response of the plant to any accidents and have an insignificant
impact on the reliability of the RTS and ESFAS signals. The RTS and
ESFAS will remain highly reliable and the proposed changes will not
result in a significant increase in the risk of plant operation.
This is demonstrated by showing that the impact on plant safety, as
measured by the increase in core damage frequency (CDF) is less than
1.0E-06 per year and the increase in large early release frequency
(LERF) is less than 1.0E-07 per year. In addition, for the CT
changes, the incremental conditional core damage probabilities
(ICCDP) and incremental conditional large early release
probabilities (ICLERP) are less than 5.0E-07 and 5.0E-08,
respectively. These changes meet the acceptance criteria in
Regulatory Guides (RGs) 1.174 and 1.177. Therefore, since the RTS
and ESFAS will continue to perform their functions with high
reliability, as originally assumed, and the increase in risk, as
measured by [Delta]CDF, [Delta]LERF, ICCDP, ICLERP risk metrics, is
within the acceptance criteria of existing regulatory guidance,
there will not be a significant increase in the consequences of any
accidents.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, or
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not alter or
prevent the ability of structures, systems, and components from
performing their intended function to mitigate the consequences of
an initiating event within the assumed acceptance limits. The
proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of any accident previously evaluated.
The proposed changes are consistent with safety analysis assumptions
and resultant consequences.
Therefore, this change does not increase the probability or
consequences of any accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated. There are no hardware changes nor are there any changes
in the method by which any safety-related plant system performs its
safety function. The proposed changes will not affect the normal
method of plant operation. No performance requirements will be
affected or eliminated. The proposed changes will not result in
physical alteration to any plant system nor will there be any change
in the method by which any safety-related plant system performs its
safety function. There will be no setpoint changes or changes to
accident analysis assumptions.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of these changes. There will be no adverse effect or challenges
imposed on any safety-related system as a result of these changes.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety?
The proposed changes do not affect the acceptance criteria for
any analyzed event nor is there a change to any Safety Analysis
Limit. There will be no effect on the manner in which safety limits,
limiting safety system settings, or limiting conditions for
operation are determined nor will there be any effect on those plant
systems necessary to assure the accomplishment of protection
functions. There will be no impact on the departure from nucleate
boiling limits, fuel centerline temperature, or any other margin of
safety. The radiological dose consequence acceptance criteria listed
in the NUREG-0800, ``Standard Review Plan for the Review of Safety
Analysis Reports for Nuclear Power Plants,'' will continue to be
met.
Redundant RTS and ESFAS trains are maintained, and diversity
with regard of the signals that provide reactor trip and engineered
safety features actuation is also maintained. All signals credited
as primary or secondary, and all operator actions credited in the
accident analyses will remain the same. The proposed changes will
not result in plant operation in a configuration outside the design
basis. The calculated impact on risk is insignificant and meets the
acceptance criteria contained in RGs 1.174 and 1.177. Although there
was no attempt to quantify any positive human factors benefit due to
increased CTs and bypass test times, it is expected that there would
be a net benefit due to a reduced potential for spurious reactor
trips and actuations associated with testing.
Implementation of the proposed changes is expected to result in
an overall improvement in safety, as follows:
Reduced testing will result in fewer inadvertent
reactor trips, less frequent actuation of ESFAS components, less
frequent distraction of operations personnel without significantly
affecting RTS and ESFAS reliability.
Improvements in the effectiveness of the operating
staff in monitoring and controlling plant operation will be
realized. This is due to less frequent distraction of the operators
and shift supervisor to attend to instrumentation Required Actions
with short CTs.
Longer repair times associated with increased CTs will
lead to higher quality repairs and improved reliability.
The CT extensions for the reactor trip breakers will
provide additional time to complete test and maintenance activities
while at power, potentially reducing the number of forced outages
related to compliance with reactor trip breaker CT, and provide
consistency with the CT for the logic trains.
