[Federal Register Volume 72, Number 117 (Tuesday, June 19, 2007)]
[Notices]
[Pages 33779-33789]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-11567]
[[Page 33779]]
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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 May 24, 2007, to June 6, 2007. The last
biweekly notice was published on June 5, 2007 (72 FR 31097).
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
[[Page 33780]]
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].
Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-317, Calvert
Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland
Date of amendment request: May 10, 2007.
Description of amendment request: In 2004, the Nuclear Regulatory
Commission (NRC) imposed a license condition that requires the
submission of a coupon surveillance program for the Unit 1 Spent Fuel
Pool (SFP) racks. The coupon surveillance program is necessary to
support an approved license amendment which established acceptable
boron concentrations in the Unit 1 SFP.
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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed surveillance program supports evaluation of
degradation of the neutron absorbing material in the Unit 1 Spent
Fuel Pool (SFP). The function of the neutron absorbing material is
to provide one means of maintaining criticality safety of the
nuclear fuel stored in the SFP.
The postulated accidents for the SFP are basically five types;
(1) dropped fuel assembly on top of the storage rack, (2) a
misloading accident, (3) an abnormal location of a fuel assembly,
(4) loss-of-normal cooling to the SFP, and (5) dilution of boron in
the SFP water.
The proposed change in the coupon surveillance program for the
Unit 1 SFP racks does not affect any of these previously evaluated
accidents. The coupon trees have been evaluated as required by our
plant modifications program and have been determined to have no
effect on accidents in the SFP.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed surveillance program supports evaluation of
degradation of the neutron absorbing material in the Unit 1 SFP. The
function of the neutron absorbing material is to provide one means
of maintaining criticality safety of the nuclear fuel stored in the
SFP.
The coupon trees have been evaluated as required by our plant
modifications program and do not create the possibility of a new or
different kind of accident in the SFP. The surveillance coupons have
existed in the SFP since the Unit 1 SFP racks were installed. The
form and function of the surveillance coupon trees is not changed
because of the need to change the coupon surveillance program. The
interaction of the coupons with the spent fuel racks and the SFP is
not changed due to the proposed surveillance program change.
The proposed change will not result in any other change in the
plant configuration or equipment design. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed coupon surveillance program supports evaluation of
degradation of the neutron absorbing material in the Unit 1 SFP. The
function of the neutron absorbing material is to provide one means
of maintaining criticality safety of the nuclear fuel stored in the
SFP. Evaluation of the coupons as part of an ongoing surveillance
program provides assurance that the fuel will remain subcritical
under all postulated conditions.
Therefore, the proposed 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 proposed to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Generation Group, LLC, 750 East Pratt Street,
17th floor, Baltimore, MD 21202.
NRC Branch Chief: Mark G. Kowal.
[[Page 33781]]
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: May 2, 2007.
Description of amendments request: The proposed amendment would
modify Technical Specification (TS) requirements for unavailable
barriers by adding Limiting Condition for Operation (LCO) 3.0.9. The
changes are consistent with the Nuclear Regulatory Commission approved
Technical Specification Task Force (TSTF)-427, Revision 2. The
availability of this TS improvement was published in the Federal
Register on October 3, 2006 (71 FR 58444) as part of the consolidated
line item improvement process.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The Proposed Change Does Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows a delay time for entering a supported
system technical specification (TS) when the inoperability is due
solely to an unavailable barrier if risk is assessed and managed.
The postulated initiating events which may require a functional
barrier are limited to those with low frequencies of occurrence, and
the overall TS system safety function would still be available for
the majority of anticipated challenges. Therefore, the probability
of an accident previously evaluated is not significantly increased,
if at all. The consequences of an accident while relying on the
allowance provided by proposed LCO 3.0.9 are no different than the
consequences of an accident while relying on the TS required actions
in effect without the allowance provided by proposed LCO 3.0.9.
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.
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).
Allowing delay times for entering supported system TS when
inoperability is due solely to an unavailable barrier, if risk is
assessed and managed, 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 will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
3. The Proposed Change Does Not Involve a Significant Reduction
in the Margin of Safety
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an unavailable
barrier, if risk is assessed and managed. The postulated initiating
events which may require a functional barrier are limited to those
with low frequencies of occurrence, and the overall TS system safety
function would still be available for the majority of anticipated
challenges. The risk impact of the proposed TS changes was assessed
following the three-tiered approach recommended in Regulatory Guide
1.177. A bounding risk assessment was performed to justify the
proposed TS changes. This application of LCO 3.0.9 is predicated
upon the licensee's performance of a risk assessment and the
management of plant risk. The net change to the margin of safety is
insignificant as indicated by the anticipated low levels of
associated risk (ICCDP and ICLERP) as shown in Table 1 of Section
3.1.1 in the Safety Evaluation (71 FR 58449). 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
amendments request involves no significant hazards consideration.
Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Generation Group, LLC, 750 East Pratt Street,
17th floor, Baltimore, MD 21202.
NRC Branch Chief: Mark G. Kowal.
Entergy Nuclear Operations, Inc., Docket No. 50-003, Indian Point, Unit
1, Buchanan, New York
Date of application for amendment: February 22, 2007.
Description of amendment request: The proposed amendment would
enable the licensee to make changes to the Final Safety Analysis Report
(FSAR) to reflect use of the non-single-failure-proof Fuel Handling
Building (FHB) 75 ton crane for dry spent fuel cask handling
operations.
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:
i. Will operation of the facility in accordance with this
proposed amendment involve a significant increase in the probability
or consequences of an accident previously evaluated.
Response: No.
The proposed amendment introduces no new mode of plant
operations and does not affect Structures, Systems, and Components
(SSCs) associated with power production, accident mitigation, or
safe plant shutdown. The SSCs affected by this proposed amendment
are the Indian Point, Unit 1 (IP-1) FHB 75-ton crane, the FHB
concrete structure, the spent fuel storage canister, the spent fuel
transfer cask, and the spent fuel inside the storage canister. A
hypothetical drop of a 30 ton spent fuel shipping cask has been
previously evaluated by the NRC and found to be acceptable based on
the physical arrangement of plant equipment and the fact that the
load path is entirely over concrete floors founded on bedrock or
engineered fill over bedrock. The increased mass of the HI-TRAC
transfer cask containing a fuel-loaded Multi-Purpose Canister
(MPC)consequently results in no change to the basis for the original
cask handling approval.
With this amendment, fewer HI-TRAC casks will be required to be
loaded, lifted, and handled, a planned total of five, than the
previous cask handling effort which involved loading and handling
120 casks. The HI-TRAC cask is within the design capability of the
IP-1 FHB 75 ton crane, therefore the probability of an accident is
not increased.
The new analyses of hypothetical drops of a loaded transfer cask
confirm that there is no release of radioactive material from the
storage canister and no unacceptable damage to the fuel, MPC, or
transfer cask.
The hypothetical drop of a spent fuel canister lid into an open,
fuel-filled canister in the cask loading pool during fuel loading
has been evaluated. [Additionally, the drop of a single spent fuel
assembly into an open fuel-filled canister in the cask loading pool,
due to the potential damage of spent fuel assemblies in the
canister, has been evaluated.] The radiological consequences of
these events are less than 2% of regulatory requirements and are
bounded by the licensing basis of IP-1.
Since the hypothetical drops result in lesser g loads on the
fuel than the design criterion, there is no rearrangement of the
fuel or deformation of the fuel basket in the canister such that a
critical geometry is created.
ii. Will operation of the facility in accordance with this
proposed amendment create the possibility of a new or different kind
of accident from any accident previously evaluated?
Response: No.
The proposed amendment introduces no new mode of plant
operations and does not affect SSCs associated with power
production, accident mitigation, fuel pool cooling, or SAFSTOR
configuration. The SSCs affected by this proposed amendment are the
non-single-failure proof 75 ton crane, structural portions of the
FHB, the spent fuel canister, the spent fuel transfer cask, and the
spent fuel inside the canister.
The design function of the IP-1 FHB 75 ton crane is not changed.
The HI-STORM System
[[Page 33782]]
load drops create the possibility of a new initiator of an accident
previously evaluated (failure of fuel cladding) caused by the
postulated non-mechanistic single failure of a component in the FHB
75 ton crane.
The current licensing basis includes evaluations of the
consequences of a spent fuel cask drop into the cask load pool. The
new initiators include the drop of a fuel transfer cask and a drop
of a spent fuel canister lid into the open, fuel filled canister in
the cask loading pool and a drop of individual assemblies into the
MPC. These new initiators create hypothetical accidents that are
comparable in consequences to and bounded by those previously
evaluated. For the drop of a spent fuel transfer cask, the
consequences of cask impact on facility SSCs are bounded by the
current licensing scenario of a shipping cask drop. That is, there
is no significant damage to the FHB structure or on any SSCs used
for safe storage of spent fuel, and there is no release of
radioactive material. These new analyses of the drop of a loaded
transfer cask confirm that there is no release of radioactive
material from the storage container and no unacceptable damage to
the fuel, MPC, or transfer cask.
