[Federal Register Volume 69, Number 3 (Tuesday, January 6, 2004)]
[Notices]
[Pages 691-705]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-8]
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UNITED STATES NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision 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 December 12, 2003, through December 23,
2003. The last biweekly notice was published on December 23, 2003 (68
FR 74262).
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.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing 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
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By February 5, 2004, the licensee may file a request for a hearing
with respect
[[Page 692]]
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 2.714, which is available at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of 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. Petitioner must provide 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 to relief. A
petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, 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 with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of continuing disruptions in delivery of mail to United States
Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the request for
hearing and petition for leave to intervene should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and because of continuing disruptions in
delivery of mail to United States Government offices, it is requested
that copies be transmitted either by means of facsimile transmission to
301-415-3725 or by e-mail to [email protected]. A copy of the
request for hearing and petition for leave to intervene should also be
sent to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
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 Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to
[email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: December 2, 2003.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TS) Surveillance Requirement (SR)
4.0.2 to extend the delay period, before entering a Limiting Condition
for Operation, following a missed surveillance. The
[[Page 693]]
delay period would be extended from the current limit of ``* * * up to
24 hours or up to the limit of the specified frequency, whichever is
less* * *'' to ``* * *up to 24 hours or up to the limit of the
specified frequency, whichever is greater.* * *'' To support this
change, the following requirement would be added to SR 4.0.2: ``A risk
evaluation shall be performed for any surveillance delayed greater than
24 hours and the risk impact shall be managed.'' Additionally, a new
section 6.2.1 will be added to provide for a TS Bases Control Program.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments
concerning missed surveillances, including a model safety evaluation
and model no significant hazards consideration (NSHC) determination,
using the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on September 28, 2001 (66 FR 49714). The licensee affirmed the
applicability of the following NSHC determination in its application.
The NSHC determination is restated below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC 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 relaxes the time allowed to perform a missed
surveillance. The time between surveillances is not an initiator of
any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The
equipment being tested is still required to be operable and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected. Any reduction
in confidence that a standby system might fail to perform its safety
function due to a missed surveillance is small and would not, in the
absence of other unrelated failures, lead to an increase in
consequences beyond those estimated by existing analyses. The
addition of a requirement to assess and manage the risk introduced
by the missed surveillance will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. A
missed surveillance will not, in and of itself, introduce new
failure modes or effects and any increased chance that a standby
system might fail to perform its safety function due to a missed
surveillance would not, in the absence of other unrelated failures,
lead to an accident beyond those previously evaluated. The addition
of a requirement to assess and manage the risk introduced by the
missed surveillance will further minimize possible concerns. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in [a] Margin of Safety
The extended time allowed to perform a missed surveillance does
not result in a significant reduction in [a] margin of safety. As
supported by the historical data, the likely outcome of any
surveillance is verification that the LCO [Limiting Condition for
Operation] is met. Failure to perform a surveillance within the
prescribed frequency does not cause equipment to become inoperable.
The only effect of the additional time allowed to perform a missed
surveillance on [a] margin of safety is the extension of the time
until inoperable equipment is discovered to be inoperable by the
missed surveillance. However, given the rare occurrence of
inoperable equipment, and the rare occurrence of a missed
surveillance, a missed surveillance on inoperable equipment would be
very unlikely. This must be balanced against the real risk of
manipulating the plant equipment or condition to perform the missed
surveillance. In addition, parallel trains and alternate equipment
are typically available to perform the safety function of the
equipment not tested. Thus, there is confidence that the equipment
can perform its assumed safety function.
Therefore, this change does not involve a significant reduction
in a margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius,
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Richard J. Laufer.
Carolina Power & Light Company, Docket No. 50-325, Brunswick Steam
Electric Plant, Unit 1, Brunswick County, North Carolina
Date of amendment request: October 31, 2003.
Description of amendment request: The proposed amendment would
revise the Minimum Critical Power Ratio (MCPR) Safety Limit contained
in Technical Specification (TS) 2.1.1.2. Currently the MCPR value is
greater than or equal to 1.12 for two recirculation loop operation and
greater than or equal to 1.14 for single recirculation loop operation.
The proposed revised MCPR would be greater than or equal to 1.11 for
two recirculation loop operation and greater than or equal to 1.12 for
single recirculation loop operation. Also, a second proposed change
would add topical report NEDE-32906P-A, ``TRACG Application for
Anticipated Operational Occurrences (AOO) Transient Analyses,'' to the
list of documents specified in TS 5.6.5 describing the approved
methodologies used to determine the core operating limits.
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.
Proposed Change 1
The proposed change to Technical Specification 2.1.1.2 does not
alter the assumptions of the accident analyses or the Technical
Specification Bases. The MCPR Safety Limit values are calculated to
ensure that greater than 99.9 percent of the fuel rods in the core
avoid transition boiling during any plant operation if the safety
limit is not violated. The derivation of the MCPR Safety Limit
values specified in the Technical Specifications has been performed
using the methods discussed in ``General Electric Standard
Application for Reactor Fuel,'' NEDE-24011-P-A-14 (i.e., GESTAR-II),
and U.S. Supplement, NEDE-24011-P-A-14-US, June 2000, which
incorporates Amendment 26. By letters dated November 10, 1999, and
March 29, 2000, GNF, the NRC approved the use of Amendment 26 to
NEDE-24011-P-A. Appropriate operational MCPR limits are applied that
ensure the MCPR Safety Limit is not exceeded during all modes of
operation and anticipated operational occurrences.
The revised MCPR Safety Limit values do not affect the
operability of any plant systems nor do these revised values
compromise any fuel performance limits; therefore, the probability
of fuel damage will not be increased as a result of this change. The
MCPR Safety Limit values do not impact the source term or pathways
assumed in accidents previously evaluated, and there are no adverse
effects on the factors contributing to offsite or onsite
radiological doses. In addition, the revised MCPR Safety Limit
values do not affect the performance of any equipment used to
mitigate the consequences
[[Page 694]]
of a previously evaluated accident and do not affect setpoints that
initiate protective or mitigative actions.
