[Federal Register Volume 69, Number 12 (Tuesday, January 20, 2004)]
[Notices]
[Pages 2735-2752]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-1104]
-----------------------------------------------------------------------
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 24, 2003, through January 8, 2004.
The last biweekly notice was published on January 6, 2003 (69 FR 691).
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 19, 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. 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
[[Page 2736]]
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, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: May 1, 2003.
Description of amendment request: The proposed amendment would
revise the Clinton Power Station (CPS) Technical Specifications to (1)
support an expansion of the core flow operating range, (2) implement an
Oscillation Power Range Monitor (OPRM) Instrumentation system, and (3)
implement the Detect and Suppress Solution--Confirmation Density
approach to automatically detect and suppress neutronic/thermal-
hydraulic instabilities. These changes will support operation at 3,473
megawatts thermal with core flow as low as 85 percent of rated core
flow. The expanded operating range is identified as Maximum Extended
Load Line Limit Analysis Plus (MELLLA+). The scope of evaluations
required to support the expansion of the core flow operating range to
MELLLA+ boundary is contained in the General Electric Licensing Topical
Report (LTR) NEDC-33006P, ``Maximum Extended Load Line Limit Analysis
Plus Licensing Topical Report.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The probability (frequency of occurrence) of a design basis
accident (DBA) occurring is not affected by the operating range
expansion, because the plant continues to comply with the regulatory
and design basis criteria established for plant equipment. The
MELLLA+ core operating range expansion does not require significant
plant hardware modifications. The core operating range expansion
involves changes to the operating power-to-flow map and a small
number of setpoints and alarms. Because there is no change in the
operating pressure, power, steam flow rate, or feedwater flow rate,
there are no significant effects on the plant hardware outside of
the Nuclear Steam Supply System (NSSS). The MELLLA+ operating range
expansion does not cause additional requirements to be imposed on
any of the safety, balance-of-plant, electrical, or auxiliary
systems. No changes to the power generation and electrical
distribution systems are required due to the introduction of
MELLLA+. An evaluation of the probabilistic safety assessment
concludes that the calculated increase in core damage frequencies
due to the MELLLA+ operating range expansion are very small. Scram
setpoints (equipment settings that initiate automatic plant
shutdowns) are established such that there is no significant
increase in scram frequency due to the MELLLA+ operating range
expansion. No new challenges to safety-related equipment result from
the MELLLA+ operating range
[[Page 2737]]
expansion. As a result, there is no significant increase in the
probability of an accident previously evaluated.
The proposed changes specify limiting conditions for operation,
required actions and surveillance requirements for the OPRM system,
and allows operation in regions of the power-to-flow map currently
restricted by the requirements of the Interim Corrective Actions
(ICAs) and certain limiting conditions of operation of TS Section
3.4.1. The restrictions of the ICAs and TS Section 3.4.1 were
imposed to ensure adequate capability to detect and suppress
conditions consistent with the onset of thermal-hydraulic
oscillations that may develop into a thermal-hydraulic instability
event. A thermal-hydraulic instability event has the potential to
challenge the Minimum Critical Power Ratio (MCPR) safety limit. The
OPRM system can automatically detect and suppress conditions
necessary for thermal-hydraulic instability. The Backup Stability
Protection (BSP), in lieu of the ICAs, will provide adequate
protection should the OPRM equipment become temporarily inoperable.
With the activation of the OPRM system, the restrictions of the ICAs
and TS Section 3.4.1 will no longer be required.
The probability of a thermal-hydraulic instability event is
impacted by power to flow conditions such that only during operation
inside specific regions of the power-to-flow map, in combination
with power shape and inlet enthalpy conditions, can the occurrence
of an instability event be postulated to occur. Operation in these
regions may increase the probability that operation with conditions
necessary for a thermal-hydraulic instability can occur.
When the OPRM is operable, the OPRM can automatically detect the
imminent onset of power oscillations and generate a trip signal.
Actuation of a Reactor Protection System (RPS) trip will suppress
conditions necessary for thermal-hydraulic instability and decrease
the probability of a thermal-hydraulic instability event. In the
event the trip capability of the OPRM is not maintained, the
proposed changes limit the period of time before an alternate method
to detect and suppress thermal-hydraulic oscillations is required.
