[Federal Register Volume 70, Number 59 (Tuesday, March 29, 2005)]
[Notices]
[Pages 15940-15955]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E5-1343]
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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 March 4, 2005, through March 17, 2005. The
last biweekly notice was published on March 15, 2005 (70 FR 12743).
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, Rules and
Directives 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
[[Page 15941]]
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].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: February 24, 2005.
Description of amendment request: The licensee proposed to revise
Table 3.1.1, ``Protective Instrumentation Requirements,'' of the
Technical Specifications to clarify the conditions under which the
reactor building closed cooling water (RBCCW) pumps and the service
water (SW) pumps will trip during a loss-of-coolant accident (LOCA).
The revised wording would state that the RBCCW and SW pumps will trip
during a LOCA only if offsite power is unavailable. The licensee also
proposed to editorially move a footnote on page 3.6-1 to its correct
place on page 3.6-2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed revision to Technical Specification (TS) Table
3.1.1 to clarify the tripping of the Service Water (SW) and Reactor
Building Closed Cooling Water (RBCCW) pumps documents the as-built
controls for these loads. Amendment No. 42 to the Oyster Creek
Licensing Application concluded that these pumps are not required to
perform any functions related to safe plant shutdown. During a loss
of coolant accident (LOCA) condition, with offsite power available,
the plant electrical busses have enough capacity and capability to
supply the SW and RBCCW pumps. This proposed change is an
administrative change only, and is being made to align the Oyster
Creek Technical Specifications with the design of the plant. No
physical changes are being made to the plant. Also, the footnote on
TS page 3.6-1 would be relocated to TS page 3.6-2 to appear on the
same TS page as the Specification to which it applies. The proposed
changes do not alter the physical design or operational procedures
associated with any plant structure, system, or component.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(2) Does the proposed change create the possibility of a new or
different kind of
[[Page 15942]]
accident from any accident previously evaluated?
Response: No.
The proposed revision to Technical Specification Table 3.1.1 to
clarify the tripping of the SW and RBCCW pumps documents as-built
controls for these loads. These pumps provide cooling to various
non-safety related plant equipment. Following a LOCA condition, with
offsite power available, these pumps will help in removing plant
heat loads. This clarification that the SW and RBCCW pumps do not
trip during a LOCA, with offsite power available, does not affect
the Emergency Diesel Generator time delayed loading sequence. The
relocation of the footnote applicable to Specification 3.6.A.4.1 is
editorial in nature and has no impact on any accident previously
evaluated. Accordingly, the proposed changes do not introduce any
new accident initiators, nor do they reduce or adversely affect the
capabilities of any plant structure or system in the performance of
their safety function.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
The proposed revision to Technical Specification Table 3.1.1 to
clarify the tripping of the SW and RBCCW pumps documents as-built
controls for these loads. The NRC Safety Evaluation Report (SER) for
Amendment 42 to the Oyster Creek Licensing Application concluded
that it is acceptable to automatically trip the SW and RBCCW pumps
during a loss of coolant accident. The NRC SER for Technical
Specification Amendment 60 concluded that the immediate tripping of
the RBCCW pump and the time delayed tripping of the SW pumps during
a LOCA was also acceptable. The clarification that the SW and RBCCW
pumps do not trip during a loss of coolant accident when offsite
power is available does not reduce any margin of safety because
these pumps are not required to mitigate the consequences of any
postulated accident. The relocation of the footnote applicable to
Specification 3.6.A.4.1 is editorial in nature and has no impact on
any accident margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LCC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Richard J. Laufer.
Dominion Nuclear Connecticut Inc., et al., Docket Nos. 50-336 and 50-
423, Millstone Power Station, Unit Nos. 2 and 3, New London County,
Connecticut
Date of amendment request: February 25, 2005.
Description of amendment request: The proposed changes would amend
Operating License DPR-65 for Millstone Power Station, Unit No. 2 (MPS2)
and Operating License NPF-49 for Millstone Power Station, Unit No. 3
(MPS3) by incorporating certain administrative changes into the MPS2
and MPS3 Technical Specifications (TSs).
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 changes are administrative in nature and do not
alter any of the requirements of the affected TS[s]. The proposed
changes do not modify any plant equipment and do not impact any
failure modes that could lead to an accident. Additionally, the
proposed changes have no effect on the consequence of any analyzed
accident since the changes do not affect any equipment related to
accident mitigation. Based on this discussion, the proposed
amendment does not increase the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes are administrative in nature. They do not
modify any plant equipment and there is no impact on the capability
of the existing equipment to perform their intended functions. No
system setpoints are being modified and no changes are being made to
the method in which plant operations are conducted. No new failure
modes are introduced by the proposed changes. The proposed amendment
does not introduce accident initiators or malfunctions that would
cause a new or different kind of accident. 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.
These changes are administrative in nature and do not alter any
of the requirements of the affected TS[s]. The proposed changes do
not affect any of the assumptions used in the accident analysis, nor
do they affect any operability requirements for equipment important
to plant safety. Therefore, the proposed changes will not result in
a significant reduction in the margin of safety as defined in the
bases for technical specifications covered in this license amendment
request.
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, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
NRC Section Chief: Darrell J. Roberts.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: March 8, 2005.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 5.5.13, Primary Containment Leakage
Rate Testing Program, for the Integrated Leak Rate Testing (ILRT)
program to add an exception to the commitment to follow the guidelines
of Regulatory Guide 1.163, ``Performance-Based Containment Leak-Test
Program.'' The effect of this request would be a one-time extension of
the interval since the last ILRT from 15 years to 15 years and 4 months
(i.e., from August 2007 to December 2007).