Therefore, the proposed change to MCPR Safety Limit values
contained in Technical Specification 2.1.1.2 does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Proposed Change 2
The proposed change to TS 5.6.5 will add General Electric
Nuclear Energy topical report NEDE-32906P-A, ``TRACG Application for
Anticipated Operational Occurrences (AOO) Transient Analyses,'' to
the list of documents describing approved methodologies for
determining core operating limits. Analyzed events are assumed to be
initiated by the failure of plant structures, systems, or
components. The core operating limits, which are developed using the
topical report being added, ensure that the integrity of the fuel
will be maintained during normal operations and that design
requirements will continue to be met. The proposed change does not
involve physical changes to any plant structure, system, or
component. Therefore, the probability of occurrence for a previously
analyzed accident is not significantly increased.
The consequences of a previously analyzed accident are dependent
on the initial conditions assumed for the analysis, the behavior of
the fuel during the analyzed accident, the availability and
successful functioning of the equipment assumed to operate in
response to the analyzed event, and the setpoints at which these
actions are initiated. Use of the analytical methodologies described
in the topical report being added to TS 5.6.5 will ensure that
applicable design and safety analyses acceptance criteria are met.
Use of these NRC-approved methodologies does not affect the
performance of any equipment used to mitigate the consequences of an
analyzed accident. As a result, no analysis assumptions are violated
and there are no adverse effects on the factors that contribute to
offsite or onsite dose as the result of an accident. Use of the
approved methodologies described in the topical report being added
to TS 5.6.5 ensures that plant structures, systems, or components
are maintained consistent with the safety analysis and licensing
bases. Based on this evaluation, there is no significant increase in
the consequences of a previously analyzed event.
Therefore, the proposed change adding General Electric Nuclear
Energy topical report NEDE-32906P-A to the TS 5.6.5 list of
documents describing approved methodologies for determining core
operating limits 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.
Proposed Change 1
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, including changes in
allowable modes of operation. The proposed revision of the MCPR
Safety Limit values does not involve installation of any new or
different equipment. No installed equipment is being operated in a
different manner than currently evaluated. No new initiating events
or transients will result from use of the revised MCPR Safety Limit
values. As a result, no new failure modes are being introduced.
Therefore, the proposed change to MCPR Safety Limit values contained
in Technical Specification 2.1.1.2 does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Proposed Change 2
The proposed change adding topical report NEDE-32906P-A to TS
5.6.5, and the use of the analytical methods described therein, does
not involve any physical alteration of plant systems, structures, or
components, other than allowing for fuel and core designs in
accordance with NRC approved methodologies. The proposed methodology
continues to meet applicable criteria for core operating limit
analysis. No new or different equipment is being installed. No
installed equipment is being operated in a different manner. There
is no alteration to the parameters within which the plant is
normally operated or in the setpoints that initiate protective or
mitigative actions. As a result no new failure modes are being
introduced.
Therefore, the proposed change adding General Electric Nuclear
Energy topical report NEDE-32906P-A to the TS 5.6.5 list of
documents describing approved methodologies for determining core
operating limits 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.
Proposed Change 1
The margin of safety is established through the design of the
plant structures, systems, and components; through the parameters
within which the plant is operated; through the establishment of
setpoints for actuation of equipment relied upon to respond to an
event; and through margins contained within the safety analyses. The
revised MCPR Safety Limit values will not adversely impact the
performance of plant structures, systems, components, and setpoints
relied upon to respond to mitigate an accident or transient. The
MCPR Safety Limit values are calculated to ensure that greater than
99.9 percent of the fuel rods in the core avoid transition boiling
during any anticipated operation occurrences if the safety limit is
not violated, thereby ensuring that fuel cladding integrity is
maintained. The revised MCPR Safety Limit values have been
calculated using NRC approved methods and procedures and preserve
the existing margin to transition boiling. Based on the assurance
that the fuel design criteria are being met, the revised MCPR Safety
Limit values do not involve a reduction in a margin of safety.
Proposed Change 2
The margin of safety is established through the design of the
plant structures, systems, and components, through the parameters
within which the plant is operated, through the establishment of the
setpoints for the actuation of equipment relied upon to respond to
an event, and through margins contained within the safety analyses.
The proposed change adding General Electric Nuclear Energy topical
report NEDE-32906P-A to the TS 5.6.5 list of documents describing
approved methodologies for determining core operating limits does
not impact the condition or performance of structures, systems,
setpoints, and components relied upon for accident mitigation. The
proposed change does not significantly impact any safety analysis
assumptions or results. Therefore, the proposed change adding
topical report NEDE-32906P-A to the TS 5.6.5 list of documents
describing approved methodologies for determining core operating
limits does not result in a significant reduction in the margin of
safety.
Based on the above, PEC concludes that the proposed amendment
presents 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: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Allen G. Howe.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois, Docket
Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle
County, Illinois, and Docket Nos. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois
Date of amendment request: November 3, 2003.
Description of amendment request: The proposed amendment would
modify Technical Specifications (TS) 3.4.1, ``Recirculation Loops
Operating,'' to add a requirement for the linear heat generation rate
(LHGR) limits specified in the Core Operating Limits Report (COLR) to
be met during single recirculation loop operation.
Technical Specification 3.4.1 for Dresden Nuclear Power Station
(DNPS) Units 2 and 3, LaSalle County Station
[[Page 695]]
(LSCS) Units 1 and 2, and Quad Cities Nuclear Power Station (QCNPS)
Units 1 and 2, currently requires limits for average planar linear heat
generation rate (APLHGR) and minimum critical power ratio (MCPR), as
well as allowable values for certain Reactor Protection System and
Control Rod Block functions, to be modified during single recirculation
loop operation. The modified limits for APLHGR and MCPR are specified
in the COLR. The proposed change adds a requirement to modify the LHGR
limit as specified in the COLR with one recirculation loop in
operation. Although there is currently no TS requirement to adjust the
LHGR limit during single recirculation loop operation, in accordance
with NRC Administrative Letter 98-10, ``Dispositioning of Technical
Specifications that Are Insufficient to Assure Plant Safety,''
administrative controls are in place at DNPS and QCNPS to ensure that
the LHGR limits are appropriately adjusted.
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 probability or consequences of an accident previously evaluated.