Since the duration of this period of time is limited, the increase
in the probability of a thermal-hydraulic instability event is not
significant. Therefore, the proposed changes do not result in a
significant increase in the probability of an accident previously
evaluated.
The DSS-CD solution is designed to identify power oscillations
upon inception and initiate control rod insertion (i.e., scram) to
terminate the oscillations prior to any significant amplitude
growth. The DSS-CD provides protection against violation of the
Safety Limit Minimum Critical Power Ratio (SLMCPR) for anticipated
oscillations. Compliance with Criterion 10, ``Reactor design.'', and
Criterion 12, ``Suppression of reactor power oscillations.'', of
10CFR50, Appendix A, ``General Design Criteria For Nuclear Power
Plants,'' is accomplished via an automatic action. A developing
instability event is suppressed by the DSS-CD system with
substantial margin to the SLMCPR and no clad damage, with the event
terminating in a scram and never developing into an accident. The
DSS-CD system does not interact with equipment whose failure could
cause an accident. Scram setpoints in the DSS-CD will be established
so that analytical limits are met. The reliability of the DSS-CD
will meet or exceed that of the existing system. No new challenges
to safety-related equipment will result from the DSS-CD solution.
Because an instability event would reliably terminate in an early
scram without impact on other safety systems, there is no
significant increase in the probability of an accident.
The spectrum of hypothetical accidents and transients has been
investigated, and are shown to meet the plant's currently licensed
regulatory criteria. In the area of core design, for example, the
fuel operating limits such as Maximum Average Planar Linear Heat
Generation Rate (MAPLHGR) and SLMCPR continue to be met. The fuel
reload analyses will show plant transients meet the criteria
accepted by the NRC as specified in NEDO-24011, ``GESTAR II,''
(Reference 12). Challenges to fuel are evaluated, and shown to still
meet the criteria of 10 CFR 50.46, ``Acceptance Criteria for
Emergency Core Cooling Systems for Light-Water Nuclear Power
Reactors.'', 10 CFR 50 Appendix K, ``ECCS Evaluation Models,'' and
Regulatory Guide 1.70, ``Standard Format and Content of Safety
Analysis Reports for Nuclear Power Plants,'' Section 6.3. Challenges
to the containment have been evaluated, and the containment and its
associated cooling systems meet Criterion 38, ``Containment heat
removal.'', and Criterion 50, ``Containment design basis.'', of the
general design criteria. Radiological release events have been
evaluated, and are shown to be below the regulatory limits of 10 CFR
100, ``Reactor Site Criteria''. Operation in the MELLLA+ region does
not result in an increase in the consequences of an accident
previously evaluated. Operation within the MELLLA+ region has been
evaluated to ensure that the CPS response to accidents and
transients remains within acceptable criteria. Thus, the proposed
changes do not involve a significant increase in the consequences of
an accident previously evaluated.
An unmitigated thermal-hydraulic instability event is postulated
to cause a violation of the MCPR safety limit. The proposed changes
ensure mitigation of thermal-hydraulic instability events prior to
challenging the MCPR safety limit if initiated from anticipated
conditions by detection of the onset of oscillations and actuation
of an RPS trip signal when the OPRM system is operable. The OPRM
also provides the capability of an RPS trip being generated for
thermal-hydraulic instability events initiated from unanticipated
but postulated conditions. These mitigative capabilities of the OPRM
system would become available as a result of the proposed changes
and have the potential to reduce the consequences of unanticipated
and postulated thermal-hydraulic instability events.
As stated above, the DSS-CD solution meets the requirements of
Criterion 10 and Criterion 12 of the GDC by automatically detecting
and suppressing design basis thermal-hydraulic oscillations prior to
exceeding the fuel SLMCPR. Proper operation of the DSS-CD system
does not affect any fission product barrier or Engineered Safety
Feature. Thus, the proposed change cannot change the consequences of
any accident previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any previously evaluated?
Response: No.