For the drop of the spent fuel canister lid, with the maximum
number of assemblies in the canister at 32, or the drop of a single
spent fuel assembly into a fuel-filled canister, doses are
calculated to be less than 2% of regulatory limits. Further the
previously analyzed 100 percent cladding failure of 160 assemblies
bounds the event. There is no rearrangement of the fuel in the
canister such that a critical geometry is created as a result of an
MPC lid drop.
iii. Will operation of the facility in accordance with this
proposed amendment involve a significant reduction in a margin of
safety?
Response: No.
The proposed amendment introduces no new mode of plant
operations and does not affect SSCs associated with spent fuel
storage, spent fuel pool cooling, or the integrity of SSCs in the
SAFSTOR mode. The SSCs affected by this proposed amendment are the
non-single-failure-proof FHB 75 ton crane, structural portions of
the FHB, the spent fuel storage canister, the spent fuel transfer
cask, and the spent fuel inside the canister. This amendment does
not affect the fuel stored in the spent fuel pool or any SSC
associated with safe storage of the fuel. The design function of the
75 ton crane is not changed. The proposed changes to plant
procedures needed to implement dry cask storage do not exceed or
alter a design basis or safety limit associated with accident
mitigation, SAFSTOR, or fuel clad integrity.
This proposed amendment results in a net benefit based upon the
larger capacity cask being used to move and store the fuel (32
assemblies per canister versus two assemblies). All the fuel can be
removed from the spent fuel pool with far fewer cask lifts, welding
evolutions, and storage placement. Because the maximum weight of the
cask loaded with spent fuel is the same as the original design and
tested capacity of the crane, design safety margins for use of the
75 ton crane remain unchanged.
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, Entergy Nuclear Operations,
Inc., 440 Hamilton Avenue, White Plains, NY 10601.
NRC Acting Branch Chief: John Buckley.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: March 15, 2007.
Description of amendment request: The proposed amendment would
change Technical Specification (TS) Section 1.4 and Section 5. Changes
to TS 1.4 would incorporate Nuclear Regulatory Commission (NRC)-
approved Technical Specification Task Force (TSTF) Standard Technical
Specification Changes TSTF-284, ``Add `Met vs. Perform' to
Specification 1.4, Frequency,'' Revision 3, TSTF-485-A, ``Correction
Example 1.4-1,'' Revision 0, and make administrative changes. Changes
to TS Section 5 would incorporate NRC-approved TSTF-258, ``Changes to
Section 5.0, Administrative Controls,'' Revision 4, NRC-approved TSTF-
273, ``[Safety Functions Determination Program] SFDP Clarifications,''
Revision 2, as amended by Westinghouse Owners Group (WOG) editorial
change WOG-ED-23, and make administrative changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed changes are administrative or provide
clarification only.
The proposed changes do not have any impact on the integrity of
any plant system, structure, or component that initiates an analyzed
event. The proposed changes will not alter the operation of, or
otherwise increase the failure probability of any plant equipment
that initiates an analyzed accident. Thus, the probability of any
accident previously evaluated is not significantly increased.
The proposed changes do not affect the ability to mitigate
previously evaluated accidents, and do not affect radiological
assumptions used in the evaluations. The proposed changes do not
change or alter the design criteria for the systems or components
used to mitigate the consequences of any design basis 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. Thus, the radiological consequences of any accident
previously evaluated are not increased.
Therefore, operation of the facility in accordance with 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 proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. 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 amendment does not involve a significant reduction in a
margin of safety. The proposed amendment does not affect any margin
of safety. The proposed amendment does not involve any physical
changes to the plant or manner in which the plant is operated.
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: William Dennis, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY
10601.
NRC Branch Chief: L. Raghavan.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: April 18, 2007.
[[Page 33783]]
Description of amendment request: The proposed amendment would
change Technical Specification (TS) Surveillance Requirement (SR)
3.5.2.9, to support resolution of containment sump issues raised in
Nuclear Regulatory Commission (NRC) Generic Letter (GL) 2004-02,
``Potential Impact of Debris Blockage on Emergency Recirculation during
Design Basis Accidents at Pressurized-Water Reactors.'' The proposed
change to TS SR 3.5.2.9 would make the surveillance consistent with the
plant design following planned modifications to the containment sump.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed changes to TS SR 3.5.2.9 do not have any
impact on the integrity of any plant system, structure, or component
(SSC) that initiates an analyzed event. The proposed changes do not
alter the operation of, or otherwise increase the failure
probability of any plant equipment that initiates an analyzed
accident. Thus, the probability of any accident previously evaluated
is not significantly increased.