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. The LHGR is a measure of the heat generation rate of a
fuel rod in a fuel assembly at any axial location. Limits on the
LHGR are specified to ensure that fuel design limits are not
exceeded anywhere in the core during normal operation, including
anticipated operational occurrences, and to ensure that the peak
cladding temperature (PCT) during a postulated design basis LOCA
does not exceed the limits specified in 10 CFR 50.46.
LHGR limits have been established consistent with the NRC-
approved GESTAR methodology to ensure that fuel performance during
normal, transient, and accident conditions is acceptable. The
proposed change establishes a requirement for LHGR limits to be
modified, as specified in the COLR, during SLO such that the fuel is
protected during SLO and during any plant transients or anticipated
operational occurrences that may occur while in SLO.
Modifying the LHGR limits during SLO does not increase the
probability of an evaluated accident. The proposed 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.
Limits on the LHGR are specified to ensure that fuel design
limits are not exceeded anywhere in the core during normal
operation, including anticipated operational occurrences, and to
ensure that the PCT during a postulated design basis LOCA does not
exceed the limits specified in 10 CFR 50.46. This will ensure that
the fuel design safety criteria (i.e., less than 1% plastic strain
of the fuel cladding and no fuel centerline melting) are met and
that the core remains in a coolable geometry following a postulated
design basis LOCA. 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.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
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, including changes in
allowable modes of operation. The proposed change does not involve
any modifications of the plant configuration or allowable modes of
operation. Requiring the LHGR limits to be modified for SLO by
applying the SLO LHGR multiplier ensures that the assumptions of the
LOCA analyses are met.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed change will not adversely affect
operation of plant equipment. The change will not result in a change
to the setpoints at which protective actions are initiated. LHGR
limits during SLO are established to ensure that the PCT during a
postulated design basis LOCA does not exceed the limits specified in
10 CFR 50.46. This will ensure that the core remains in a coolable
geometry following a postulated design basis LOCA. The proposed
change will ensure the appropriate level of fuel protection.
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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Vice President,
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way,
Kennett Square, PA 19348.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: December 16, 2003.
Description of amendment request: The proposed amendment would
change Technical Specification (TS) Section 3/4.4.5, ``Reactor Coolant
System--Steam Generators,'' to allow a one-time extension of the steam
generator tube inservice inspection interval from March 2004 to March
2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their 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 steam generator tubes perform both an accident prevention
and an accident mitigation function. Steam generator tube integrity
is necessary to prevent the loss of reactor coolant system inventory
to the secondary system and to provide a barrier to fission product
release to the environment. The layup and storage conditions of the
steam generator during the extended outage have been assessed and
determined to not adversely affect steam generator conditions. An
operational assessment of the steam generators for approximately 1.4
effective full power year has been performed to assure acceptable
structural integrity during the extended surveillance interval. The
operational assessment for the steam generators has determined that
primary-to-secondary leakage following a steam line break, which is
the limiting event (other than a tube rupture), would continue to be
acceptable. 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 introduce any new or different
failure mechanism for the steam generators. Steam generator tube
integrity will be maintained as previously analyzed following
postulated events. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
[[Page 696]]
Response: No.
The layup and storage conditions of the steam generator during
the extended outage have been assessed and determined to not
adversely affect steam generator condition. The operational
assessment for the mid-cycle outage has shown that structural
margins are greater at approximately 1.4 EFPY then they would be at
the end of a typical full cycle of operation. Accident induced
leakage is projected to be the same for the surveillance interval
extension period as it would be for a full cycle of operation.
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: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: January 30, 2003.
Description of amendment request: This license amendment request
proposes a revision to the reactor pressure vessel (RPV) material
surveillance program described within the Perry Nuclear Power Plant
(PNPP) Updated Safety Analysis Report (USAR) from a plant-specific
program to the Boiling-Water Reactor Vessel and Internals Project
(BWRVIP) Integrated Surveillance Program (ISP).
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.
NRC [Nuclear Regulatory Commission] regulations impose
requirements upon the reactor coolant system to ensure that adequate
safety margins against nonductile or rapidly propagating failures
exits during normal operation, anticipated operational occurrences,
and system hydrostatic tests. These requirements are set forth in 10
CFR 50, Appendix A, ``General Design Criteria for Nuclear Power
Plants,'' Criterion 31, ``Fracture Prevention of Reactor Coolant
Pressure Boundary,'' Appendix G, ``Fracture Toughness
Requirements,'' and Appendix H requires that changes in the fracture
toughness properties of reactor vessel materials, resulting from the
neutron irradiation and the thermal environment, are monitored by a
material surveillance program. To determine the effects of neutron
fluence on the nil-ductility reference temperature of reactor vessel
materials, the methods provided in Regulatory Guide (RG) 1.99,
``Radiation Embrittlement of Reactor Vessel Materials,'' Revision 2
are used.
As described in the PNPP USAR, the current PNPP material
surveillance program is a plant-specific program which complies with
10 CFR 50, Appendix H.
The proposed amendment involves changing the material
surveillance program from a plant-specific program to an integrated
surveillance program. The use of an integrated program is consistent
with the requirements of 10 CFR 50, Appendix H, Paragraph III.C. The
integrated program proposed by PNPP is the BWRVIP ISP. The BWRVIP
ISP has been reviewed and approved by the NRC staff as an acceptable
program and is in conformance with 10 CFR 50, Appendix H. Use of the
ISP, among its many benefits, will increase the number of data
points used in the evaluation of changes in vessel material
properties. This will improve compliance with the aforementioned NRC
regulations. The methods contained in RG 1.99, Revision 2, will
still be used to determine the effects of neutron fluence upon the
nil-ductility reference temperature of the PNPP reactor vessel
materials.
This change will not affect the reactor pressure vessel, as no
physical changes are involved. The proposed change will not cause
the reactor pressure vessel or interfacing systems to be operated
outside of any design or testing limits. Furthermore, the proposed
changes will not alter any assumptions previously made in evaluating
the radiological consequences of any accident. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change revises the PNPP licensing bases to reflect
participation in the BWRVIP ISP. The ISP was approved by the NRC
staff as an acceptable material surveillance program which complies
with 10 CFR 50, Appendix H. The proposed change will not impact the
manner in which the plant is designed or operated. No new accident
types or failure modes will be introduced as a result of the
proposed change. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from that
previously evaluated.
3. The proposed change will not involve a significant reduction
in the margin of safety.