Therefore, the proposed changes do 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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell A. Gibbs.
[[Page 14307]]
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: January 30, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) Surveillance Requirement (SR)
3.5.1.3.b to correctly state that the required pressure at which the
Alternate Nitrogen System is determined to be operable should be
greater than or equal to 410 psig, not the currently stated pressure of
greater than or equal to 220 psig. The safety-related Alternate
Nitrogen System provides an alternate pressure source to equipment
required during or following an accident. The licensee has determined
that the current acceptance value specified by SR 3.5.1.3.b is non-
conservative and needs to be corrected to the higher value.
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).
The NRC staff reviewed the licensee's analysis, and has performed its
own analysis as follows:
(1) Does the proposed amendment involve a significant increase
in the probability or consequences of an accident previously
evaluated?
No. The proposed amendment would only correct the acceptance
value specified by SR 3.5.1.3.b. The acceptance value of the
nitrogen supply was not considered to be a precursor to, and does
not affect the probability of, an accident. In addition, there is no
design or operation change associated with the proposed amendment.
Therefore, the proposed amendment does not increase the
probability of an accident previously evaluated.
The corrected, higher pressure of the Alternate Nitrogen System
will ensure that nitrogen is available to operate equipment after an
accident, as designed. The increased acceptance value will not
decrease the functionality of the Alternate Nitrogen System, or the
functionality of the plant equipment it supports. Therefore, the
plant systems required to mitigate accidents will remain capable of
performing their design functions. As a result, the proposed
amendment will not lead to a significant change in the consequences
of any accident.
(2) Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
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 revised acceptance value.
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?
No. The proposed amendment only changes the acceptance value of
the Alternate Nitrogen System. There will be no modification of any
TSs limiting condition for operation, no change to any limit on
previously analyzed accidents, no change to how previously analyzed
accidents or transients would be mitigated, no change in any
methodology used to evaluate consequences of accidents, and no
change in any operating procedure or process. Therefore, the
proposed amendment does not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
the NRC staff's own analysis above, it appears that the three standards
of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the proposed amendment involves no significant hazards
consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Branch Chief: L. Raghavan.
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: February 8, 2007.
Description of amendment requests: This license amendment request
will (1) revise Technical Specification (TS) Surveillance Requirement
(SR) 3.3.7.3.a to lower the allowable value for dropout and raise the
allowable value for pickup of the degraded voltage function, and (2)
revise TS SR 3.8.1 to lower the diesel generator minimum output voltage
due to lower settings for the degraded voltage function.
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.
This proposed change revises the Technical Specification (TS)
Surveillance Requirement (SR) 3.3.7.3.a allowable values of the
Degraded Voltage Function and SRs 3.8.1.2, .7, .9, .11, .12, .15,
.16, .17, .19, and .20 for Diesel Generator (DG) minimum operable
voltage. This proposed change will allow Southern California Edison
(SCE) to widen the operating band while maintaining adequate
conservatism for the degraded relay settings and overall loop
uncertainties while keeping 218 kV as the minimum voltage on the
offsite transmission grid necessary to support operability of the
immediate access offsite power source (also referred to as the
normal preferred power source). This will be accomplished by
lowering the dropout and increasing the pickup settings of the
degraded voltage protection relays. Following approval of this
proposed change, the 4.16 kV Class 1E buses would remain on the
normal preferred power source at or above a grid voltage of 218 kV
while protecting all Class 1E equipment from degraded grid
conditions.
The degraded voltage protection circuits are designed to protect
electrical equipment against the effects of degraded voltage on the
offsite transmission networks. Therefore, these circuits are
generally not considered to be accident initiators. However,
spurious actuation of the degraded voltage protection relays could
result in the loss of the preferred power source (offsite source of
alternating current (AC) power). The proposed change lowers the
allowable value for dropout and raises the allowable value for
pickup for the degraded voltage protection relays. This results in
an increase in operating band and a lower probability of spurious
actuation of these degraded voltage signals. Therefore, there is no
increase in the probability of a Loss of Offsite Power (preferred
power source) as a result of this proposed change.
