Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 15940-15955 [E5-1343]
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Federal Register / Vol. 70, No. 59 / Tuesday, March 29, 2005 / Notices
call (recording)—(301) 415–1292.
Contact person for more information:
Dave Gamberoni, (301) 415–1651.
The NRC Commission Meeting
Schedule can be found on the Internet
at: https://www.nrc.gov/what-we-do/
policy-making/schedule.html.
The NRC provides reasonable
accommodation to individuals with
disabilities where appropriate. If you
need a reasonable accommodation to
participate in these public meetings, or
need this meeting notice or the
transcript or other information from the
public meetings in another format (e.g.,
braille, large print), please notify the
NRC’s Disability Program Coordinator,
August Spector, at (301) 415–7080,
TDD: (301) 415–2100, or by e-mail at
aks@nrc.gov. Determinations on
requests for reasonable accommodation
will be made on a case-by-case basis.
This notice is distributed by mail to
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longer wish to receive it, or would like
to be added to the distribution, please
contact the Office of the Secretary,
Washington, DC 20555 (301) 415–1969).
In addition, distribution of this meeting
notice over the Internet system is
available. If you are interested in
receiving this Commission meeting
schedule electronically, please send an
electronic message to dkw@nrc.gov.
Dated: March 24, 2005.
R. Michelle Schroll,
Office of the Secretary.
[FR Doc. 05–6239 Filed 3–25–05; 9:07 am]
BILLING CODE 7590–01–M
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.
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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 60day 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
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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, https://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
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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
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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,
HearingDocket@nrc.gov; 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 email to OGCMailCenter@nrc.gov. 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, https://
www.nrc.gov/reading-rm/adams.html. If
you do not have access to ADAMS or if
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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
pdr@nrc.gov.
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
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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 nonsafety 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.
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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
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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 15years 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.
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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
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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
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15943
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
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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
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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
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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
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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
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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
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15945
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
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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.
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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?
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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.
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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.
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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
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15947
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
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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
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consideration, which is presented
below:
and does not involve any significant safety
hazards.
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,
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.
PO 00000
Frm 00132
Fmt 4703
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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)
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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
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NRC Web site, https://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
pdr@nrc.gov.
AmerGen Energy Company, LLC, Docket
No. 50–461, Clinton Power Station, Unit
1, DeWitt County, Illinois
15949
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.
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
Energy Northwest, Docket No. 50–397,
Technical Specifications.
Columbia Generating Station, Benton
Date of initial notice in Federal
Register: October 26, 2004 (69 FR 62469). County, Washington
Date of application for amendment:
The Commission’s related evaluation
September 27, 2004.
of the amendment is contained in a
Brief description of amendment: The
Safety Evaluation dated March 2, 2005.
amendment eliminated the technical
No significant hazards consideration
specification requirements to submit a
comments received: No.
monthly operating report and an annual
Carolina Power & Light Company,
occupational radiation exposure report.
Docket No. 50–324, Brunswick Steam
Date of issuance: March 9, 2005.
Electric Plant, Unit 2, Brunswick
Effective date: March 9, 2005.
County, North Carolina
Amendment No.: 190.
Facility Operating License No. NPF–
Date of application for amendment:
21: The amendment revised the
August 16, 2004.
Technical Specifications.
Brief Description of amendment: The
Date of initial notice in Federal
amendment adds topical report NEDE–
Register: October 26, 2004 (69 FR 62472).
32906P–A, ‘‘TRACG Application for
The Commission’s related evaluation
Anticipated Operational Occurrences
of the amendment is contained in a
(AOO) Transient Analyses,’’ to the
Safety Evaluation dated March 9, 2005.
documents listed in Technical
No significant hazards consideration
Specification 5.6.5 describing the
comments received: No.
approved methodologies used to
Entergy Operations, Inc., Docket No. 50–
determine the core operating limits.
368, Arkansas Nuclear One Unit No. 2,
Date of issuance: March 4, 2005.
Pope County, Arkansas
Effective date: March 4, 2005.
Date of application for amendment:
Amendment No.: 262.
April 15, 2004, as supplemented
Facility Operating License No DPR–
January 20, 2005.