The material surveillance program requirements contained in 10
CFR 50, Appendix H, provide assurance that adequate margins of
safety exist for the reactor coolant system against nonductile or
rapidly propagating failures during normal operation, anticipated
operational occurrences, and safety hydrostatic tests. The BWRVIP
ISP has been approved by the NRC staff as an acceptable material
surveillance program which complies with 10 CFR 50, Appendix H. The
ISP will provide the material surveillance data which will ensure
that the safety margins require by NRC regulations are maintained
for the PNPP reactor coolant system. Therefore, the proposed change
does not involve a significant reduction in any margins 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: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: August 14, 2003.
Description of amendment request: This license amendment request
(LAR) proposes a revision to increase the analytical limit and the
resulting Technical Specification (TS) allowable value (AV) related to
the setpoint for the Main Steam Line Turbine Building Temperature--
High, system isolation function. This LAR revises the main steam line
trip setpoint AV based on improved computer modeling of the expected
building temperature transients in the event of a larger steam leak.
The proposed change improves the operating margins and reduces
challenges to the plant by avoiding unnecessary plant shutdown
transients from turbine building high temperatures from other than a
main steam line leak.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The nuclear boiler Leak Detection System (LDS) instrumentation
associated with the proposed amendment assists in the detection of a
small steam leak to prevent a significant release of radioactive
material created by conditions other than a break within the Reactor
Coolant Pressure Boundary (RCPB).
The proposed amendment establishes a new steam leak system
isolation temperature limit in the Turbine Building.
[[Page 697]]
There is no accident analysis or transient that credits the
subject LDS instrumentation. The subject instrumentation is for the
detection of small steam leaks and not a pipeline break as described
in the Updated Safety Analysis Report (USAR) Chapter 15 accident
analysis. The detection of main steam line flow is the parameter
used in the accident analysis to signal a steam line break outside
of containment.
The proposed amendment does not impact the physical design or
location of the LDS instrumentation. This proposed amendment is
associated only with the results of a main steam line leak in the
non-safety related Turbine Building and has no impact on the
initiation of this leak. The analysis completed in support of the
proposed amendment indicates that the radiological effects
associated with the new steam leak system isolation limit remains
bounded by the existing large main steam line break analysis
contained within the PNPP [Perry Nuclear Power Plant] USAR. The
proposed leakage limit does not alter the current function of the
LDS that isolates the Main Steam system prior to the leakage
degrading to a point where the system integrity, i.e., piping
integrity and makeup capability, is challenged. Therefore, the
proposed amendment ensures that the criteria for acceptance as
established in the original licensing bases and the requirements of
the original design basis remain valid. It has been determined that
the service life, i.e., Equipment Qualification (EQ) and structural
integrity of the Structures, Systems and Components (SSC) in the
affected areas are not adversely impacted by the proposed amendment.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed amendment does not impact the physical design or
location of the associated LDS instrumentation. The instruments will
still promptly initiate the automatic isolation of the appropriate
Containment and Drywell isolation valves to mitigate steam leakage
as credited in the original licensing bases. This proposed amendment
is associated only with the results of a main steam line break in
the non-safety related Turbine Building and has no impact on the
initiation of this leak. The analysis completed in support of the
proposed amendment indicates that the radiological effects
associated with the new steam leak system isolation limit remains
bounded by the existing large main steam line break analysis
contained within the PNPP USAR. The EQ and structural integrity of
any SSC located within the non-safety related Turbine Building are
not affected by the proposed amendment. Therefore, the proposed
amendment does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. The proposed change will not involve a single reduction in
the margin of safety.
The analysis performed for the proposed amendment proves that
the appropriate instruments will still promptly initiate automatic
system isolation, upon sensing temperatures in excess of their
setpoints. The radiological effects associated with the proposed
small steam leak to be detected remain bounded by the existing large
main steam line break analysis contained within the USAR. Steam
leaks in the affected area of the Turbine Building will be detected
on a timely basis so that the Main Steam system will be isolated
before such degradation could become sufficiently severe to
jeopardize the safety of the system. Also, steam leaks will be
detected before the leakage could increase to a level beyond the
capability of the makeup system. Therefore, the proposed amendment
ensures that the criteria for acceptance as established in the
original licensing bases and the requirements of the original design
basis remain valid. There is no accident analysis or transient that
credits the associated leak detection instrumentation, and the LDS
Main Steam Line Turbine Building Temperature--High function is
categorized as non-risk significant. Further, the proposed amendment
reduces the challenges to SSCs due to unnecessary plant shutdowns
created by conditions other than a main steam line leak. The EQ and
structural integrity of any SSC located within the Turbine Building
are not affected by the proposed amendment. Therefore, the proposed
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: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: October 29, 2003.
Description of amendment request: The proposed license amendments
would allow relocation of specific pressure and flow values for the
boric acid makeup (BAM) pumps, containment spray (CS) pumps, high
pressure safety injection (HPSI) pumps, and low pressure safety
injection (LPSI) pumps from the St. Lucie Units 1 and 2 Technical
Specifications to the Updated Final Safety Analysis Reports (UFSARs).
This is consistent with the Combustion Engineering Improved Standard
Technical Specifications and the Nuclear Regulatory Commission Final
Policy Statement on Technical Specification Improvements for Nuclear
Power Reactors.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Would operation of the facility in accordance with the
proposed amendments involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes to relocate the BAM, CS, HPSI, and LPSI
pump surveillance verification details in the aforementioned
Technical Specifications surveillance requirements to the St. Lucie
UFSARs do not adversely affect accident initiators or precursors nor
alter the design assumptions, conditions, configuration of the
facility, or the manner in which it is operated. The proposed
changes do not alter or prevent the ability of structures, systems,
or components to perform their intended function to mitigate the
consequences of an initiating event within the acceptance limits
assumed in the St. Lucie UFSARs.
The subject surveillance requirement criteria relocated to the
St. Lucie UFSARs will continue to be administratively controlled.
Changes to the St. Lucie UFSARs are evaluated and controlled under
10 CFR 50.59 prior to implementation. Therefore, the proposed
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
(2) Would operation of the facility in accordance with the
proposed amendments create the possibility of a new different kind
of accident from any accident previously evaluated?
The proposed changes do not alter the design assumptions,
conditions, or configuration of the facility or the manner in which
the plant is operated.