The safety function of the degraded voltage protection circuits
is to ensure the operability of Class 1E equipment. SCE has
performed calculations that demonstrate that operation in accordance
with this proposed change will not result in operation of plant
equipment at degraded voltages. Therefore, there is no increase in
the consequences of any accident previously evaluated.
Therefore, the proposed 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 allowable values of the degraded voltage relays and
the DG minimum operating voltage will provide an acceptable level of
protection for plant equipment.
This proposed change affects only the voltage settings of the
degraded voltage protection relays and voltage regulator setting of
the DG for lowering the required bus voltage. There is no other
change to the degraded voltage function. There are no physical
modifications necessary to the degraded voltage protection relays or
the DG. There are no changes to the actions performed by the relays
or the DG following actuation. Therefore, there are no new failure
[[Page 14308]]
modes or effects introduced by this proposed change.
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 proposed degraded voltage protection schemes are designed to
ensure that plant equipment will not operate at a degraded voltage
and the DG Automatic Voltage Regulator (AVR) is set to provide
adequate voltage for resetting of the relays and satisfactory
operation of the Safety Related equipment. The proposed degraded
voltage allowable values will not affect the existing protection
criterion for plant equipment. This maintains the existing margin of
safety for plant equipment.
Therefore, there is no significant reduction in a margin of
safety as a result of the proposed amendment.
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.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia
Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units
1 and 2, Appling County, Georgia.
Date of amendment request: February 13, 2007.
Description of amendment request: The proposed amendment would
modify the licensee's Technical Specification (TS) Section 3.9.1,
``Refueling Equipment Interlocks,'' to add required actions to allow
insertion of a control rod withdrawal block and verification that all
control rods are fully inserted as alternate actions to suspending in-
vessel fuel movement in the event that one or more required refueling
equipment interlocks are inoperable. These changes are based on
Technical Specification Task Force (TSTF) change TSTF-225, Revision 2,
``Fuel movement with inoperable refueling equipment interlocks'' and
are consistent with the current Boiling Water Reactor (BWR)/4 Standard
Technical Specifications (STS), NUREG-1433, Volume 1, Revision 3.0.
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?
The proposed change provides additional actions for an
inoperable refueling equipment interlock. The proposed actions will
allow fuel movement with inoperable refueling interlocks, however,
those actions will require the insertion of a continuous control rod
withdrawal block, as well as verification that all control rods are
fully inserted, before the commencement of fuel movement. Since fuel
movement with the refueling interlocks operable allows control rod
withdrawal under some circumstances, complete prevention of control
rod withdrawal with the refueling interlocks inoperable does not
increase the likelihood of a reactivity event, and may in fact
decrease its probability of occurrence.
The refueling interlocks are not designed or otherwise intended
to prevent or mitigate the consequences of the fuel handling
accident. This proposed change does not involve those structures
that could have an effect on the fuel handling accident and its
consequences, such as the fuel design, the integrity of the
refueling platform, and the integrity of the refueling mast and
grapple. Furthermore, the consequences of the refueling accident are
not increased since, should that accident occur while operating
under the provisions of the alternate actions, all control rods will
be fully inserted. The consequences of the fuel assembly insertion
error event during refueling are not increased since this proposed
change preserves the initial conditions of that transient event,
i.e., all control rods inserted.
Implementing these changes will not increase the likelihood of
an equipment failure resulting from the use of the refueling cranes
and hoists. Such protection is afforded by other plant (owner
controlled) specifications and procedures. These documents require
testing and maintenance of these components separate from the
requirements of [Limiting Condition for Operation] LCO 3.9.1.
This submittal does not affect any other system, structure or
component that is important with respect to the prevention and
mitigation of other accidents or transients.
For the above reasons, this proposed Technical Specifications
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?