62: Amendment revises the Technical
Brief Description of amendments: The
Specifications.
licensee has proposed to change the
Date of initial notice in Federal
existing reactor coolant system (RCS)
Register: October 26, 2004 (69 FR 62470). cooldown curve to a single 32 effective
The Commission’s related evaluation
full power year pressure/temperature
of the amendment is contained in a
limit curve that is applicable for
Safety Evaluation dated March 4, 2005.
cooldowns at a rate of 100 °F/hour or 50
°F in any half-hour step. The licensee’s
No significant hazards consideration
proposed curve is applicable to RCS
comments received: No.
PO 00000
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Federal Register / Vol. 70, No. 59 / Tuesday, March 29, 2005 / Notices
FirstEnergy Nuclear Operating
Units 1 and 2, Rock Island County,
cold-leg temperatures ranging from 50
Illinois
°F to 560 °F.
Company, et al., Docket Nos. 50–334
Date of application for amendments:
Date of issuance: March 7, 2005.
and 50–412, Beaver Valley Power
Effective date: As of the date of
April 30, 2004.
Station, Unit Nos. 1 and 2 (BVPS–1 and
Brief description of amendments: The 2), Beaver County, Pennsylvania
issuance to be implemented within 60
amendments modify Technical
days from the date of issuance.
Date of amendment request: June 1,
Amendment No.: 256.
Specifications (TS) requirements to
2004, as supplemented July 23, 2004,
Facility Operating License No. NFP–6: adopt the provisions of Industry/TS
and February 18, 2005.
Amendment revised the Technical
Task Force (TSTF) change TSTF–359,
Description of amendment request:
Specifications.
‘‘Increased Flexibility in Mode
These amendments lowered the BVPS–
Date of initial notice in Federal
Restraints.’’
2 overpressure protection system enable
Register: May 11, 2004 (69 FR 26188). The Date of issuance: March 10, 2005.
temperature, allowed one inoperable
supplemental letter provided additional
Effective date: As of the date of
residual heat removal loop during
information that clarified the
issuance and shall be implemented
surveillance testing, removed the BVPS–
application, did not expand the scope of within 180 days.
1 list of figures and list of tables from
Amendment Nos.: 212/204/223/218.
the application as originally noticed,
the Index of the BVPS–1 Technical
Facility Operating License Nos. DPR–
and did not change the staff’s original
Specifications (TSs), and made minor
19, DPR–25, DPR–29 and DPR–30. The
proposed no significant hazards
changes to achieve consistency between
amendments revised the Technical
consideration determination as
units and with the Standard TSs for
Specifications.
published in the Federal Register.
Date of initial notice in Federal
Westinghouse plants and with some TS
The Commission’s related evaluation
Register: October 26, 2004 (69 FR 62474). Task Force changes.
of the amendments is contained in a
The Commission’s related evaluation
Safety Evaluation dated March 7, 2005.
Date of issuance: March 11, 2004.
of the amendments is contained in a
No significant hazards consideration
Effective date: As of the date of
Safety Evaluation dated March 10, 2005. issuance, to be implemented within 30
comments received: No.
No significant hazards consideration
days.
Entergy Nuclear Operations, Inc.,
comments received: No.
Amendment Nos.: 265 and 146.
Docket No. 50–293, Pilgrim Nuclear
Facility Operating License Nos. DPR–
Exelon Generation Company, LLC,
Power Station, Plymouth County,
66 and NPF–73: Amendments revised
Docket Nos. 50–254 and 50–265, Quad
Massachusetts
the Technical Specifications.
Cities Nuclear Power Station, Units 1
Date of application for amendment:
Public comments requested as to
and 2, Rock Island County, Illinois
April 14, 2004.
proposed no significant hazards
Date of application for amendments:
Brief description of amendment: The
consideration (NSHC): Yes. February 25,
amendment revised the Pilgrim Nuclear June 10, 2004, and supplemented July
2005 (70 FR 9391). The notice provided
19 and July 21, 2004 and January 21,
Power Station Technical Specifications
an opportunity to submit comments on
2005.
(TSs) by adding a new limiting
Brief description of amendments: The the Commission’s proposed NSHC
condition for operation (LCO) 3.0.7 to
determination by March 11, 2005. No
amendments revise the Quad Cities
Section 3.0, ‘‘Limiting Condition for
comments have been received. The
Nuclear Power Station Technical
Operation (LCO) Applicability,’’ a new
notice also provided an opportunity to
Specifications to change the allowable
TS Section 3.14, ‘‘Special Operations,’’
request a hearing by April 26, 2005, but
value and add Surveillance
and a new LCO 3.14.A, ‘‘Inservice Leak
indicated that if the Commission makes
Requirements for the Main Steam Line
and Hydrostatic Testing Operation,’’ to
Flow-High initiation of Group 1 Primary a final NSHC determination, any such
the TSs. These changes permit the
hearing would take place after issuance
Containment Isolation System and
licensee to perform inservice
of the amendment.