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. 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 to TS 5.5.13 allows a one-time extension
to the current interval for the ILRT. The current interval of
fifteen years, based on past performance, would be extended on a
one-time basis to 15-years and 4 months from the date of the last
test. The proposed extension to the ILRT cannot increase the
probability of an accident since there are no design or operating
changes involved and the test is not an accident initiator. The
proposed extension of the test interval does not involve a
significant increase in the consequences since analysis has shown
that, the proposed extension of the ILRT and DWBT [Drywell Bypass
Test] frequency has a minimal impact on plant risk. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
[[Page 15943]]
2. 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 extension to the interval for the ILRT does not
involve any design or operational changes that could lead to a new
or different kind of accident from any accidents previously
evaluated. The tests are not being modified, but are only being
performed after a longer interval. The proposed change does 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. Therefore, the proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Will operation of the facility in accordance with this
proposed amendment involve a significant reduction in a margin of
safety?
Response: No.
An evaluation of extending the ILRT DWBT surveillance frequency
from once in 10 years to once in 15 years and 4 months has been
performed using methodologies based on the approved ILRT
methodologies. This evaluation assumed that the DWBT frequency was
being adjusted in conjunction with the ILRT frequency. This analysis
used realistic, but still conservative, assumptions with regard to
developing the frequency of leakage classes associated with the ILRT
and DWBT. The results from this conservative analysis indicates that
the proposed extension of the ILRT frequency has a minimal impact on
plant risk and therefore, the proposed change does not involve a
significant reduction in a margin of safety.
Based on the above, Entergy concludes that the proposed
amendment(s) present no significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Allen G. Howe.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: December 20, 2004.
Description of amendment request: Entergy Operations, Inc. is
proposing that the Arkansas Nuclear One Unit 2 (ANO-2) Facility
Operating License be amended to revise the requirements for ensuring
containment structural integrity. The proposed changes modify the
Containment Structural Integrity Technical Specification (TS) 3.6.1.5
to delete the existing Surveillance Requirements (SR) and add a new SR
to verify containment structural integrity in accordance with a new
Containment Tendon Surveillance Program. A new Containment Tendon
Surveillance Program is added to TS 6.5.6 and a new reporting
requirement is being added to TS 6.6.6. The proposed changes are
generally consistent with NUREG 1432, ``Standard Technical
Specifications Combustion Engineering Plants,'' Revision 3. This
request for amendment also contains proposed administrative changes
related to page numbering.
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:
Criterion 1--Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The containment building is not considered to be the initiator
of any accident previously evaluated, but serves to mitigate
accidents that could allow a release to the environment. The
proposed TS change will provide for containment tendon inspections
as required by 10 CFR 50.55a and prevent or inhibit release from the
containment building as designed. Through appropriate inspections
and implementation of corrective actions for any degradation
discovered during the inspections that might lead to containment
structural failures, the probability or consequences of accidents
will not be increased.
Criterion 2--Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed change does not change the design, configuration,
or method of operation of the plant. By implementing corrective
actions for any degradation discovered during the required
inspections of the containment, the possibility of a new or
different kind of accident will not be created. Implementation of
the requirements of Subsection IWL of the ASME code [American
Society of Mechanical Engineers Boiler and Pressure Vessel Code] and
those of 10 CFR 50.55a(b)(2) provide an equally acceptable
containment inspection program.
Criterion 3--Does Not Involve a Significant Reduction in the
Margin of Safety.
The proposed change to incorporate the applicable requirements
of Subsection IWL of the ASME Code and of 10 CFR 50.55a(b)(2) into
the ANO-2 containment inspection program has no impact on any safety
analysis assumptions. The addition of structural integrity
requirements to ANO-2 TS Specification 3.6.1.5 imposes consistent
requirements with those previously specified in the ANO-2 TSs. The
requirements of ASME IWL are more restrictive than those currently
provided in the existing ANO-2 technical specifications. As a
result, the margin of safety is not reduced by the proposed change.
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 are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Allen G. Howe.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: February 25, 2005.
Description of amendment requests: The proposed amendments would
modify the Technical Specifications by revising the near-end-of-life
Moderator Temperature Coefficient (MTC) Surveillance Requirement by
placing a set of conditions on core performance, which, if met, would
allow conditional exemption from the required MTC measurement.
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 probability or consequences of accidents previously
evaluated in the Updated Final Safety Analysis Report (UFSAR) are
unaffected by this proposed change because there is no change to any
equipment response or accident mitigation scenario. 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 new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed change.
The proposed change does not challenge the performance or integrity
of any safety-related system.
Therefore, the proposed change does not create the possibility
of a new or different
[[Page 15944]]
kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety associated with the acceptance criteria of
any accident is unchanged. The proposed change will have no affect
on the availability, operability, or performance of the safety-
related systems and components. A change to a surveillance
requirement is proposed, but the limiting conditions for operation
required by the Technical Specifications (TS) are not changed.
The Technical Specifications Bases are founded in part on the
ability of the regulatory criteria to be satisfied assuming the
limiting conditions for operation are met for the various systems.
Conformance to the regulatory criteria for operation with the
conditional exemption from the near-end of life moderator
temperature coefficient (MTC) measurement is demonstrated and the
regulatory limits are not exceeded. Therefore, the margin of safety
as defined in the TS is not reduced.
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 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 Section Chief: L. Raghavan.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: March 8, 2005.
Description of amendment request: The proposed amendment would
revise Technical Specification 2.1.1.2 for the single recirculation
loop Safety Limit Minimum Critical Power Ratio (SLMCPR) value to
reflect results of a cycle-specific calculation for Cycle 23
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:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident.
Changing the SLMCPR does not increase the probability of an
evaluated accident. The change does not require any physical plant
modifications, physically affect any plant components, or entail
changes in plant operation. Therefore, no individual precursors of
an accident are affected.
The consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits have been established, consistent with NRC
approved methods, to ensure that fuel performance during normal,
transient, and accident conditions is acceptable. The proposed
change conservatively establishes the safety limit for the minimum
critical power ratio for CNS Cycle 23 such that the fuel is
protected during normal operation and during any plant transients or
anticipated operational occurrences.
The proposed change revises the SLMCPR to protect the fuel
during normal operation as well as during any transients or
anticipated operational occurrences. Operational limits Minimum
Critical Power Ratio (MCPR) are established based on the proposed
SLMCPR to ensure that the SLMCPR is not violated during all modes of
operation. This will ensure that the fuel design safety criteria
(i.e., that at least 99.9% of the fuel rods do not experience
transition boiling during normal operation and anticipated
operational occurrences) is met. Since the operability of plant
systems designed to mitigate any consequences of accidents has not
changed, the consequences of an accident previously evaluated are
not expected to increase.
Based on the above, NPPD concludes that the proposed changes do
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors of
that accident. New accident precursors may be created by
modifications of the plant configuration or changes in allowable
modes of operation. The proposed change does not involve any
modifications of the plant configuration or allowable modes of
operation. The proposed change to the SLMCPR assures that safety
criteria are maintained for Cycle 23.
Based on the above, NPPD concludes that the proposed changes do
not create the possibility of a new or different kind of accident
from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The value of the proposed SLMCPR provides a margin of safety by
ensuring that no more than 0.1% of the rods are expected to be in
boiling transition if the MCPR limit is not violated. The proposed
change will ensure the appropriate level of fuel protection is
maintained. Additionally, operational limits are established based
on the proposed SLMCPR to ensure that the SLMCPR is not violated
during all modes of operation. This will ensure that the fuel design
safety criteria (i.e., that at least 99.9% of the fuel rods do not
experience transition boiling during normal operation as well as
anticipated operational occurrences) are met.
Based on the above, NPPD concludes that the proposed changes do
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Allen G. Howe.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of amendment request: February 3, 2005.
Description of amendment request: The proposed amendments would
modify the Technical Specifications (TSs) by revising TS 6.16.b.1,
``Radioactive Effluent Controls Program,'' to be consistent with the
intent of 10 CFR 20 and NUREG-1431, ``Standard Technical Specifications
Westinghouse Plants'' (STS).
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?
NMC [Nuclear Management Company, LLC] Response:
No. Updating the specification to be consistent with 10 CFR 20
and the STS has no impact on plant structures, systems, or
components, does not affect any accident initiators, and does not
change any safety analysis. Therefore, the changes do not involve an
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?
NMC Response:
No. Updating the specification to be consistent with 10 CFR 20
and the STS will not change any equipment, require new equipment to
be installed, or change the way current equipment operates. No
credible new failure mechanisms, malfunctions, or
[[Page 15945]]
accident initiators are created by the proposed changes. Therefore,
the changes do 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?
NMC Response:
No. Updating the specification to be consistent with 10 CFR 20
and the STS has no impact on inputs to the safety analysis or to
automatic plant actions. It also does not impact plant equipment or
operation. Therefore, the change does not reduce 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: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: October 15, 2004.
Description of amendment request: The proposed amendment revises TS
5.5.6, ``Reactor Coolant Pump Flywheel Inspection Program,'' to extend
the allowable inspection interval to 20 years.
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 October 22, 2003 (68 FR 60422). The licensee
affirmed the applicability of the model NSHC determination in its
application dated October 15, 2004.
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 to the RCP flywheel examination frequency
does not change the response of the plant to any accidents. The RCP
will remain highly reliable and the proposed change will not result
in a significant increase in the risk of plant operation. Given the
extremely low failure probabilities for the RCP motor flywheel
during normal and accident conditions, the extremely low probability
of a loss-of-coolant accident (LOCA) with loss of offsite power
(LOOP), and assuming a conditional core damage probability (CCDP) of
1.0 (complete failure of safety systems), the core damage frequency
(CDF) and change in risk would still not exceed the NRC's acceptance
guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per
year). Moreover, considering the uncertainties involved in this
evaluation, the risk associated with the postulated failure of an
RCP motor flywheel is significantly low. Even if all four RCP motor
flywheels are considered in the bounding plant configuration case,
the risk is still acceptably low.
The proposed change does not adversely affect accident
initiators or precursors, nor alter the design assumptions,
conditions, or configuration of the facility, or the manner in which
the plant is operated and maintained; alter or prevent the ability
of structures, systems, components (SSCs) from performing their
intended function to mitigate the consequences of an initiating
event within the assumed acceptance limits; or affect the source
term, containment isolation, or radiological release assumptions
used in evaluating the radiological consequences of an accident
previously evaluated. Further, the proposed change does not increase
the type or amount of radioactive effluent that may be released
offsite, nor significantly increase individual or cumulative
occupational/public radiation exposure. The proposed change is
consistent with the safety analysis assumptions and resultant
consequences. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
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 in flywheel inspection frequency does not
involve any change in the design or operation of the RCP. Nor does
the change to examination frequency affect any existing accident
scenarios, or create any new or different accident scenarios.
Further, the change does not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or alter the methods governing normal plant operation. In
addition, the change does not impose any new or different
requirements or eliminate any existing requirements, and does not
alter any assumptions made in the safety analysis. The proposed
change is consistent with the safety analysis assumptions and
current plant operating practice. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3--The proposed change does not involve a significant
reduction in a margin of safety.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not impacted by this change. The proposed change will not result
in plant operation in a configuration outside of the design basis.
The calculated impact on risk is insignificant and meets the
acceptance criteria contained in RG 1.174. There are no significant
mechanisms for inservice degradation of the RCP flywheel. Therefore,
the proposed 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: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Section Chief: L. Raghavan.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: September 27, 2004.
Description of amendment request: The proposed amendment would
revise the reactor coolant pump (RCP) flywheel inspection surveillance
requirements to extend the allowable inspection interval to 20 years.