The proposed change provides additional actions (the insertion
of a control rod block and verification that all control rods are
fully inserted) for inoperable refueling interlocks. This change
does not involve any permanent alterations to plant systems or
components. Nor does it involve changes to operational
configurations or to the maintenance and testing of systems or
components. Consequently, no new modes of operation are being
introduced. Therefore, the 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
the margin of safety?
The proposed change provides additional actions for an
inoperable required refueling equipment interlock. The new actions
will require that all control rods be fully inserted and that a
control rod block be in effect. Under the current specifications,
control rod withdrawal is allowed during fuel movement under certain
conditions.
The alternate actions of the proposed specifications will not
allow rod withdrawal under any circumstances during fuel movement
operations, therefore, this proposed change provides a level of
safety at least equivalent to the existing actions.
Consequently, the 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: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Branch Chief: Evangelos C. Marinos.
Virginia Electric and Power Company
Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and
2, Surry County, Virginia.
Date of amendment request: February 26, 2007.
Description of amendment request: The proposed change adds an
operating license condition and revises the Technical Specifications to
permit the replacement of main control room (MCR) and emergency
switchgear room (ESGR) air-conditioning system (ACS) chilled water
piping by using temporary 45-day and 14-day allowed outage times (AOTs)
four times in a 24-month time span.
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 license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The proposed change has been evaluated using the risk-informed
processes described
[[Page 14309]]
in Regulatory Guide (RG) 1.174, ``An Approach for Using
Probabilistic Risk assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis,'' and RG 1.177, ``An
Approach for Plant-Specific, Risk-Informed Decision Making:
Technical Specifications.''
The risk associated with the proposed change was found to be
acceptably ``small'' and therefore not a significant increase in the
probability or consequences of an accident previously evaluated.
In addition, the proposed change does not affect the initiators
of analyzed events or the assumed mitigation of accident or
transient events. During the temporary 45-day and 14-day AOT
entries, equipment availability restrictions will restrict or limit
the out-of-service time of risk significant plant equipment due to
surveillance testing, preventive maintenance, and elective
maintenance. In addition, during the replacement activities,
compensatory actions will be in place to ensure the availability of
chilled water or to provide backup cooling. Therefore, the ACS will
continue to perform its required function. As a result, the proposed
change to the Surry TS does not involve any significant increase in
the probability or the consequences of any accident or malfunction
of equipment important to safety previously evaluated since neither
accident probabilities nor consequences are being affected by this
proposed change.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
The proposed change does not involve a change in the methods
used to respond to plant transients. There is no alteration to the
parameters within which the plant is normally operated or in the
setpoints, which initiate protective or mitigative actions. The MCR
and ESGR ACS will continue to perform its required function. This is
assured by the planned implementation of compensatory actions,
including provisions for backup cooling. Consequently, no new
failure modes are introduced by the proposed change. Therefore, the
proposed Surry TS change does not create the possibility of a new or
different kind of accident or malfunction of equipment important to
safety from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Margin of safety is established through the design of the plant
structures, systems, and components, the parameters within which the
plant is operated, and the establishment of the setpoints for the
actuation of equipment relied upon to respond to an accident or
transient event. The proposed change does not affect the ability of
the MCR and ESGR ACS to perform its required function. This is
assured by the planned implementation of compensatory actions,
including provisions for backup cooling. Furthermore, the proposed
change has been evaluated using the risk-informed processes
described in Regulatory Guide (RG) 1.174, ``An approach for Using
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis,'' and RG 1.177, ``An
Approach for Plant-Specific, Risk-Informed Decision Making:
Technical Specifications.''
The risk associated with the proposed change was found to be
acceptably small. Therefore, the proposed change to the Surry TS
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: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Branch Chief: Evangelos C. Marinos.
Virginia Electric and Power Company
Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and
2, Surry County, Virginia.
Date of amendment request: March 6, 2007.