There are no changes to the source term or radiological release
assumptions used in evaluating the radiological consequences in the
St. Lucie UFSARs. The proposed changes have no adverse impact on the
component or system interactions. The proposed changes will not
adversely degrade the ability of systems, structures and components
important to safety to perform their safety function nor change the
response of any system, structure or component important to safety
as described in the UFSARs. The proposed changes do not change the
level of programmatic and procedural details of assuring operation
of the facility in a safe manner. Since there are no changes to the
design assumptions, conditions, configuration of the facility, or
the manner in which the plant is operated and surveilled, the
proposed changes do not create the possibility of a new different
kind of accident from any previously analyzed.
(3) Would operation of the facility in accordance with the
proposed amendments involve a significant reduction in a margin of
safety?
[[Page 698]]
There is no adverse impact on equipment design or operation and
there are no changes being made to the Technical Specification
required safety limits or safety system settings that would
adversely affect plant safety. The proposed changes do not reduce
the level of programmatic or procedural controls associated with the
activities presently performed via the aforementioned surveillance
requirements.
Future changes to the relocated technical requirements will
require an evaluation pursuant to the provisions of 10 CFR 50.59
prior to implementation.
Therefore, relocation of the specific pump pressure and flow
criteria contained in the aforementioned Technical Specification
Surveillance Requirements to the St. Lucie Units 1 and 2 UFSARs does
not involve a significant reduction in the margin of safety provided
in the existing specifications.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Allen G. Howe.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: November 21, 2003.
Description of amendment request: The proposed amendments would
transfer Technical Specification (TS) requirements 6.5 (Review and
Audit), 6.8.2 and 6.8.3 (procedures and programs review specifics), and
6.10 (Record Retention) to the quality assurance plan (a licensee
controlled document) for St. Lucie Units 1 and 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. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes to the St. Lucie Plant TS do not adversely
affect accident initiators or precursors, nor alter the design
assumptions, conditions, and configuration of the facility or the
manner in which the plant is operated and maintained. In addition,
the proposed changes do not affect the manner in which the plant
responds in normal operation, transient, or accident conditions, nor
do they change any of the procedures related to operation of the
plant. The proposed changes do not alter or prevent the ability of
structures, systems, and components (SSCs) to perform their intended
function to mitigate the consequences of an initiating event within
the acceptance limits assumed in the Updated Final Safety Analysis
Report (UFSAR). The proposed changes are administrative for the
purpose of updating TS to reflect current NRC and industry
initiatives.
The proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of an accident previously evaluated in
the St. Lucie UFSARs. Further, the proposed changes do not increase
the types and amounts of radioactive effluent that may be released
off site, nor significantly increase individual or cumulative
occupational/public radiation exposures.
Therefore, it is concluded that these proposed revisions do not
involve a significant increase in the probability or consequence of
an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed changes to the St. Lucie Plant TS do not change the
operation or the design basis of any plant system or component
during normal or accident conditions. The proposed changes do not
include any physical changes to the plant. In addition, the proposed
changes do not change the function or operation of plant equipment
or introduce any new failure mechanisms. The plant equipment will
continue to respond per the design and analyses and there will not
be a malfunction of a new or different type introduced by the
proposed changes.
The proposed changes are administrative in nature and only
correct, update and clarify the St. Lucie Plant Technical
Specifications to reflect NRC guidance, i.e., AL 95-06. The proposed
changes do not modify the facility nor do they affect the plant's
response to normal, transient, or accident conditions. The changes
do not introduce a new mode of plant operation. The changes are an
enhancement and do not affect plant safety. The plant's design and
design basis are not revised and the current safety analyses remains
in effect.
Thus, these proposed revisions to the St Lucie Plant TS do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The proposed changes are administrative changes to the St. Lucie
Plant Technical Specifications. The safety margins established
through Limiting Conditions for Operation, Limiting Safety System
Settings and Safety Limits as specified in the Technical
Specifications are not revised nor is the plant design or its method
of operation revised by the proposed changes.
Thus, it is concluded that these proposed revisions to the St.
Lucie Plant TS 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Allen G. Howe.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: December 6, 2003.
Description of amendment request: The proposed amendment would
revise the Unit 1 and 2 Technical Specifications (TSs) by adding a
requirement to apply linear heat generation rate [LHGR] limits if the
main turbine bypass system becomes inoperable.
Basis for proposed no significant hazards consideration
determination:
As required by 10 CFR 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration, which is
presented below:
1. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
Response: No.
The proposed change to the TS 3.7.6 does not directly or
indirectly affect any plant system, equipment, component, or change
the processes used to operate the plant. Further, the MCPR [minimum
critical power ratio] and LHGR limits documented in the unit/cycle
specific COLRs [core operating limits report] for Main Turbine
Bypass System operable and inoperable are generated using NRC
[Nuclear Regulatory Commission] approved methodology and meet the
applicable acceptance criteria. The COLR operating limits thus
assure that the MCPR Safety Limit and LHGR Limit will not be
exceeded during normal operation or anticipated operational
occurrences. Thus, this proposed amendment does not involve a
significant increase in the probability of occurrence 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 change to TS 3.7.6 does not directly or indirectly
affect any plant system, equipment, or component and therefore does
not affect the failure modes of any of these items. Thus, the
proposed changes do not create the possibility of a previously
unevaluated operator error or a new single failure.
[[Page 699]]
Therefore, this proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Since the proposed changes do not alter any plant system,
equipment, component, or the processes used to operate the plant,
the proposed change will not jeopardize or degrade the function or
operation of any plant system or component governed by Technical
Specifications. The proposed change to TS 3.7.6 does not involve a
significant reduction in the margin of safety as currently defined
in the Bases of the applicable Technical Specification sections,
because the MCPR and LHGR limits calculated for Main Turbine Bypass
System operable and inoperable preserve the required margin of
safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc, General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101,1179.
NRC Section Chief: Richard J. Laufer.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: October 22, 2003 (TSC 03-12).