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 October 22, 2003 (68 FR 60422). The licensee
affirmed the applicability of the model NSHC determination in its
application dated September 27, 2004.
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 to the RCP flywheel examination frequency
does not change the response of the plant to any accidents. The RCP
will remain highly reliable and the proposed change will not result
in a significant increase in the risk of plant operation. Given the
extremely low failure probabilities for the RCP motor flywheel
during normal and accident conditions, the extremely low probability
of a loss-of-coolant accident (LOCA) with loss of offsite power
(LOOP), and assuming a conditional core damage probability (CCDP) of
1.0 (complete failure of safety systems), the core damage frequency
(CDF) and change in risk would still not exceed the NRC's acceptance
guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per
year). Moreover, considering the uncertainties involved in this
evaluation, the risk associated with the postulated failure of an
RCP motor flywheel is significantly low. Even if all four RCP motor
flywheels are considered in the
[[Page 15946]]
bounding plant configuration case, the risk is still acceptably low.
The proposed change does not adversely affect accident
initiators or precursors, nor alter the design assumptions,
conditions, or configuration of the facility, or the manner in which
the plant is operated and maintained; alter or prevent the ability
of structures, systems, components (SSCs) from performing their
intended function to mitigate the consequences of an initiating
event within the assumed acceptance limits; or affect the source
term, containment isolation, or radiological release assumptions
used in evaluating the radiological consequences of an accident
previously evaluated. Further, the proposed change does not increase
the type or amount of radioactive effluent that may be released
offsite, nor significantly increase individual or cumulative
occupational/public radiation exposure. The proposed change is
consistent with the safety analysis assumptions and resultant
consequences. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
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 in flywheel inspection frequency does not
involve any change in the design or operation of the RCP. Nor does
the change to examination frequency affect any existing accident
scenarios, or create any new or different accident scenarios.
Further, the change does not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or alter the methods governing normal plant operation. In
addition, the change does not impose any new or different
requirements or eliminate any existing requirements, and does not
alter any assumptions made in the safety analysis. The proposed
change is consistent with the safety analysis assumptions and
current plant operating practice. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3--The proposed change does not involve a significant
reduction in a margin of safety.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not impacted by this change. The proposed change will not result
in plant operation in a configuration outside of the design basis.
The calculated impact on risk is insignificant and meets the
acceptance criteria contained in RG 1.174. There are no significant
mechanisms for inservice degradation of the RCP flywheel. Therefore,
the proposed 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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: Darrell J. Roberts.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Docket No. 50-354, Hope Creek Generating Station, Salem County, New
Jersey Date of amendment request: January 11, 2005. Description of
amendment request: The proposed amendment would delete the Technical
Specification (TS) requirements to submit monthly operating reports and
occupational radiation exposure reports.
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in licensing amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated January 11, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating report
of shutdown experience and operating statistics if the equivalent
data is submitted using an industry electronic database. It also
eliminates the TS reporting requirement for an annual occupational
radiation exposure report, which provides information beyond that
specified in NRC regulations. The proposed change involves no
changes to plant systems or accident analyses. As such, the change
is administrative in nature and does not affect initiators of
analyzed events or assumed mitigation of accidents or transients.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
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.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: Darrell J. Roberts.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: February 15, 2005.
Description of amendment request: The proposed amendment will
revise the Salem, Unit Nos. 1 and 2 Technical Specifications to reflect
the deletion of Reactor Coolant System (RCS) volume from design
features Section 5.4.2. This design feature information will continue
to be maintained in the plant's updated final safety analysis report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated.
Response: No.
The proposed change to remove this information from T/S
[technical specifications] does not affect any accident initiators
or precursors. Elimination of the RCS volume information from the T/
S does not change the methods for plant operation or actions to be
taken in the event of an accident. The quantity of radioactive
material available for release in the event of an accident is not
increased.
Barriers to release of radioactive material are not eliminated
or otherwise changed. More detailed RCS component and piping volume
information is included in the Salem UFSAR [updated final safety
analysis report], and changes to that information would be evaluated
prior to implementation in accordance with 10 CFR 50.59.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of accidents 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.
[[Page 15947]]
The deletion of the RCS volume information from the T/S does not
change the methods of plant operation or modify plant systems,
structures, or components. No new methods of plant operation are
created. As such, the proposed change does not affect any accident
initiators or precursors or create new accident initiators or
precursors. More detailed and complete RCS component and piping
volume information is included in the Salem UFSAR, and any changes
to that information would be evaluated prior to implementation in
accordance with 10 CFR 50.59.
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 deletion of the RCS volume information from the T/S does not
affect safety limits or limiting safety system settings. Plant
operational parameters are not affected. The proposed change does
not modify the quantity of radioactive material available for
release in the event of an accident. As such, the change will not
affect any previous safety margin assumptions or conditions. The
actual volume of the RCS is not affected by the change, only the
location of the text describing the volume. More detailed and
complete RCS component and piping volume information is included in
the Salem UFSAR, and any changes to that information would be
evaluated prior to implementation in accordance with 10 CFR 50.59.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: Darrell J. Roberts.
Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco
Nuclear Generating Station, Sacramento County, California
Date of amendment request: January 24, 2005.
Description of amendment request: The proposed license amendment
removes unnecessary and obsolete information from the facility license.
The proposed changes are editorial and administrative in nature and
will remove inappropriate and unnecessary information from the license
given that the facility is permanently shutdown and defueled.
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.
No. The proposed changes are administrative and involve deleting
unnecessary and obsolete information from the facility operating
license. These changes do not affect possible initiating events for
accidents previously evaluated or alter the configuration or
operation of the facility. Safety limits, limiting safety system
settings, and limiting control systems are no longer applicable to
Rancho Seco in the permanently defueled mode, and are therefore not
relevant.