Description of amendment request: The proposed amendments would
revise the licensing basis (Updated Final Safety Analysis Report
(UFSAR)) to permit irradiation of the fuel assemblies beginning with
Surry Power Station, Unit Nos. 1 and 2, improved fuel assemblies with
ZIRLO (Westinghouse trademark) cladding to a lead rod average burnup of
62,000 MWD/MTU.
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. The probability of occurrence or the consequence of an
accident previously evaluated is not significantly increased.
The activity being evaluated is a slight increase in the lead
rod average burnup limit for the fuel assemblies. No change in fuel
design or fuel enrichment will be required to increase the lead rod
average burnup. The fuel rods at the extended lead rod average
burnup will continue to meet the design limits with respect to fuel
rod growth, clad fatigue, rod internal pressure and corrosion. There
will be no impact on the capability to engage the fuel assemblies
with the handling tools. Therefore, it is concluded that the change
will not result in an increase in the probability of occurrence of
any accident previously evaluated in the UFSAR. The impact of
extending the lead rod average burnup to 62,000 MWD/MTU from 60,000
MWD/MTU on the core kinetics parameter, core thermal-hydraulics/
[departure from nucleate boiling ratio]DNBR, specific accident
considerations, and radiological consequences was considered. Based
on the evaluation of these considerations, it is concluded that
increasing the lead rod average burnup limit to 62,000 MWD/MTU will
not result in a significant increase in the consequences of the
accidents previously evaluated in the Surry UFSAR.
2. The possibility for a new or different type of accident from
any accident previously evaluated is not created.
The fuel is the only component affected by the change in the
burnup limit. The change does not affect the thermal hydraulic
response to any transient or accident. The existing fuel rod design
criteria continue to be met at the higher burnup limit. Thus, the
change does not create the possibility of an accident of a different
type.
3. The margin of safety as defined in the Bases to the Surry
Technical Specifications is not significantly reduced.
The operation of the Surry cores with a limited number of fuel
assemblies with some fuel rods irradiated to a lead rod average
burnup of 62,000 MWD/MTU will not change the performance
requirements of any system or component such that any design
criteria will be exceeded. The normal limits on core operation
defined in the Surry Technical Specifications will remain applicable
for the irradiation of the fuel to a lead rod average burnup of
62,000 MWD/MTU. Therefore, the margin of safety as defined in the
Bases to the Surry Technical Specifications is not significantly
reduced.
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: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Branch Chief: Evangelos C. Marinos.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
[[Page 14310]]
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Florida Power Corporation, et al.
Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant,
Citrus County, Florida.
Date of amendment request: February 8, 2007.
Description of amendment request: To change the basis for
protection of spent fuel stored in the spent fuel pool (SFP) in order
to eliminate the Final Safety Analysis Report commitment for
maintaining the SFP missile shields.
Date of publication of individual notice in the Federal Register:
March 13, 2007. (72 FR 11381).
Expiration date of individual notice: May 14, 2007.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments.
If the Commission has prepared an environmental assessment under
the special circumstances provision in 10 CFR 51.22(b) and has made a
determination based on that assessment, it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the Internet at the NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Carolina Power & Light Company
Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2,
Darlington County, South Carolina.
Date of application for amendment: May 30, 2006, as supplemented by
letter dated November 20, 2006.
Brief description of amendment: The amendment revises the existing
steam generator tube surveillance program at H. B. Robinson Steam
Electric Plant, Unit No. 2.
Date of issuance: March 12, 2007.
Effective date: This license amendment is effective as of the date
of issuance and shall be implemented within 60 days.
Amendment No. 212.
Renewed Facility Operating License No. DPR-23. Amendment revises
the Technical Specifications.
Date of initial notice in Federal Register: December 19, 2007 (71
FR 75990). The November 20, 2006, supplemental letter provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated: March 12, 2007.
No significant hazards consideration comments received: No.
Dominion Energy Kewaunee, Inc.