The proposed changes do not affect the ability to mitigate
previously evaluated accidents, and do not affect radiological
assumptions used in the evaluations. The proposed changes to TS SR
3.5.2.9 do not change or alter the design criteria for the systems
or components used to mitigate the consequences of any design basis
accident. The proposed amendment does not involve operation of the
required structures, systems, or components in a manner or
configuration different from those previously recognized or
evaluated. The proposed changes to TS SR 3.5.2.9 provide assurance
that the sump flowpath is unrestricted and stays in proper operating
condition. Thus, the radiological consequences of any accident
previously evaluated are not increased.
Therefore, operation of the facility in accordance with 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 proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed amendment to modify TS SR [3.]5.2.9 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 involve a significant reduction
in a margin of safety. The proposed changes do not adversely affect
any plant safety limits, set points, or design parameters. The
proposed changes do not adversely affect the fuel, fuel cladding,
primary coolant system (PCS), or containment integrity. The proposed
TS SR 3.5.2.9 changes ensure that the containment sump is
unrestricted and stays in proper operating condition. The proposed
changes would make the surveillance consistent with the plant design
following planned modifications to the containment sump.
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: William Dennis, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY
10601.
NRC Branch Chief: L. Raghavan.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: April 22, 2007.
Description of amendment request: The proposed amendments would
delete the Unit 2 license condition that requires reporting violations
of other requirements conditions and delete Technical Specifications
(TS) 6.6 for both units that require the NRC be notified of reportable
events pursuant to 10 CFR 50.73. This request also includes an
administrative TS change for both Units by changing references of the
``Topical Quality Assurance Report'' to the ``Quality Assurance Topical
Report.'' The NRC staff issued a notice of opportunity to comment in
the Federal Register on August 29, 2005 (70 FR 51098), on possible
amendments to eliminate the license condition involving reporting of
violations of other requirements (typically in License Condition 2.C)
in the operating license, 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 model for referencing in license
amendment applications in the Federal Register on November 4, 2005 (70
FR 67202).
The licensee affirmed the applicability of the NSHC determination
in its application dated April 22, 2007.
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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change involves the deletion of a reporting
requirement. The change does not affect plant equipment or operating
practices and therefore does not significantly increase 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 is administrative in that it deletes a
reporting requirement. The change does not add new plant equipment,
change existing plant equipment, or affect the operating practices
of the facility. Therefore, the change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change deletes a reporting requirement. The change
does not affect plant equipment or operating practices and therefore
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: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Thomas H. Boyce.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Unit 1 and 2, Berrien County, Michigan
Date of amendment request: May 11, 2007.
Description of amendment request: The proposed amendment would
modify Surveillance Requirement (SR) 3.3.1.18, pertaining to the
reactor trip on turbine trip function, in the Technical
[[Page 33784]]
Specifications (TS). The existing SR requires that the SR be met before
reaching the P-7 interlock (approximately at 10 percent reactor power).
The licensee proposed to change the SR such that the SR will be met
before reaching the P-8 interlock (approximately at 31 percent reactor
power). This proposed change would ensure consistency between the SR
and the mode of applicability for the reactor trip on turbine trip
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 of occurrence or consequences of an accident
previously evaluated?
Response: No.
The proposed change revises a Technical Specification (TS)
[s]urveillance [r]equirement (SR) [f]requency associated with the
reactor trip on turbine trip function to be consistent with the mode
of applicability for the function. The change to the frequency from
prior to exceeding the P-7 interlock to prior to exceeding the P-8
interlock does not create any new credible single failure. The P-7
and P-8 interlocks are not accident initiators. The reactor trip on
turbine trip function is an anticipatory trip, and the safety
analysis does not credit this trip for protecting the reactor core.
The consequences of accidents previously evaluated are unaffected by
this change because no change to any accident mitigation scenario
has resulted and there are no additional challenges to fission
product barrier integrity.
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.
No changes are being made to the plant that would introduce any
new accident causal mechanisms. The proposed change to the interlock
at which the surveillance is performed in support of a reactor trip
on turbine trip does not adversely affect previously identified
accident initiators and does not create any new accident initiators.