The proposed changes do not affect the boundaries used to
evaluate compliance with liquid or gaseous effluent limits, and have
no impact on plant operations. Therefore, the proposed license
amendment 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 accident from any previously evaluated.
No. As described above, the proposed changes are administrative.
The safety analysis for the facility remains complete and accurate.
There are no physical changes to the facility and the plant
conditions for which the design basis accidents have been evaluated
are still valid.
The operating procedures and emergency procedures are not
affected. The proposed changes do not affect the emergency planning
zone, the boundaries used to evaluate compliance with liquid or
gaseous effluent limits, and have no impact on plant operations.
Consequently, no new failure modes are introduced as the result of
the proposed changes. Therefore, the proposed changes will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
No. As described above, the proposed changes are administrative.
There are no changes to the design or operation of the facility. The
proposed changes do not affect the emergency planning zone, the
boundaries used to evaluate compliance with liquid or gaseous
effluent limits, and have no impact on plant operations.
Accordingly, neither the design basis nor the accident assumptions
in the Defueled Safety Analysis Report (DSAR), nor the Technical
Specification Bases are affected. Therefore, the proposed changes do
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's significant hazards
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: Arlen Orchard, Esq., General Counsel,
Sacramento Municipal Utility District, 6201 S Street, P.O. Box 15830,
Sacramento, CA 95817-1899.
NRC Section Chief: Claudia M. Craig.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of amendment request: August 16, 2004 (TS-433).
Description of amendment request: The proposed amendment extends
the frequency of ``once-per-cycle'' from 18 months to 24 months in
several Technical Specification Surveillance Requirements. This change
will allow the adoption of a 24-month refueling cycle.
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?
No. The proposed amendment changes the surveillance frequency
from 18 months to 24 months for Surveillance Requirements in the
Unit 1 Technical Specification[s] that are normally a function of
the refueling interval. Under certain circumstances, Surveillance
Requirement 3.0.2 would allow a maximum surveillance interval of 30
months for these surveillances. TVA's evaluations have shown that
the reliability of protective instrumentation and equipment will be
preserved for the maximum allowable surveillance interval. The
proposed changes do not involve any change to the design or
functional requirements of plant systems and the surveillance test
methods will be unchanged. The proposed changes will not give rise
to any increase in operating power level, fuel operating limits, or
effluents. The proposed change does not affect any accident
precursors. In addition, the proposed changes will not significantly
increase any radiation levels. Based on the foregoing considerations
and the evaluations completed in accordance with the guidance of
Generic Letter 91-04, it is concluded that 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?
No. The proposed amendment does not require a change to the
plant design, nor the mode of plant operation. The proposed changes
do not create the possibility of any
[[Page 15948]]
new failure mechanisms. No new external threats or release pathways
are created. Therefore, the implementation of the proposed amendment
will not create a possibility for an accident of a new or different
type than those previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
No. The proposed amendment changes the surveillance frequency
from 18 months to 24 months for Surveillance Requirements in the
Unit 1 Technical Specification[s] that are normally a function of
the refueling interval. Under certain circumstances, Surveillance
Requirement 3.0.2 would allow a maximum surveillance interval of 30
months for these surveillances. Although the proposed Technical
Specification changes will result in an increase in the interval
between surveillance tests, the impact on system availability is
small based on other, more frequent testing or redundant systems or
equipment. There is no evidence of any failures that would impact
the availability of the systems. This change does not alter the
existing setpoints, Technical Specification allowable values or
analytical limits. The assumptions in the current safety analyses
are not impacted and the proposed amendment does not reduce a margin
of safety. Therefore, the proposed license amendment 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Michael L. Marshall, Jr.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear
Plant (BFN), Unit 1, Limestone County, Alabama
Date of amendment request: October 12, 2004 (TS-438).
Description of amendment request: The proposed amendment request
changes the frequency requirement for Technical Specification
Surveillance Requirement (SR) 3.6.1.3.8 by allowing a representative
sample (approximately 20 percent) of excess flow check valves (EFCVs)
to be tested every 24 months, so that each EFCV is tested once every
120 months. The current SR requires testing of each EFCV every 24
months.
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?
No. The current EFCV frequency requires that each reactor
instrument line EFCV be tested every 24 months. The EFCVs are
designed to automatically close upon excessive differential pressure
including failure of the down stream piping or instrument and will
reopen when appropriate. This proposed change will allow a reduction
in the number of EFCVs that are verified tested every 24 months, to
approximately 20 percent of the valves each cycle. BFN and industry
operating experience demonstrates high reliability of these valves.
Neither the EFCVs nor their failure is capable of initiating a
previously evaluated accident. Therefore, there is no increase in
the probability of occurrence of an accident previously evaluated.
The instrument lines going to the Reactor Coolant Pressure
boundary with EFCVs installed have flow restricting devices upstream
of the EFCV. The consequences of an unisolable failure of an
instrument line have been previously evaluated and meet the intent
of NRC Safety Guide 11. The offsite exposure has been calculated to
be substantially below the limits of 10 CFR 50.67. The total control
room Total Effective Dose Equivalent (TEDE) doses are less than the
5 REM limit and the offsite TEDE doses are less than 10% of the 25
REM limit. Additionally, coolant lost from such a break is
inconsequential compared to the makeup capabilities of normal and
emergency makeup systems. Although not expected to occur as a result
of this change, the affects of a postulated failure of an EFCV to
isolate and [sic] instrument line break as a result of reduced
testing are bounded by TVA analysis.