Docket No. 50-305, Kewaunee Power Station, Kewaunee County,
Wisconsin.
Date of application for amendment: January 30, 2006, as
supplemented by letter dated January 23, 2007.
Brief description of amendment: The amendment modifies the
radiological accident analyses and associated technical specifications.
Date of issuance: March 8, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 190.
Facility Operating License No. DPR-43: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 14, 2006 (71 FR
13172).
The supplemental letter contained clarifying information and did
not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 8, 2007.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc.
Docket No. 50-423, Millstone Power Station, Unit No 3, New London
County, Connecticut.
Date of application for amendment: February 7, 2006, as
supplemented by letters dated August 14, 2006, and January 2, 2007.
Brief description of amendments: The amendment revised the
Millstone Power Station, Unit No. 3 Technical Specifications to permit
an increase in the allowed outage time from 72 hours to 7 days for the
inoperablity of the steam supply to the turbine-driven auxiliary
feedwater pump (AFW) or the inoperability of the turbine-driven AFW
pump under certain operating mode restrictions.
Date of issuance: February 28, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 235.
Facility Operating License No NPF-49: Amendment revised the License
and Technical Specifications.
Date of initial notice in Federal Register: April 11, 2006 (70 FR
18372).
The supplements dated August 14, 2006, and January 2, 2007,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 28, 2007.
No significant hazards consideration comments received: No.
Entergy Operations, Inc.
Docket Nos. 50-313 and 50-368, Arkansas Nuclear One, Units 1 and 2,
Pope County, Arkansas.
Date of amendment request: October 25, 2005, as supplemented by
letter dated March 20, 2006.
[[Page 14311]]
Brief description of amendments: The changes addressed inventory
and inspection requirements associated with the emergency cooling pond,
which is a common cooling water source for both units during conditions
that may render the normal cooling water source (Dardanelle Reservoir)
unavailable.
Date of issuance: March 9, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: Unit 1-229, Unit 2-271.
Renewed Facility Operating License Nos. DPR-51 and NPF-6:
Amendments revised the Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: October 24, 2006 (71 FR
62309). The supplemental letter dated March 20, 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 amendments is contained
in a Safety Evaluation dated March 9, 2007.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC
Docket No. 50-220, Nine Mile Point Nuclear Station, Unit No. 1,
Oswego County, New York.
Date of application for amendment: October 19, 2006, as
supplemented by letter dated January 5, 2007.
Brief description of amendment: The amendment revises the
Surveillance Requirement (SR) in Technical Specification (TS) 4.1.1.c,
``Scram Insertion Times,'' to modify the conditions under which scram
time testing (STT) of control rods is required, and to add a
requirement to perform STT on a defined portion of control rods, at a
specified frequency, during the operating cycle. The amendment also
revises the SR in TS 4.1.7.c, ``Minimum Critical Power Ratio (MCPR),''
to add a requirement to determine the MCPR operating limits following
completion of control rod STT per TS 4.1.1.c.
Date of issuance: March 15, 2007.
Effective date: March 15, 2007.
Amendment No.: 193.
Facility Operating License No. DPR-63: Amendment revised the
License and Technical Specifications.
Date of initial notice in Federal Register: December 5, 2006 (71 FR
70562) The supplemental letter dated January 5, 2007, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 15, 2007.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC
Docket No. 50-311, Salem Nuclear Generating Station, Unit No. 2,
Salem County, New Jersey.
Date of application for amendment: April 6, 2006.
Brief description of amendment: The amendment changed the Technical
Specifications (TSs) to reduce the maximum allowable reactor power
level when two main steam safety valves are inoperable.
Date of issuance: March 7, 2007.
Effective date: As of the date of issuance and shall be implemented
prior to restart from the steam generator replacement outage.
Amendment No.: 259.
Facility Operating License No. DPR-75: The amendment revised the
TSs and the License.
Date of initial notice in Federal Register: November 7, 2006 (71 FR
65144).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 7, 2007.