The change does not affect how the associated trip function
operates. No new single failures or accident scenarios are created
by the proposed change and the proposed change does not result in
any event previously deemed incredible being made credible.
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.
No safety analyses were changed or modified as a result of the
proposed change in the surveillance frequency. All margins
associated with the current safety analyses acceptance criteria are
unaffected. The current safety analyses remain bounding. The safety
systems credited in the safety analyses will continue to be
available to perform their mitigation functions. The proposed change
does not affect the availability or operability of safety-related
systems and components.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The Nuclear Regulatory Commission (NRC) staff has reviewed the
licensee's analysis and, based on 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: James M. Petro, Jr., Esquire, One Cook
Place, Bridgman, MI 49106.
NRC Acting Branch Chief: Travis L. Tate.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: May 16, 2007.
Description of amendment request: A change is proposed to the
standard technical specifications (STS) (NUREGs 1430 through 1434) and
plant-specific technical specifications (TS), to strengthen TS
requirements regarding control room envelope (CRE) habitability by
changing the action and surveillance requirements associated with the
limiting condition for operation operability requirements for the CRE
emergency ventilation system, and by adding a new TS administrative
controls program on CRE habitability. Accompanying the proposed TS
change are appropriate conforming technical changes to the TS Bases.
The proposed revision to the Bases also includes editorial and
administrative changes to reflect applicable changes to the
corresponding STS Bases, which were made to improve clarity, conform
with the latest information and references, correct factual errors, and
achieve more consistency among the STS NUREGs. The proposed revision to
the TS and associated Bases is consistent with STS as revised by STS
change traveler TS Task Force (TSTF)-448, Revision 3, ``Control Room
Envelope Habitability.''
The proposed amendment would revise the TS to modify requirements
regarding CRE habitability using the Consolidated Line Item Improvement
Process, based on the NRC-approved to TSTF-448, Revision 3. The NRC
staff issued a notice of opportunity for comment in the Federal
Register on October 17, 2006 (71 FR 61075), on possible amendments
adopting TSTF-448, 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 January 17,
2007 (72 FR 2022). The licensee affirmed the applicability of the
following NSHC determination in its application dated May 16, 2007.
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 does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Accident Previously
Evaluated.
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its
[[Page 33785]]
functioning during accident conditions as assumed in the licensing
basis analyses of design basis accident radiological consequences to
CRE occupants. No new or different accidents result from performing
the new surveillance or following the new program. The proposed
change does not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed) or a
significant change in the methods governing normal plant operation.
The proposed change does not alter any safety analysis assumptions
and is consistent with current plant operating practice. Therefore,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition. Therefore, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the 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: James R. Curtiss, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Thomas G. Hiltz.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: May 21, 2007.
Description of amendment request: The proposed amendment would add
Technical Specification (TS) Limiting Condition for Operation (LCO)
3.0.8 to allow a delay time for entering a supported system TS when the
inoperability is due solely to an inoperable snubber, if risk is
assessed and managed consistent with the program in place for complying
with the requirements of 10 CFR 50.65(a)(4).
The NRC staff issued a notice of availability of a model safety
evaluation and model no significant hazards consideration (NSHC)
determination for referencing in license amendment applications in the
Federal Register on May 4, 2005 (70 FR 23252) for model safety
evaluation and November 24, 2004 (69 FR 68420) for NSHC. The licensee
affirmed the applicability of the model NSHC determination in its
application dated May 21, 2007.
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. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. Therefore, the probability
of an accident previously evaluated is not significantly increased,
if at all. The consequences of an accident while relying on
allowance provided by proposed LCO 3.0.8 are no different than the
consequences of an accident while relying on the TS required actions
in effect without the allowance provided by proposed LCO 3.0.8.
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.
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).
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is
assessed and managed, 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 will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in RG 1.177. A bounding risk assessment was performed to
justify the proposed TS changes. This application of LCO 3.0.8 is
predicated upon the licensee's performance of a risk assessment and
the management of plant risk. 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 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.
RC Branch Chief: Evangelos C. Marinos.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: March 14, 2007, as supplemented by
letters dated April 18 and May 9, 2007.