Control Room Emergency Ventilation
hydrostatic testing and system leakage
The Commission’s related evaluation
System isolation.
pressure testing of the reactor coolant
of the amendment, finding of exigent
Date of issuance: March 15, 2005.
system at temperatures greater than 212
circumstances, state consultation, and
Effective date: As of the date of
°F with the reactor shut down.
final NSHC determination are contained
issuance and shall be implemented
Date of issuance: March 16, 2005.
in a safety evaluation dated March 11,
within 90 days for Unit 1 and no later
Effective Date: As of the date of
2005.
than 90 days after the start of the Unit
issuance, and shall be implemented
Attorney for licensee: Mary O’Reilly,
2 refueling outage currently scheduled
within 30 days.
FirstEnergy Nuclear Operating
for March 2006 for Unit 2.
Amendment No.: 211.
Company, FirstEnergy Corporation, 76
Amendment Nos.: 224, 219
Facility Operating License No. DPR–
South Main Street, Akron, OH 44308.
Facility Operating License Nos. DPR–
35: The amendment revised the TSs.
29 and DPR–30: The amendments revise Indiana Michigan Power Company,
Date of initial notice in Federal
the Technical Specifications.
Register: December 21, 2004 (69 FR
Docket Nos. 50–315 and 50–316, Donald
Date of initial notice in Federal
76489).
C. Cook Nuclear Plant, Units 1 and 2,
Register: August 31, 2004 (69 FR 53107). Berrien County, Michigan
The Commission’s related evaluation
The supplemental letters contained
of the amendment is contained in a
Date of application for amendments:
Safety Evaluation dated March 16, 2005. clarifying information and did not
change the initial no significant hazards April 13, 2004.
No significant hazards consideration
Brief description of amendments: The
consideration determination and did not
comments received: No.
expand the scope of the original Federal amendments change the design basis as
Exelon Generation Company, LLC,
described in the Updated Final Safety
Register notice.
Docket Nos. 50–237 and 50–249,
Analysis Report to allow the use in
The Commission’s related evaluation
Dresden Nuclear Power Station, Units 2 of the amendments is contained in a
control rod drive missile shield
and 3, Grundy County, Illinois
Safety Evaluation dated March 15, 2005. structural calculations of a reinforcing
bar (rebar) yield strength value based on
No significant hazards consideration
Docket Nos. 50–254 and 50–265,
measured material properties, as
comments received: No.
Quad Cities Nuclear Power Station,
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requirements to maintain hydrogen
documented in the licensee rebar
recombiners and hydrogen analyzers.
acceptance tests.
Date of issuance: March 11, 2005.
Date of issuance: March 8, 2005.
Effective date: As of the date of
Effective date: As of the date of
issuance and shall be implemented
issuance and shall be implemented
within 45 days.
within 60 days.
Amendment Nos.: 286, 268.
Amendment Nos.: 167 and 159.
Facility Operating License Nos. DPR–
Facility Operating License Nos. NPF–
58 and DPR–74: Amendments revised
2 and NPF–8: Amendments revise the
the design basis.
Technical Specifications.
Date of initial notice in Federal
Date of initial notice in Federal
Register: October 12, 2004 (69 FR 60682). Register: September 28, 2004 (69 FR
The Commission’s related evaluation
57994)
of the amendments is contained in a
The Commission’s related evaluation
Safety Evaluation dated March 11, 2005. of the amendments is contained in a
No significant hazards consideration
Safety Evaluation dated March 8, 2005.
comments received: No.
No significant hazards consideration
comments received: No
Omaha Public Power District, Docket
15951
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.