No significant hazards consideration comments received: No.
R.E. Ginna Nuclear Power Plant, LLC
Docket No. 50-244, R.E. Ginna Nuclear Power Plant, Wayne County,
New York.
Date of application for amendment: May 1, 2006, as supplemented by
letter dated November 3, 2006.
Brief description of amendment: The amendment revises the steam
generator tube integrity Technical Specifications consistent with the
Nuclear Regulatory Commission's approved Technical Specification Task
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
449, ``Steam Generator Tube Integrity,'' Revision 4.
Date of issuance: March 1, 2007.
Effective date: As of the date of issuance to be implemented within
90 days.
Amendment No.: 100.
Renewed Facility Operating License No. DPR-18: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: June 6, 2006 (71 FR
32605).
The supplemental letter dated November 3, 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 March 1, 2007.
No significant hazards consideration comments received: No.
Tennessee Valley Authority (TVA)
Docket No. 50-259, Browns Ferry Nuclear Plant (BFN), Unit 1,
Limestone County, Alabama.
Date of application for amendment: September 22, 2006.
Brief description of amendment: The amendment supplements a June
28, 2004, request to increase the licensed thermal power from 3293
megawatt thermal (MWt) to 3952 MWt, an approximate 20% increase in
thermal power. This supplement requests interim approval of an increase
in licensed thermal power from 3293 MWt to 3458 MWt with an attendant
30-psi increase in reactor pressure. This represents an approximate 5%
increase above the original licensed thermal power of 3293 MWt. An
interim approval would provide for operation at 105% power until such
time as certain steam dryer analyses can be completed. The NRC staff's
review of the remainder of the June 2004 application would resume upon
receipt of the satisfactorily completed steam dryer analyses.
Date of issuance: March 6, 2007.
Effective date: Date of issuance, to be implemented prior to
restart.
Amendment No.: 269.
Renewed Facility Operating License No. DPR-33: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: October 10, 2006 (71 FR
59532). The Commissions related evaluation of the amendment is
contained in an Environmental Assessment dated February 12, 2007 (72 FR
6612), and in a Safety Evaluation dated March 6, 2007.
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: March 28, 2006, as supplemented
by letter dated November 17, 2006.
Brief description of amendment: The amendment deleted references to
specific isolation valves in the chemical
[[Page 14312]]
and volume control system (CVCS) and modified to allow (1) an exception
for decontamination activities and (2) an exception for CVCS resin
vessel operation. These are changes to TS 3.3.9, ``Boron Dilution
Mitigation System (BDMS),'' and TS 3.9.2, ``Unborated Water Source
Isolation Valves.''
Date of issuance: March 8, 2007.
Effective date: As of its date of issuance and shall be implemented
within 90 days of the date of issuance.
Amendment No.: 181.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: May 9, 2006 (71 FR
27004). The supplemental letter dated November 17, 2006, 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 published in the Federal Register. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated March 8, 2007.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company
Docket Nos. 50-338 and 50-339, North Anna Power Station, Units 1
and 2, Louisa County, Virginia.
Date of application for amendment: October 3, 2006, as supplemented
by letter dated January 24, 2007.
Brief description of amendment: The proposed amendments revised the
Technical Specifications (TSs) and licensing basis to support the
resolution of the Nuclear Regulatory Commission's (NRC's) Generic
Safety Issue (GSI) 191, assessment of debris accumulation on
containment sump performance and its impact on emergency recirculation
during an accident, and NRC Generic Letter (GL) 2004-02.
Date of issuance: March 13, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 250 and 230.
Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments
change the licenses and the technical specifications.
Date of initial notice in Federal Register: December 5, 2006 (71 FR
70563). The supplement dated January 24, 2007, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated March 13, 2007.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 19th day of March 2007.
For the Nuclear Regulatory Commission.
John W. Lubinski,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. E7-5342 Filed 3-26-07; 8:45 am]
BILLING CODE 7590-01-P