Description of amendment request: The proposed change involves the
extension from 1 hour to 24 hours of the completion time (CT) for
Action (a) of Technical Specification (TS) 3.5.1.1, which defines
requirements for accumulators. Accumulators are part of the emergency
core cooling system and consist of tanks partially filled with borated
water and pressurized with nitrogen gas. The contents of the tank are
discharged to the reactor coolant system (RCS) if, as during a loss-of-
coolant accident, the coolant pressure decreases to below the
accumulator pressure. Action (a) of TS 3.5.1.1 specifies a CT to
restore an accumulator to operable status when it has been declared
inoperable for a reason other than the boron concentration of the water
in the accumulator not being within the required range. This change was
proposed by the Westinghouse Owners Group participants in the TS Task
Force (TSTF) and is designated TSTF-370. TSTF-370 is supported by NRC-
approved topical report WCAP-15049-A, ``Risk-Informed Evaluation of an
Extension to Accumulator Completion Times,'' submitted on May 18, 1999.
The NRC staff issued a notice of opportunity for comment in the Federal
Register on July 15, 2002 (67 FR 46542), on possible amendments
concerning TSTF-370, including a model safety evaluation and model no
significant hazards consideration (NSHC) determination, using the
consolidated line item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on March 12,
2003 (68 FR 11880). The licensee affirmed the applicability of the
following NSHC determination in its application dated October 22, 2003.
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 basis for the accumulator limiting condition for operation
(LCO), as discussed in Bases Section 3.5.1.1, is to ensure that a
sufficient volume of borated water will be immediately forced into
the core through each of the cold legs in the event the RCS pressure
falls below the pressure of the accumulators, thereby providing the
initial cooling mechanism during large RCS pipe ruptures. As
described in Section 9.2 of WCAP-15049-A, the proposed change will
allow plant operation with an inoperable accumulator for up to 24
hours, instead of 1 hour, before the plant would be required to
begin shutting down. The impact of the increase in the accumulator
CT on core damage frequency for all the cases evaluated in WCAP-
15049-A is within the acceptance limit of 1.0E-06/yr for a total
plant core damage frequency (CDF) less than 1.0E-03/yr. The
incremental conditional core damage probabilities calculated in
WCAP-15049-A for the accumulator CT increase meet the criterion of
5E-07 in Regulatory Guides (RG) 1.174, ``An Approach for using
Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-
Specific Changes to the Licensing Basis,'' and 1.177, ``An Approach
for Plant-Specific, Risk-Informed Decisionmaking: Technical
Specifications,'' for all cases except those that are based on
design basis success criteria. As indicated in WCAP-15049-A, design
basis accumulator success criteria are not considered necessary to
mitigate large break loss-of-coolant accident (LOCA) events, and
were only included in the WCAP-15049-A evaluation as a worst case
data point. In addition, WCAP-15049-A states that the NRC has
indicated that an incremental conditional core damage frequency
(ICCDP) greater than 5E-07 does not necessarily mean the change is
unacceptable.
The proposed technical specification change does not involve any
hardware changes nor does it affect the probability of any event
initiators. There will be no change to normal plant operating
parameters, engineered safety feature (ESF) actuation setpoints,
accident mitigation capabilities, accident analysis assumptions or
inputs.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed change. As described in Section 9.1 of the WCAP-
15049-A evaluation, the plant design will not be changed with this
proposed technical specification CT increase. All safety systems
still function in the same manner and there is no additional
reliance on additional systems or procedures. The proposed
accumulator CT increase has a very small impact on core damage
frequency. The WCAP-15049-A evaluation demonstrates that the small
increase in risk due to increasing the CT for an inoperable
accumulator is within the acceptance criteria provided in RGs 1.174
and 1.177. No new accidents or transients can be introduced with the
requested change and the likelihood of an accident or transient is
not impacted.
The malfunction of safety related equipment, assumed to be
operable in the accident analyses, would not be caused as a result
of the proposed technical specification change. No new failure mode
has been created and no new equipment performance burdens are
imposed.
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not involve a significant reduction in
a margin of safety. There will be no change to the departure from
nucleate boiling ratio (DNBR) correlation limit, the design DNBR
limits, or the safety analysis DNBR limits.
The basis for the accumulator LCO, as discussed in Bases Section
3.5.1.1, is to ensure that a sufficient volume of borated water will
be immediately forced into the core through each of the cold legs in
the event the RCS pressure falls below the pressure of the
accumulators, thereby providing the initial cooling mechanism during
large RCS pipe ruptures. As described in Section 9.2 of WCAP-15049-
A, the proposed change will allow plant operation with an inoperable
accumulator for up to 24 hours, instead of 1 hour, before the plant
would be required to begin shutting down. The impact of this on
plant risk was evaluated and found to be very small. That
[[Page 700]]
is, increasing the time the accumulators will be unavailable to
respond to a large LOCA event, assuming accumulators are needed to
mitigate the design basis event, has a very small impact on plant
risk.
Since the frequency of a design basis large LOCA (a large LOCA
with loss of offsite power) would be significantly lower than the
large LOCA frequency of the WCAP-15049-A evaluation, the impact of
increasing the accumulator CT from 1 hour to 24 hours on plant risk
due to a design basis large LOCA would be significantly less than
the plant risk increase presented in the WCAP-15049-A evaluation.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: December 8, 2003.
Description of amendment request: The licensee is proposing to
revise Technical Specification (TS) Section 5.5.6, ``Containment Tendon
Surveillance Program,'' for consistency with the requirements of 10 CFR
50.55a(g)(4) for components classified as Code Class CC. The proposed
revision to TS 5.5.6 is to indicate that the Containment Tendon
Surveillance Program, inspection frequencies, and acceptance criteria
shall be in accordance with Section XI, Subsection IWL of the ASME
Boiler and Pressure Vessel Code and the applicable addenda as required
by 10 CFR 50.55a, except where an exemption or relief has been
authorized by the NRC. The licensee has also proposed to delete the
provisions of Surveillance Requirement 3.0.2 from this TS. In addition,
the licensee is proposing to revise TS 5.5.16, ``Containment Leakage
Rate Testing Program,'' to add exceptions to Regulatory Guide 1.163,
``Performance-Based Containment Leak-Testing Program.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change revises the TS administrative controls
programs for consistency with the requirements of 10 CFR
50.55a(g)(4) for components classified as Code Class CC. The revised
requirements do not affect the function of the containment post-
tensioning system components. The post-tensioning systems are
passive components whose failure modes could not act as accident
initiators or precursors.