No. 50–285, Fort Calhoun Station, Unit
Southern Nuclear Operating Company,
Brief description of amendments: The
No. 1, Washington County, Nebraska
Inc., Docket Nos. 50–348 and 50–364,
amendments revised the Technical
Joseph M. Farley Nuclear Plant, Units 1
Date of amendment request:
Specifications (TS) limiting conditions
and 2, Houston County, Alabama
September 7, 2004.
for operation 3.8.4, 3.8.5, and 3.8.6, on
Brief description of amendment: The
direct current sources, operating and
Date of amendments request: July 28,
amendment revised Technical
shutdown, and battery cell parameters.
2004.
Specification (TS) 5.9.5, ‘‘Core
Brief description of amendments: The The proposed amendments creates TS
Operating Limits Report,’’ to be
amendments delete the technical
5.5.19, for a battery monitoring and
consistent with Specification 5.6.5 of
specification requirements to submit
maintenance program. The TS Bases are
NUREG–1432, ‘‘Standard Technical
monthly operating reports and annual
revised to be consistent with these
Specifications Combustion Engineering
occupational radiation exposure reports. changes. The proposed amendments are
Plants.’’ In addition, the list of core
Date of issuance: March 8, 2005.
based on Technical Specification Task
Effective date: As of the date of
reload analysis methodologies
Force (TSTF) Traveler, TSTF–360,
issuance and shall be implemented
contained in TS 5.9.5b used to
Revision 1.
determine the core operating limits, has within 60 days from the date of
Date of issuance: March 2, 2005.
issuance.
been updated. Many of these references
Effective date: As of the date of
Amendment Nos.: 168 and 160.
were moved to the Omaha Public Power
issuance and shall be implemented
Facility Operating License Nos. NPF–
District core reload analysis
within 30 days from the date of
2 and NPF–8: Amendments revise the
methodology documents OPPD–NA–
issuance.
Technical Specifications.
Amendment Nos.: 133 and 112.
8301, 8302, and 8303, which are also
Date of initial notice in Federal
Facility Operating License Nos. NPF–
listed in TS 5.9.5b. However, OPPD–
Register: October 12, 2004 (69 FR 60686) 68 and NPF–81: Amendments revised
NA–8302 has been revised to
The Commission’s related evaluation
the Technical Specifications.
incorporate use of the code CASMO–4
of the amendments is contained in a
Date of initial notice in Federal
in lieu of the previously approved
Safety Evaluation dated March 8, 2005.
Register: January 20, 2004 (69 FR 2746).
CASMO–3 code.
No significant hazards consideration
The supplements dated April 12 and
Date of issuance: March 11, 2005.
comments received: No.
October 28, 2004, provided clarifying
Effective date: March 11, 2005, and
information that did not change the
shall be implemented within 90 days
Southern Nuclear Operating Company,
scope of the October 13, 2003,
from the date of issuance.
Inc., Georgia Power Company,
application nor the initial proposed no
Amendment No.: 233.
Oglethorpe Power Corporation,
Renewed Facility Operating License
significant hazards consideration
Municipal Electric Authority of Georgia,
No. DPR–40: The amendment revised
determination.
City of Dalton, Georgia, Docket Nos. 50–
The Commission’s related evaluation
the Technical Specifications.
321 and 50–366, Edwin I. Hatch Nuclear
of the amendments is contained in a
Date of initial notice in Federal
Plant, Units 1 and 2, Appling County,
Safety Evaluation dated March 2, 2005.
Register: October 12, 2004 (69 FR 60683)
Georgia
No significant hazards consideration
The Commission’s related evaluation
Date of application for amendments:
comments received: No.
of the amendment is contained in a
safety evaluation dated March 11, 2005. July 28, 2004.
Brief description of amendments: The Southern Nuclear Operating Company,
No significant hazards consideration
Inc., Docket Nos. 50–424 and 50–425,
amendments revised the Technical
comments received: No.
Vogtle Electric Generating Plant, Units 1
Specifications by deleting the
Southern Nuclear Operating Company,
and 2, Burke County, Georgia
requirements for monthly operating
Inc., Docket Nos. 50–348 and 50–364,
Date of application for amendments:
reports and occupational radiation
Joseph M. Farley Nuclear Plant, Units 1
May 21, 2004.
exposure reports.
and 2, Houston County, Alabama
Brief description of amendments: The
Date of issuance: March 8, 2005.
amendments revised the Technical
Effective date: As of the date of
Date of amendments request: May 21,
Specifications to delete the
issuance and shall be implemented
2004.
requirements to maintain hydrogen
Brief description of amendments: The within 60 days from the date of
recombiners and change requirements
amendments revised the Technical
issuance.
for hydrogen analyzers.