Therefore, the proposed change does not involve a significant
increase in the consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
No. The proposed reduction in EFCV test frequency is bounded by
previous evaluation of a line rupture. The proposed change does not
introduce new equipment, which could create a new or different kind
of accident. No new external threats, release pathways, or equipment
failure modes are created. Therefore, the implementation of the
proposed change will not create a possibility for an accident of a
new or different type than those previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
No. The consequences of an unisolable rupture of an instrument
line have been previously evaluated and meet the intent NRC Safety
Guide 11. The proposed change does not involve a significant
reduction in a margin of safety. Therefore, the proposed revised
surveillance frequency does not adversely affect the public health
and safety, and does not involve any significant safety hazards.
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 are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Michael L. Marshall, Jr.
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.
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit No. 3, New London County, Connecticut
Date of amendment request: February 10, 2005.
Brief description of amendment request: The proposed amendment
would extend the allowed outage time for the Emergency Generator Load
Sequencer (Technical Specification 3/4.3.2, Table 3.3-3, Functional
Unit 10) from 6 hours to 12 hours.
Date of publication of individual notice in Federal Register:
February 22, 2005 (70 FR 8641).
Expiration date of individual notice: March 24, 2005 (public
comments) and April 25, 2005 (hearing requests).
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: July 23, 2004, and January 6, 2005.
Brief description of amendment request: The proposed revision would
modify the Technical Specification (TS)
[[Page 15949]]
definition of OPERABILITY with respect to requirements for availability
of normal and emergency power. Additionally, the proposed revision
would modify the required actions for shutdown power TSs.
Date of publication of individual notice in Federal Register: March
1, 2005.
Expiration date of individual notice: March 31, 2005 (public
comments), and May 2, 2005 (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 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) The
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of application for amendment: April 30, 2004.
Brief description of amendment: The amendment modifies requirements
in the Technical Specifications (TS) to adopt the provisions of
Industry/TS Task Force (TSTF) change TSTF-359, ``Increased Flexibility
in Mode Restraints.''
Date of issuance: March 2, 2005.
Effective date: As of the date of issuance and shall be implemented
within 180 days.
Amendment No.: 163.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 26, 2004 (69 FR
62469).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 2, 2005.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket No. 50-324, Brunswick Steam
Electric Plant, Unit 2, Brunswick County, North Carolina
Date of application for amendment: August 16, 2004.
Brief Description of amendment: The amendment adds topical report
NEDE-32906P-A, ``TRACG Application for Anticipated Operational
Occurrences (AOO) Transient Analyses,'' to the documents listed in
Technical Specification 5.6.5 describing the approved methodologies
used to determine the core operating limits.
Date of issuance: March 4, 2005.
Effective date: March 4, 2005.
Amendment No.: 262.
Facility Operating License No DPR-62: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: October 26, 2004 (69 FR
62470).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 4, 2005.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: May 27, 2004.
Brief description of amendments: The amendments revised the
Technical Specifications by eliminating the requirements associated
with hydrogen recombiners and hydrogen monitors.
Date of issuance: March 1, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 219 and 214 .
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 28, 2004 (69
FR 57982).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 1, 2005.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: September 27, 2004.
Brief description of amendment: The amendment eliminated the
technical specification requirements to submit a monthly operating
report and an annual occupational radiation exposure report.
Date of issuance: March 9, 2005.
Effective date: March 9, 2005.
Amendment No.: 190.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 26, 2004 (69 FR
62472).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 9, 2005.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One Unit
No. 2, Pope County, Arkansas
Date of application for amendment: April 15, 2004, as supplemented
January 20, 2005.
Brief Description of amendments: The licensee has proposed to
change the existing reactor coolant system (RCS) cooldown curve to a
single 32 effective full power year pressure/temperature limit curve
that is applicable for cooldowns at a rate of 100 [deg]F/hour or 50
[deg]F in any half-hour step. The licensee's proposed curve is
applicable to RCS
[[Page 15950]]
cold-leg temperatures ranging from 50 [deg]F to 560 [deg]F.
Date of issuance: March 7, 2005.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 256.
Facility Operating License No. NFP-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 11, 2004 (69 FR
26188). The supplemental letter 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 7, 2005.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: April 14, 2004.
Brief description of amendment: The amendment revised the Pilgrim
Nuclear Power Station Technical Specifications (TSs) by adding a new
limiting condition for operation (LCO) 3.0.7 to Section 3.0, ``Limiting
Condition for Operation (LCO) Applicability,'' a new TS Section 3.14,
``Special Operations,'' and a new LCO 3.14.A, ``Inservice Leak and
Hydrostatic Testing Operation,'' to the TSs. These changes permit the
licensee to perform inservice hydrostatic testing and system leakage
pressure testing of the reactor coolant system at temperatures greater
than 212 [deg]F with the reactor shut down.
Date of issuance: March 16, 2005.
Effective Date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 211.
Facility Operating License No. DPR-35: The amendment revised the
TSs.
Date of initial notice in Federal Register: December 21, 2004 (69
FR 76489).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 16, 2005.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: April 30, 2004.
Brief description of amendments: The amendments modify Technical
Specifications (TS) requirements to adopt the provisions of Industry/TS
Task Force (TSTF) change TSTF-359, ``Increased Flexibility in Mode
Restraints.''
Date of issuance: March 10, 2005.
Effective date: As of the date of issuance and shall be implemented
within 180 days.
Amendment Nos.: 212/204/223/218.
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 26, 2004 (69 FR
62474).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 10, 2005.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: June 10, 2004, and supplemented
July 19 and July 21, 2004 and January 21, 2005.
Brief description of amendments: The amendments revise the Quad
Cities Nuclear Power Station Technical Specifications to change the
allowable value and add Surveillance Requirements for the Main Steam
Line Flow-High initiation of Group 1 Primary Containment Isolation
System and Control Room Emergency Ventilation System isolation.
Date of issuance: March 15, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days for Unit 1 and no later than 90 days after the start of
the Unit 2 refueling outage currently scheduled for March 2006 for Unit
2.