The proposed change affects the frequency of visual examinations
that will be performed for the concrete surfaces of the containment
for the purpose of the Containment Leakage Rate Testing Program. In
addition, the proposed change allows those examinations to be
performed during power operation as opposed to during a refueling
outage. The frequency of visual examinations of the concrete
surfaces of the containment and the mode of operation during which
those examinations are performed has no relationship to or adverse
impact on the probability of any of the initiating events assumed in
the accident analyses. The proposed change would allow visual
examinations that are performed pursuant to NRC approved ASME
Section XI Code requirements (except where relief has been granted
by the NRC) to meet the intent of visual examinations [as] required
by Regulatory Guide 1.163, [because of the commitment in Appendix 3A
of the Callaway Final Safety Analysis Report,] without requiring
additional visual examinations pursuant to the Regulatory Guide. The
intent of early detection of deterioration will continue to be met
by the more rigorous requirements of the Code required visual
examinations. As such, the safety function of the containment as a
fission product barrier is maintained.
The proposed change does not impact any accident initiators or
analyzed events or assumed mitigation of accident or transient
events. They do not involve the addition or removal of any
equipment, or any design changes to the facility.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change revises the TS administrative controls
programs for consistency with the requirements of 10 CFR
50.55a(g)(4) for components classified as Code Class CC. The
function of the containment post-tensioning system components are
not altered by this change. The change affects the frequency of
visual examinations that will be performed for the concrete surfaces
[of the containment]. In addition, the proposed change allows those
examinations to be performed during power operation as opposed to
during a refueling outage. The proposed change does not involve a
modification to the physical configuration of the plant (i.e., no
new equipment will be installed) or change in the methods governing
normal plant operation. The proposed change will not impose any new
or different requirements or introduce a new accident initiator,
accident precursor, or malfunction mechanism. Additionally, there is
no change in the types or increases in the amounts of any
effluent[s] that may be released off-site and there is no increase
in individual or cumulative occupational exposure.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change revises the TS administrative controls
programs for consistency with the requirements of 10 CFR
50.55a(g)(4) for components classified as Code Class CC. The
function of the containment post-tensioning system components are
not altered by this change. The change affects the frequency of
visual examinations that will be performed for the concrete surfaces
[of the containment]. In addition, the proposed change allows those
examinations to be performed during power operation as opposed to
during a refueling outage. The safety function of the containment as
a fission product barrier will be maintained.
Therefore, 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: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: October 30, 2003.
Description of amendment request: The amendment would revise the
Technical Specifications (TSs) for alternating current (AC) sources--
operating (TS 3.8.1) and electrical power distribution systems--
operating (TS 3.8.9) by extending the required action completion times
(CTs). For TS 3.8.1, the amendment would extend the CT to restore a
single inoperable diesel generator (DG) to operable status by adding a
note to the CT for Required Action B.4. A note would also be added to
the CT for Required Action A.3 to restore a single inoperable offsite
circuit to operable status to account for the note that would be added
to the CT for Required Action B.4.
For TS 3.8.9, the CT for Required Action C.1 (to restore a single
inoperable AC vital bus subsystem to
[[Page 701]]
operable status) would be extended to 24 hours. The second CTs, from
the discovery of the failure to meet the limiting condition for
operation (LCO), for Required Actions B.1 (to restore a single
inoperable AC electrical power distribution subsystem to operable
status), C.1 (given above), and D.1 (to restore a single inoperable
direct current (DC) electrical power distribution subsystem to operable
status) would be extended to 34 hours.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to the Completion Times do not change the
response of the plant to any accidents and have an insignificant
impact on the reliability of the electrical power sources and
distribution systems. The proposed changes to the second Completion
Times are administrative in nature and only intended to prevent the
plant from successively entering and exiting ACTIONS associated with
different systems governed by one LCO without ever meeting the LCO.
The electrical power sources and distribution subsystems will remain
highly reliable and the proposed changes will not result in a
significant increase in the risk of plant operation. This is
demonstrated by showing that the impact on plant safety as measured
by core damage frequency (CRF) and large early release frequency
(LERF) is acceptable. In addition, for the Completion Time change,
the incremental conditional core damage probabilities (ICCDP) and
incremental conditional large early release probabilities (ICLERP)
are also acceptable. These changes are consistent with the
acceptance criteria in Regulatory Guides 1.174 and 1.177. Therefore,
since the electrical sources and distribution subsystems will
continue to perform their [safety] functions with high reliability
as originally assumed, and the increase in risk as measured by CDF,
LERF, ICCDP, [and] ICLERP is acceptable, there will not be a
significant increase in the consequences of any accidents
[previously evaluated].
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, or
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not alter or
prevent the ability of structures, systems, and components (SSCs)
from performing their intended [safety] function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed changes do not affect the source term,
containment isolation, or radiological release assumptions used in
evaluating the radiological consequences of an accident previously
evaluated. Further, the proposed changes do not increase the types
or amounts of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures. The proposed changes are consistent with the
safety analysis assumptions and resultant [radiological]
consequences.
Therefore, the proposed change[s do] not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not result in a change in the manner in
which the electrical distribution subsystems provide plant
protection. The use of the Sharpe Station will provide an alternate
AC power source in the event of emergent inoperability of a WCGS
[Wolf Creek Generating Station] DG or a complete loss of all WCGS
emergency AC power. The changes do not alter assumptions made in the
safety analysis. The changes to Completion Times do not change any
existing accident scenarios, nor create any new or different
accident scenarios. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
Therefore, the proposed change[s do] 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 a margin of safety.
The proposed changes do 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 these changes. The proposed changes will not
result in plant operation in a configuration outside the design
basis. The calculated impact on risk is insignificant and is
consistent with the acceptance criteria contained in Regulatory
Guides 1.174 and 1.177.
Therefore, the proposed change[s 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: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
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, located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendment: November 8, 2002.
Brief description of amendment: The amendment revised the Technical
Specifications to delete the requirements for the auxiliary and fuel
handling building air treatment system.
Date of issuance: December 12, 2003.
[[Page 702]]
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 248.
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 24, 2002 (67
FR 78517).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 12, 2003.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: May 28, 2003, as supplemented
October 8, 2003.