Description of amendment request: The amendment would revise
Technical Specifications (TSs) 3.3.2, ``Engineered Safety Features
Actuation System Instrumentation''; 3.7.2, ``Main Steam Isolation
Valves (MSIVs)''; and 3.7.3, ``Main Feedwater Isolation Valves
(MFIVs).'' The proposed TS changes address the following changes to the
plant and/or plant TSs: (1) The modification of the main steam and
feedwater isolation system (MSFIS), which provides the signal to
actuate the MSIVs and MFIVs, and changes to TS 3.3.2; (2) the
replacement of the MSIVs and MFIVs, and associated actuators; (3) the
addition of the main feedwater regulating valves (MFRVs), and
associated MFRV bypass valves, to TS 3.7.3; (4) the relocation of the
MSIV and MFIV isolation times from TSs 3.7.2 and 3.7.3 to the TS Bases;
and (5) the changes to page numbers in the TS Table of Contents.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) [Do] the proposed change[s] involve a significant increase
in the probability or consequences of an accident previously
evaluated?
Response: No.
[[Page 33786]]
Evaluations and/or reanalysis assessing the impact of the
replacement MSFIS, MSIVs and MFIVs and actuators, and the increased
closure time on non-LOCA [non-loss-of-coolant accident] transients;
SBLOCA [small-break LOCA] transients; main steam line break mass and
energy releases inside and outside containment; containment pressure
and temperature response to a postulated main steam line break;
environmental qualification of equipment; and the steam generator
tube rupture transients and associated radiological consequences,
were performed. The increase in closure times and the changes to the
MSFIS, MSIVs, and MFIVs either do not provide an adverse impact or
do not result in accident acceptance criteria being challenged.
The modifications to the MSFIS controls will not affect any
design basis accidents since the logic which currently exists will
continue to be performed. The replacement controls are functionally
the same as the current system since the same logic functions are
performed, the same inputs received, and the same outputs produced.
The replacement of the MSFIS controls, replacement of the MSIV
and MFIVs, and replacement of the electro-hydraulic actuators with
system-medium actuators [with the longer closure time] will not
result in a significant increase in the probability or consequence
of an accident previously evaluated. [The replacement equipment for
the MSFIS, MSIVs, and MFIVs does not reduce the reliability of the
existing equipment being replaced.]
The relocation of the specific isolation times from the TSs to
the TS Bases does not impact the design safety function of the
valves to close. The TS requirements continue to provide the same
level of assurance as before that the MSIVs and MFIVs are capable of
performing their intended safety function. The addition of the MFRVs
and MFRV bypass valves and extending the Completion Time for one or
more MFIVs inoperable, is not an accident initiator and does not
change the probability that an accident will occur. The increase in
time that the MFIV is unavailable is small and the probability of an
event occurring during this time period which would require
isolation of the flow path is low. The redundancy provided by the
MFRVs and MFRV bypass valves, which have the same actuation signals,
provides adequate assurance that automatic feedwater isolation will
occur.
Based on all of the above, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously analyzed.
(2) [Do] the proposed change[s] create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The increase in MSIV and MFIV closure time as a result of the
replacement of the MSFIS controls, MSIVs and MFIVs and associated
actuators, will not prevent the Main Steam System, Main Feedwater
System, or Auxiliary Feedwater System from performing their safety
functions. The increased closure time will not affect the normal
method of plant operation. No new accident scenarios, transient
precursors, failure mechanisms, or limiting single failures are
introduced with the proposed modifications and increased closure
times. Although the modification does alter the design of the MSFIS
and MSIV and MFIV actuators, it does not prevent the systems,
subsystems, and components from performing their safety functions.
[The replacement equipment for the MSFIS, MSIVs, and MFIVs are not
initiators of accidents.]
The relocation of the specific isolation times from the TSs to
the TS Bases and the addition of the MFRVs and MFRV bypass valves
and extending the Completion Time for one or more MFIVs inoperable
does not affect the assumptions of any accident analysis or the
OPERABILITY of plant equipment.
Therefore, the proposed change[s] [do] not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) [Do] the proposed change[s] involve a significant reduction
in a margin of safety?
Response: No.
The replacement of the MSFIS controls, replacement of the MSIVs
and MFIVs and associated actuators and resulting increased closure
time, does not affect the manner in which safety limits or limiting
safety system settings are determined, nor will there be any adverse
effect on those plant systems necessary to assure the accomplishment
of protection functions. There will be no significant impact on the
overpower limit, departure from nucleate boiling ratio limits, heat
flux hot channel factor, nuclear enthalpy rise hot channel factor,
LOCA peak cladding temperature, peak local density, or any other
margin of safety. The radiological dose consequence acceptance
criteria listed in the Standard Review Plan will continue to be met.
Therefore, the proposed change[s] [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
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Thomas G. Hiltz.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Florida Power and Light Company, Docket No. 50-250, Turkey Point Plant
Unit 3, Miami-Dade County, Florida
Date of application for amendments: May 17, 2007.