Amendment Nos.: 224, 219
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: August 31, 2004 (69 FR
53107). 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 March 15, 2005.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of amendment request: June 1, 2004, as supplemented July 23,
2004, and February 18, 2005.
Description of amendment request: These amendments lowered the
BVPS-2 overpressure protection system enable temperature, allowed one
inoperable residual heat removal loop during surveillance testing,
removed the BVPS-1 list of figures and list of tables from the Index of
the BVPS-1 Technical Specifications (TSs), and made minor changes to
achieve consistency between units and with the Standard TSs for
Westinghouse plants and with some TS Task Force changes.
Date of issuance: March 11, 2004.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment Nos.: 265 and 146.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. February 25, 2005 (70 FR 9391). The notice
provided an opportunity to submit comments on the Commission's proposed
NSHC determination by March 11, 2005. No comments have been received.
The notice also provided an opportunity to request a hearing by April
26, 2005, but indicated that if the Commission makes a final NSHC
determination, any such hearing would take place after issuance of the
amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated March 11, 2005.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: April 13, 2004.
Brief description of amendments: The amendments change the design
basis as described in the Updated Final Safety Analysis Report to allow
the use in control rod drive missile shield structural calculations of
a reinforcing bar (rebar) yield strength value based on measured
material properties, as
[[Page 15951]]
documented in the licensee rebar acceptance tests.
Date of issuance: March 11, 2005.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 286, 268.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the design basis.
Date of initial notice in Federal Register: October 12, 2004 (69 FR
60682).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 11, 2005.
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: September 7, 2004.
Brief description of amendment: The amendment revised Technical
Specification (TS) 5.9.5, ``Core Operating Limits Report,'' to be
consistent with Specification 5.6.5 of NUREG-1432, ``Standard Technical
Specifications Combustion Engineering Plants.'' In addition, the list
of core reload analysis methodologies contained in TS 5.9.5b used to
determine the core operating limits, has been updated. Many of these
references were moved to the Omaha Public Power District core reload
analysis methodology documents OPPD-NA-8301, 8302, and 8303, which are
also listed in TS 5.9.5b. However, OPPD-NA-8302 has been revised to
incorporate use of the code CASMO-4 in lieu of the previously approved
CASMO-3 code.
Date of issuance: March 11, 2005.
Effective date: March 11, 2005, and shall be implemented within 90
days from the date of issuance.
Amendment No.: 233.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: October 12, 2004 (69 FR
60683)
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated March 11, 2005.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: May 21, 2004.
Brief description of amendments: The amendments revised the
Technical Specifications to delete the requirements to maintain
hydrogen recombiners and hydrogen analyzers.
Date of issuance: March 8, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 167 and 159.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: September 28, 2004 (69
FR 57994)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 8, 2005.
No significant hazards consideration comments received: No
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: July 28, 2004.
Brief description of amendments: The amendments delete the
technical specification requirements to submit monthly operating
reports and annual occupational radiation exposure reports.
Date of issuance: March 8, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 168 and 160.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: October 12, 2004 (69 FR
60686)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 8, 2005.
No significant hazards consideration comments received: No.
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 application for amendments: July 28, 2004.
Brief description of amendments: The amendments revised the
Technical Specifications by deleting the requirements for monthly
operating reports and occupational radiation exposure reports.
Date of issuance: March 8, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 245 and 189.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 12, 2004 (69 FR
60686).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 8, 2005.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of application for amendments: October 13, 2003, as
supplemented by letters dated April 12 and October 28, 2004.
Brief description of amendments: The amendments revised the
Technical Specifications (TS) limiting conditions for operation 3.8.4,
3.8.5, and 3.8.6, on direct current sources, operating and shutdown,
and battery cell parameters. The proposed amendments creates TS 5.5.19,
for a battery monitoring and maintenance program. The TS Bases are
revised to be consistent with these changes. The proposed amendments
are based on Technical Specification Task Force (TSTF) Traveler, TSTF-
360, Revision 1.
Date of issuance: March 2, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 133 and 112.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 20, 2004 (69 FR
2746). The supplements dated April 12 and October 28, 2004, provided
clarifying information that did not change the scope of the October 13,
2003, application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 2, 2005.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of application for amendments: May 21, 2004.
Brief description of amendments: The amendments revised the
Technical Specifications to delete the requirements to maintain
hydrogen recombiners and change requirements for hydrogen analyzers.
[[Page 15952]]
Date of issuance: March 7, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 134 and 113.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 28, 2004 (69
FR 57995).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 7, 2005.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of application for amendments: July 28, 2004.
Brief description of amendments: The amendments delete the
technical specification requirements to submit monthly operating
reports and annual occupational radiation exposure reports.
Date of issuance: March 8, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 135 and 114.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 12, 2004 (69 FR
60686)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 8, 2005.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: May 13, 2003, as supplemented by letters
dated October 6, 2004, November 30, 2004, and January 20, 2005.
Brief description of amendments: The amendments approve revisions
to the RETRAN-02 methodology that is used to evaluate certain design
basis transients and accidents.
Date of issuance: March 7, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: Unit 1--171; Unit 2--159.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the RETRAN-02 methodology.
Date of initial notice in Federal Register: November 12, 2003 (68
FR 64138). The supplements dated October 6, 2004, November 30, 2004,
and January 20, 2005, 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 7, 2005.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant, Units 2 and 3, Limestone County, Alabama
Date of application for amendments: July 8, 2004, as supplemented
in a letter dated November 24, 2004 (TS-448).
Brief description of amendments: The amendments modify Technical
Specification Section 5.5.12 ``Primary Containment Leakage Rate Testing
Program'' to allow a one-time 5-year extension to the 10-year frequency
of the performance-based leakage rate testing program for Type A tests.