Amendment Nos.: 245 and 189.
Specifications to delete the
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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.
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
Southern Nuclear Operating Company,
Nuclear Plant, Units 2 and 3, Limestone
Inc., et al., Docket Nos. 50–424 and 50–
County, Alabama
425, Vogtle Electric Generating Plant,
Date of application for amendments:
Units 1 and 2, Burke County, Georgia
July 8, 2004, as supplemented in a letter
Date of application for amendments:
dated November 24, 2004 (TS–448).
July 28, 2004.
Brief description of amendments: The
Brief description of amendments: The amendments modify Technical
amendments delete the technical
Specification Section 5.5.12 ‘‘Primary
specification requirements to submit
Containment Leakage Rate Testing
monthly operating reports and annual
Program’’ to allow a one-time 5-year
occupational radiation exposure reports. extension to the 10-year frequency of
Date of issuance: March 8, 2005.
the performance-based leakage rate
Effective date: As of the date of
testing program for Type A tests. The
issuance and shall be implemented
first Unit 2 Type A test performed after
within 60 days from the date of
the November 6, 1994, Type A test shall
issuance.
be performed no later than November 6,
Amendment Nos.: 135 and 114.
2009, and the first Unit 3 Type A test
Facility Operating License Nos. NPF–
performed after the October 10, 1998,
68 and NPF–81: Amendments revised
Type A test shall be performed no later
the Technical Specifications.
than October 10, 2013. The local leakage
Date of initial notice in Federal
Register: October 12, 2004 (69 FR 60686) rate tests (Type B and Type C),
including their schedules, are not
The Commission’s related evaluation
affected by this request.
of the amendments is contained in a
Date of issuance: March 9, 2005.
Safety Evaluation dated March 8, 2005.
Effective date: As of date of issuance
No significant hazards consideration
and shall be implemented within 30
comments received: No.
days.
STP Nuclear Operating Company,
Amendment Nos.: 293 and 252.
Docket Nos. 50–498 and 50–499, South
Facility Operating License Nos. DPR–
Texas Project, Units 1 and 2, Matagorda 52 and DPR–68: Amendments revise the
County, Texas
Technical Specifications.
Date of initial notice in Federal
Date of amendment request: May 13,
Register: August 3, 2004 (69 FR 46592).
2003, as supplemented by letters dated
The Commission’s related evaluation
October 6, 2004, November 30, 2004,
of the amendment is contained in a
and January 20, 2005.
Brief description of amendments: The Safety Evaluation dated March 9, 2005.
No significant hazards consideration
amendments approve revisions to the
comments received: No.
RETRAN–02 methodology that is used
to evaluate certain design basis
Tennessee Valley Authority, Docket
transients and accidents.
Nos. 50–327 and 50–328, Sequoyah
Date of issuance: March 7, 2005.
Nuclear Plant, Units 1 and 2, Hamilton
Effective date: As of the date of
County, Tennessee
issuance and shall be implemented
Date of application for amendments:
within 30 days of issuance.
August 18, 2004.
Amendment Nos.: Unit 1—171; Unit
Brief description of amendments: The
2—159.
amendments revised Technical
Facility Operating License Nos. NPF–
Specification (TS) 3/4.4.2, ‘‘Safety
76 and NPF–80: The amendments
Valves—Shutdown,’’ TS 3/4.4.3, ‘‘Safety
revised the RETRAN–02 methodology.
and Relief Valves—Operating,’’ and TS
Date of initial notice in Federal
3/4.5.2, ‘‘ECCS Subsystems—T avg
Register: November 12, 2003 (68 FR
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Greater Than or Equal to 350°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 TemperatureOverpressurization) 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.
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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 primaryto-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.
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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
PO 00000
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15953
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
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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, https://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
pdr@nrc.gov.
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, https://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 email to pdr@nrc.gov. 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
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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.
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/
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.
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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,
HearingDocket@nrc.gov; 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 email to OGCMailCenter@nrc.gov. 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
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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]
SECURITIES AND EXCHANGE
COMMISSION
Issuer Delisting; Notice of Application
of Hythiam, Inc. to Withdraw its
Common Stock, $.0001 par value, From
Listing and Registration on the
American Stock Exchange LLC File No.