Description of amendments request: The proposed amendment would
allow the use of an alternate method of determining rod position for a
control rod with inoperable rod position indication.
Date of publication of individual notice in the Federal Register:
May 24, 2007 (72 FR 29186).
Expiration date of individual notice: June 25, 2007 (Public
comments) and July 23, 2007 (Hearing requests).
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these 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
[[Page 33787]]
made a determination based on that assessment, it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit No. 1, DeWitt County, Illinois
Date of application for amendment: June 30, 2006.
Brief description of amendment: The amendment revises the note
preceding Technical Specification Surveillance Requirement 3.4.6.1 to
be consistent with the wording in NUREG-1434, ``Standard Technical
Specifications for General Electric Plants, BWR/6,'' Revision 3.
Specifically, the note will be revised to read, ``Not required to be
performed in MODE 3.''
Date of issuance: May 24, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 176.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: August 15, 2006 (71 FR
46930) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 24, 2007.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit Nos. 2, New London County, Connecticut
Date of application for amendments: June 13, 2006, as supplemented
by letter dated March 6, 2007.
Brief description of amendments: The amendment revised the
Millstone Power Station, Unit No. 2 (MPS2) Technical Specifications to
modify the MPS2 licensing basis in the area of radiological dose
analysis for design-basis accidents using the alternative source term
permitted by Title 10 of the Code of Federal Regulations 50.67,
``Accident source term''. Additionally, the amendment revises the MPS2
Technical Specifications consistent with the amended licensing-basis.
Date of issuance: May 31, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No: 298.
Facility Operating License Nos. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 29, 2006 (71 FR
51226). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 31, 2007.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336 and 50-423,
Millstone Power Station, Unit Nos. 2 and 3, New London County,
Connecticut
Date of application for amendments: May 31, 2006, as supplemented
by letters dated February 14 and April 26, 2005.
Brief description of amendments: The amendments revised the
Millstone Power Station, Unit Nos. 2 and 3 Technical Specifications
(TSs) related to steam generator (SG) tube integrity. Specifically, the
amendment revises the SG tube surveillance program consistent with the
Nuclear Regulatory Commission-approved TS Task Force (TSTF) Standard TS
Change Traveler, TSTF-449, ``Steam Generator Tube Integrity,'' Revision
4. TSTF-449 is part of the consolidated line item improvement process.
Date of issuance: May 31, 2007
Effective date: As of the date of issuance and shall be implemented
within 180 days.
Amendment Nos: 299 and 238
Facility Operating License Nos. DPR-65 and NPF-49: Amendments
revised the TSs.
Date of initial notice in Federal Register: December 19, 2006 (71
FR 75992).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 31, 2007. No significant hazards
consideration comments received: No.
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 application for amendment: February 8, 2007.
Brief description of amendment: The amendment modified Grand Gulf
Nuclear Station, Unit 1 (GGNS) technical specifications (TSs)
requirements for MODE change limitations in Limiting Condition of
Operation (LCO) 3.0.4 and Surveillance Requirement (SR) 3.0.4. The TS
changes are consistent with Revision 9 of NRC-approved Industry TS Task
Force (TSTF) Standard TS Change Traveler, TSTF-359, ``Increase
Flexibility in MODE Restraints.'' In addition, the amendment also
changed TS Section 1.4, ``Frequency,'' Example 1.4-1, ``Surveillance
Requirements,'' to accurately reflect the changes made by TSTF-359,
which is consistent with NRC-approved TSTF-485, Revision 0, ``Correct
Example 1.4-1.''
Date of issuance: May 30, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 175.
Facility Operating License No. NPF-29: The amendment revises the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: March 27, 2007 (72 FR
14304).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 30, 2007.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of application for amendment: May 25, 2006, as supplemented
January 22, and April 16, 2007.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) consistent with the NRC-approved Revision 4 to
Technical Specification Task Force (TSTF) Standard TS Change Traveler,
TSTF-449, ``Steam Generator Tube Integrity.''
Date of Issuance: May 29, 2007.
Effective Date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 147.
Renewed Facility Operating License No. NPF-16: Amendment revised
the TSs.
[[Page 33788]]
Date of initial notice in Federal Register: July 18, 2006 (71 FR
40747). The January 22, and April 16, 2007, supplements did not affect
the original proposed no significant hazards determination, or expand
the scope of the request as noticed in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 29, 2007.
No significant hazards consideration comments received: No.
GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear
Station, Unit 2, Dauphin County, Pennsylvania
Date of amendment request: December 13, 2006.
Brief description of amendment: The amendment deletes Technical
Specification 6.8.1.3, which provided the requirement for submittal of
the annual occupational radiation exposure report.
Date of issuance: May 25, 2007.
Effective date: May 25, 2007.
Amendment No.: 62.
Possession Only License No. DPR-73: The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 13, 2007 (72
FR 6780)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation Report, dated May 25, 2007.
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 21, 2005, supplemented
by letters dated June 16, August 31, September 29, and October 30,
2006, March 15, and May 10, 2007.
Brief description of amendments: The amendments extend the Required
Action Completion Times (CT) specified in technical specification (TS)
3.8.1, ``AC Sources--Operating,'' to restore an inoperable emergency
diesel generator (EDG) to operable status from the current 7 days to 14
days. Specifically, the proposed changes would revise the current 7-day
CT specified in TS 3.8.1 Required Action B.4 to allow 14 days to
restore an inoperable EDG to operable status.
Date of issuance: May 30, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 178 and 168.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 3, 2006 (71 FR
151).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 30, 2007.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: December 20, 2006.
Brief description of amendment: The amendment deleted the Technical
Specification requirements associated with the hydrogen purge system.
The change is consistent with revisions of 10 CFR 50.44, ``Combustible
gas control for nuclear power reactors,'' that became effective on
October 16, 2003. This operating license improvement was made available
by the U.S. Nuclear Regulatory Commission on September 25, 2003 (68 FR
55416) as part of the consolidated line item improvement process
(CLIIP).
Date of issuance: June 6, 2007.
Effective date: As of its date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment No.: 250.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 30, 2007 (72 FR
4309)
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated June 6, 2007.
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: November 7, 2006.
Brief description of amendments: The amendments revise TS 3.7.1,
``Main Steam Safety Valves,'' operability requirements and Linear Power
Level High Trip setpoints.
Date of issuance: June 5, 2007.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment Nos.: Unit 2-212; Unit 3-204.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: December 19, 2006 (71
FR 75999).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 5, 2007.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-328, Sequoyah Nuclear Plant,
Unit 2, Hamilton County, Tennessee.
Date of application for amendments: February 15, 2006, as
supplemented August 7, 2006, August 30, 2006, November 30, 2006, and
April 2, 2007.
Brief description of amendments: The amendment revises the existing
steam generator (SG) tube surveillance program through technical
specification (TS) changes modeled after TS Task Force (TSTF) traveler
TSTF-449, Revision 4, ``Steam Generator Tube Integrity,'' and the model
safety evaluation prepared by the NRC and published in the Federal
Register on March 2, 2005 (70 FR 10298). The amendment includes changes
to the definition of leakage, changes to the primary-to-secondary
leakage requirements, changes to the SG tube surveillance program,
changes to the SG reporting requirements, and associated changes to the
TS Bases.
The amendment also deletes condition 2.C(8)(b) of Facility
Operating License No. DPR-79.
This license condition references previous commitments for SG
inspection that are bounded by the above TS changes.
Date of issuance: May 22, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 305.
Facility Operating License No. DPR-79: Amendment revised the
license and technical specifications.
Date of initial notice in Federal Register: March 28, 2006 (71 FR
15488).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 22, 2007.
No significant hazards consideration comments received: No.
[[Page 33789]]
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: January 31, 2006, as
supplemented on February 23, June 21, and July 28, 2006.
Brief Description of amendments: These amendments revised the
Technical Specifications to incorporate the changes to the operation of
the containment, as discussed in Generic Letter 2004-02, ``Potential
Impact of Debris Blockage on Emergency Recirculation During Design-
Basis Accidents at Pressurized-Water Reactor,'' dated September 13,
2004.
Date of issuance: October 12, 2006.
Effective date: Unit 1 (fall 2007 refueling outage) and Unit 2
(fall 2006 refueling outage).
Amendment Nos.: 250 and 249.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments changed the license and the technical specifications.
Date of initial notice in Federal Register: March 14, 2006 (71 FR
13182).
The February 23, June 21, and July 28, 2006, supplements contained
clarifying information only and did not change the initial proposed no
significant hazards consideration determination or expand the scope of
the initial application.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 12, 2006.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 11th day of June 2007.
For The Nuclear Regulatory Commission.
Timothy J. McGinty,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. E7-11567 Filed 6-18-07; 8:45 am]
BILLING CODE 7590-01-P