The first Unit 2 Type A test performed after the November 6, 1994, Type
A test shall be performed no later than November 6, 2009, and the first
Unit 3 Type A test performed after the October 10, 1998, Type A test
shall be performed no later than October 10, 2013. The local leakage
rate tests (Type B and Type C), including their schedules, are not
affected by this request.
Date of issuance: March 9, 2005.
Effective date: As of date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 293 and 252.
Facility Operating License Nos. DPR-52 and DPR-68: Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: August 3, 2004 (69 FR
46592).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 9, 2005.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: August 18, 2004.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3/4.4.2, ``Safety Valves--Shutdown,'' TS 3/4.4.3,
``Safety and Relief Valves--Operating,'' and TS 3/4.5.2, ``ECCS
Subsystems--T avg Greater Than or Equal to 350[deg]F.'' TS
3/4.4.2 is eliminated because overpressure protection of the reactor
coolant system does not rely upon the pressurizer safety valves during
plant operation in Modes 4 and 5. TS 3/4.4.3 is revised to remove
redundancy and add improvements consistent with NUREG-1431, Revision 3,
``Standard Technical Specifications for Westinghouse Plants.'' TS 3/
4.5.2 is revised by adding a note to the Limiting Condition for
Operation (LCO) supporting transition to and from LCO 3.4.12, ``Low
Temperature Overpressure Protection (LTOP) System.''
Date of issuance: March 9, 2005.
Effective date: As of the date of issuance. Unit 1 shall be
implemented by May 15, 2005, and Unit 2 shall be implemented by
completion of the 2005 Cycle 13 Refueling Outage.
Amendment Nos.: 299 and 288.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the TSs.
Date of initial notice in Federal Register: November 9, 2004 (69 FR
64991)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 9, 2005.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: September 15, 2004.
Brief description of amendment: The amendment modifies technical
specification (TS) requirements for mode change limitations in Limiting
Condition for Operation 3.0.4 and Surveillance Requirement 3.0.4
consistent with Industry/TS Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-359, Revision 9, ``Increased
Flexibility in Mode Restraints.'' In addition, the amendment modifies
TS requirements consistent with TSTF-153, Revision 0, ``Clarify
Exception Notes to be Consistent with the Requirement Being Excepted,''
in part, and TSTF-285, Revision 1, ``Charging Pump Swap LTOP (Low
Temperature-Overpressurization) Allowance.''
Date of issuance: March 3, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 55.
Facility Operating License No. NPF-90: Amendment revises the TSs.
Date of initial notice in Federal Register: January 18, 2005 (70 FR
2901) and February 1, 2005 (70 FR 5226).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 3, 2005.
[[Page 15953]]
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: September 8, 2003, as
supplemented by letter dated September 11, 2003.
Brief description of amendment: The amendment revised the Updated
Final Safety Analysis Report (UFSAR) by modifying the design and
licensing basis to increase the postulated primary-to-secondary leakage
in the faulted steam generator following a main steamline break
accident from 1 to 3 gallons per minute.
Date of issuance: March 10, 2005.
Effective date: As of the date of issuance and shall be implemented
as part of the next UFSAR update made in accordance with 10 CFR
50.71(e).
Amendment No.: 56
Facility Operating License No. NPF-90: Amendment revised the UFSAR.
Date of initial notice in Federal Register: September 18, 2003 (68
FR 54745).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 10, 2005.
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: October 27, 2004.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) by eliminating the requirements in TSs 5.6.1 and
5.6.4 to submit monthly operating reports and annual occupational
radiation exposure reports.
Date of issuance: March 8, 2005.
Effective date: March 8, 2005, and shall be implemented within 90
days of the date of issuance.
Amendment No.: 166.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 4, 2005 (70 FR
406).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 8, 2005.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: July 22, 2004.
Brief description of amendment: The amendment revises Technical
Specification Figure 3.5.5-1, ``Seal Injection Flow Limits,'' to
reflect flow limits that allow a higher seal injection flow for a given
differential pressure between the charging pump discharge header and
the reactor coolant system.
Date of issuance: March 16, 2005.
Effective date: March 16, 2005, and shall be implemented prior to
startup from Refueling Outage 14.
Amendment No.: 160.
Facility Operating License No. NPF-42: The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 31, 2004 (69 FR
53115).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 16, 2005.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments To Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
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 for the
amendment 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.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection
[[Page 15954]]
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 System's
(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].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a
request for a hearing or petition for leave to intervene is filed by
the above date, 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 identify 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
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
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.\1\ 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 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.
---------------------------------------------------------------------------
\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
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. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
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 or
[[Page 15955]]
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).
Duke Energy Corporation, et al., Docket No. 50-414, Catawba Nuclear
Station Unit 2, York County, South Carolina
Date of amendment request: February 5, 2005, as supplemented by
letter dated February 7, 2005.
Description of amendment request: The amendment revises the system
bypass leakage acceptance criterion for the charcoal adsorber in the 2B
Auxiliary Building Filtered Ventilation Exhaust System train as listed
in Technical Specification 5.5.11, ``Ventilation Filter Testing
Program.''
Date of issuance: February 7, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 213.
Renewed Facility Operating License No. NPF-52: Amendments revised
the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC):
No.
The Commission's related evaluation of the amendment, finding of
emergency circumstances, state consultation, and final NSHC
determination are contained in a safety evaluation dated February 7,
2005.
Attorney for licensee: Ms. Anne Cottingham, Esquire.
NRC Section Chief: John A. Nakoski.
The Commission's related evaluation of the amendment, finding of
emergency circumstances, state consultation, and final NSHC
determination are contained in a safety evaluation dated February 7,
2005.
Attorney for licensee: Ms. Anne Cottingham, Esquire.
NRC Section Chief: John A. Nakoski.
Dated at Rockville, Maryland, this 21st day of March 2005.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. E5-1343 Filed 3-28-05; 8:45 am]
BILLING CODE 7590-01-P