1–31932
17:01 Mar 28, 2005
Jkt 205001
imposed by the Commission for the
protection of investors. All comment
letters may be submitted by either of the
following methods:
Electronic Comments
• Send an e-mail to rulecomments@sec.gov. Please include the
File Number 1–31932 or;
March 22, 2005.
Paper Comments
On March 7, 2005, Hythiam, Inc., a
Delaware corporation (‘‘Issuer’’), filed
an application with the Securities and
Exchange Commission (‘‘Commission’’),
pursuant to section 12(d) of the
Securities Exchange Act of 1934
(‘‘Act’’)1 and Rule 12d2–2(d)
thereunder,2 to withdraw its common
stock, $.0001 par value (‘‘Security’’),
from listing and registration on the
American Stock Exchange LLC
(‘‘Amex’’).
On March 4, 2005, the Board of
Directors (‘‘Board’’) of the Issuer
unanimously approved resolutions to
withdraw the Security from listing and
registration on Amex and to list the
Security on the Nasdaq National Market
(‘‘Nasdaq’’). The Board determined that
it is in the best interest of the Issuer and
its stockholders to withdraw the
Security from listing on the Amex and
to list the Security on Nasdaq. The
Board believed that listing the Security
on Nasdaq will enable the Issuer and its
stockholders to benefit from increased
visibility to investors, an open market
structure, and an efficient electronic
trading platform. In addition, the Board
stated that the Issuer has met the initial
listing requirements of Nasdaq, and the
application for listing the Security on
Nasdaq has been approved.
The Issuer stated that it has met the
requirements of Amex’s rules governing
an issuer’s voluntary withdrawal of a
security from listing and registration by
complying with all the applicable laws
in effect in Delaware, in which it is
incorporated.
The Issuer’s application relates solely
to the withdrawal of the Security from
listing on the Amex and from
registration under section 12(b) of the
Act,3 and shall not affect its obligation
to be registered under section 12(g) of
the Act.4
Any interested person may, on or
before April 15, 2005, comment on the
facts bearing upon whether the
application has been made in
accordance with the rules of the Amex,
and what terms, if any, should be
• Send paper comments in triplicate
to Jonathan G. Katz, Secretary,
Securities and Exchange Commission,
450 Fifth Street, NW., Washington, DC
20549–0609.
All submissions should refer to File
Number 1–31932. This file number
should be included on the subject line
if e-mail is used. To help us process and
review your comments more efficiently,
please use only one method. The
Commission will post all comments on
the Commission’s Internet Web site
(https://www.sec.gov/rules/delist.shtml).
Comments are also available for public
inspection and copying in the
Commission’s Public Reference Room.
All comments received will be posted
without change; we do not edit personal
identifying information from
submissions. You should submit only
information that you wish to make
available publicly.
The Commission, based on the
information submitted to it, will issue
an order granting the application after
the date mentioned above, unless the
Commission determines to order a
hearing on the matter.
BILLING CODE 7590–01–P
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1 15
U.S.C. 78l(d).
2 17 CFR 240.12d2–2(d).
3 15 U.S.C. 78l(b).
4 15 U.S.C. 78l(g).
Frm 00139
Fmt 4703
Sfmt 4703
For the Commission, by the Division of
Market Regulation, pursuant to delegated
authority.5
Jonathan G. Katz,
Secretary.
[FR Doc. E5–1377 Filed 3–28–05; 8:45 am]
BILLING CODE 8010–01–P
SECURITIES AND EXCHANGE
COMMISSION
[Release No. 34–51418; File No. SR–BSE–
2005–01]
Self-Regulatory Organizations; Notice
of Filing of Proposed Rule Change,
and Amendment No. 1 Thereto, by the
Boston Stock Exchange, Inc. Relating
to the Price Improvement Period Under
the Rules of the Boston Options
Exchange Facility
March 23, 2005.
Pursuant to section 19(b)(1) of the
Securities Exchange Act of 1934
5 17
E:\FR\FM\29MRN1.SGM
CFR 200.30–3(a)(1).
29MRN1
Agencies
[Federal Register Volume 70, Number 59 (Tuesday, March 29, 2005)]
[Notices]
[Pages 15940-15955]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E5-1343]
-----------------------------------------------------------------------
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, https://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, HearingDocket@nrc.gov;
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
OGCMailCenter@nrc.gov. 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, https://
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 pdr@nrc.gov.
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