Brief description of amendment: This amendment eliminates the need
to credit Boraflex neutron-absorbing material for reactivity control in
the spent fuel storage pool.
Date of issuance: December 22, 2003.
Effective date: December 22, 2003.
Amendment No.: 198.
Facility Operating License No. DPR-23: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 8, 2003 (68 FR
40710). The October 8, 2003, supplement contained clarifying
information only and did not change the initial proposed no significant
hazards consideration determination or expand the scope of the initial
application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 22, 2003.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: December 11, 2002, as
supplemented June 24, 2003.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) related to N-1 loop operation. Specifically, the
changes eliminate N-1 loop operation from particular sections of the
TSs and makes other changes that are clarifying and/or administrative
in nature. In addition, the TS Bases are revised to address the
proposed changes.
Date of issuance: December 10, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 217.
Facility Operating License No. NPF-49: The amendment revised the
TSs.
Date of initial notice in Federal Register: January 21, 2003 (68 FR
2800). The January 24, 2003, supplement contained clarifying
information and did not change the staff's proposed finding of no
significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 10, 2003.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-413 and 50-414, Catawba Nuclear
Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: March 24, 2003, as supplemented
by letters dated June 25 and October 15, 2003.
Brief description of amendments: The amendments revise the
Technical Specifications (TS) to relocate certain reactor coolant
system cycle-specific parameter limits from the TSs to the Core
Operating Limits Report, and revises the minimum allowable reactor
coolant system flow rate.
Date of issuance: December 19, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 210 and 204.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 18, 2003 (68
FR 54749), November 18, 2003 (68 FR 65090).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 19, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: May 19, 2003, as supplemented
October 27, 2003.
Brief description of amendment: This amendment revises the
Technical Specifications to require ``flow indication,'' rather than
``safety-grade flow indication,'' to satisfy Surveillance Requirement
4.7.1.7.e.2 for the motor driven feedwater pump.
Date of issuance: December 18, 2003.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 261.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 10, 2003 (68 FR
34669).
The supplement dated October 27, 2003, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated December 18, 2003.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: July 15, 2003.
Brief description of amendment: The amendment revises the Technical
Specification (TS) requirements for surveillance of the status of
Secondary Containment Isolation Valves and Blind Flanges in
Surveillance Requirement 3.6.4.2.1, consistent with TS Task Force
Traveler-45 Revision 2.
Date of issuance: December 5, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 202.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 28, 2003 (68 FR
61479).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 5, 2003.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of application for amendment: April 11, 2003.
Brief description of amendment: The amendment revises TS Section
5.0, ``Administrative Controls,'' to make various administrative,
editorial, and typographical changes.
Date of issuance: December 15, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
[[Page 703]]
Amendment No.: 213.
Facility Operating License No. DPR-20: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2003 (68
FR 64136).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 15, 2003.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket No. 50-206, San
Onofre Nuclear Generating Station, Unit 1, San Diego County, California
Date of application for amendment: July 25, 2003, supplemented by
letters dated October 3, 2003, and December 3, 2003.
Brief description of amendment: This amendment approves the use of
the modified Unit 1 turbine gantry crane and turbine building support
structure in a single failure proof application and at a rated capacity
of 105 tons for handling of spent fuel casks as documented in the
Defueled Safety Analysis Report (DSAR). The DSAR changes approved by
this amendment are needed to permit use of the modified turbine gantry
crane and turbine building support structure for lifting and handling
of the spent fuel casks from the SONGS Unit 1 spent fuel pool to the
Independent Spent Fuel Storage Installation.
Date of issuance: December 18, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: Unit 1-162.
Facility Operating License No.DPR-13: Amendment revises the license
to permit use of the turbine building gantry crane in a single failure
proof application at a rated capacity of 105 tons for handling of spent
fuel casks.
Date of initial notice in Federal Register: September 18, 2003 (68
FR 54751).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 18, 2003.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: December 19, 2002, as
supplemented October 20, 2003.
Brief description of amendments: These amendments correct various
typographical, editorial, and other administrative errors currently in
the Technical Specifications for Surry Power Station, Units 1 and 2.
Date of issuance: December 16, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 238 and 237.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments change the Technical Specifications.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5684). The supplement dated October 20, 2003, provided clarifying
information only and did not change the initial proposed no significant
hazards consideration determination or expand the scope of the initial
application.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 16, 2003.
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 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action, see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety
[[Page 704]]
Evaluation and/or Environmental Assessment, as indicated. All of these
items are available for public inspection at the Commission's Public
Document Room, 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 Assess and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by e-mail to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By February 5, 2004, 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 2.714, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, and
electronically on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are problems in accessing the
document, contact the PDR Reference staff at 1-800-397-4209, 301-415-
4737, or by e-mail to [email protected]. If a request for a hearing or
petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of 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. Petitioner must provide 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 to relief. A
petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross-
examine witnesses. 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 with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of the continuing disruptions in delivery of mail to United
States Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the petition for
leave to intervene and request for hearing should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and because of continuing disruptions in
delivery of mail to United States Government offices, it is requested
that copies be transmitted either by means of facsimile transmission to
301-415-3725 or by e-mail to [email protected]. A copy of the
request for hearing and petition for leave to intervene should also be
sent to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
STP Nuclear Operating Company, Docket No. 50-499, South Texas Project,
Unit 2, Matagorda County, Texas
Date of amendment request: December 23, 2003.
Description of amendment request: The amendments revise Technical
Specification (TS) 3.8.1, ``AC Sources--Operating,'' to extend the
allowed outage time for Unit 2 Standby Diesel Generator 22 from 14 days
to 21 days as a one-time change for the purpose of collecting data
associated with failure of SDG-22.
Date of issuance: December 23, 2003.
Effective date: December 23, 2003.
Amendment Nos.: Unit No. 2: 148.
Facility Operating License Nos. NPF-76 and NPF-80: Amendments
revise the Technical Specifications.
[[Page 705]]
Public comments requested as to final 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 December 23,
2003.
Attorney for licensee: A. H. Gutterman, Esquire, Morgan, Lewis &
Bockius, LLP, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: Robert A. Gramm.
Dated at Rockville, Maryland, this 24th day of December, 2003.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 04-8 Filed 1-5-04; 8:45 am]
BILLING CODE 7590-01-P