Equipment that could be affected by MELLLA+ has been evaluated
and no new operating mode, safety related equipment lineup, accident
scenario, or equipment failure mode was identified. The full
spectrum of accident considerations, defined in the CPS Updated
Safety Analysis Report (USAR), has been evaluated, and no new or
different kind of accident has been identified. The MELLLA+
operating range expansion uses existing technology and NRC approved
safety analysis methodology, and applies them within the
capabilities of already existing plant equipment in accordance with
presently existing regulatory and industry criteria. The MELLLA+
operating range expansion will not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes specify limiting conditions for operations,
required actions and surveillance requirements of the OPRM system
and allows operation in regions of the power-to-flow map currently
restricted by the requirements of the ICAs and TS Section 3.4.1. The
OPRM system uses input signals shared with the Average Range Power
Monitor (APRM) system and rod block functions to monitor core
conditions and generate an RPS trip when required. Quality
requirements for software design, testing, implementation and module
self-testing of the OPRM system provide assurance that no new
equipment malfunctions due to software errors are created. The
design of the OPRM system also ensures that neither operation nor
malfunction of the OPRM system will adversely impact the operation
of the other systems and no accident or equipment malfunction of
these other systems could cause the OPRM system to malfunction or
cause a different kind of accident. No new failure modes of either
the new OPRM equipment or of the existing APRM equipment have been
introduced. Therefore, operation with the OPRM system does not
create the possibility of a new or different kind of accident from
any previously evaluated.
The DSS-CD solution operates within the existing Option III OPRM
hardware. Implementation of the DSS-CD will require a software/
hardware change to the existing Option III system. No new operating
mode, safety-related equipment lineup, accident scenario, system
interaction, or equipment failure mode was identified. Therefore,
the DSS-CD solution will not adversely affect plant equipment.
Because there are no significant hardware changes, there is no
[[Page 2738]]
change in the possibility or consequences of a failure. The worst-
case failure of the equipment is a failure to initiate mitigating
action (i.e., scram), but no failure can cause an accident of a new
or different kind than any previously evaluated.
As such the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The calculated loads on all affected structures, systems and
components have been shown to remain within design allowables for
all design basis event categories. No NRC acceptance criteria are
exceeded. The margins of safety currently included in the design of
the plant are not affected by the MELLLA+ operating range expansion.
Because the plant configuration and response to transients and
hypothetical accidents do not result in exceeding the presently
approved NRC acceptance limits, operation in the MELLLA+ region does
not involve a significant reduction in a margin of safety.
The OPRM system monitors small groups of LPRM signals for
indication of local variations of core power consistent with
thermal-hydraulic oscillations and generates an RPS trip when
conditions consistent with the onset of oscillations are detected.
An unmitigated thermal-hydraulic instability event has the potential
to result in a challenge to the MCPR safety limit. The OPRM system
provides the capability to automatically detect and suppress
conditions which might result in a thermal-hydraulic instability
event and thereby maintains the margin of safety by providing
automatic protection for the MCPR safety limit while reducing the
burden on the control room operators significantly. The BSP, in lieu
of the ICAs, will provide adequate protection should the OPRM
equipment become temporarily inoperable. Operation with the OPRM
system does not involve a significant reduction in a margin of
safety.
The DSS-CD solution is designed to identify the power
oscillations upon inception and initiate control rod insertion to
terminate (i.e., scram) the oscillations prior to any significant
amplitude growth. The DSS-CD solution algorithm will maintain or
increase the margin to the SLMCPR for anticipated instability
events. The safety analyses in NEDC-33075P demonstrate the margin to
the SLMCPR for postulated bounding stability events. In addition,
the current Option III algorithms are retained to provide defense-
in-depth protection for unanticipated reactor instability events. As
a result, there is no impact on the MCPR Safety Limit identified for
an instability event.
Therefore, operation of CPS in accordance with the proposed
changes will 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: Edward J. Cullen, Deputy General Counsel
Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: December 23, 2003.
Description of amendment request: The licensee proposed to revise
Section 3.4.A and 3.5.A.2 of the Technical Specifications to clarify
requirements for inoperable components and allow meeting the water
availability requirements during periods of core spray system
inoperability (e.g., when the plant is shutdown) in an alternate
manner. Specifically, this would allow the required water volume for
core spray system operability be located in the torus, condensate
storage tank, or a combination of both, in order to provide operational
flexibility in water management and outage work scheduling.
Additionally, the licensee proposed to improve consistency of
verification requirements within the specifications and provide more
definitive bases for the specifications. No physical changes to the
plant are involved, and the requirements in the current specifications
will be maintained.
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. The NRC staff has reviewed the licensee's analysis
against the three standards of 10 CFR 50.92(c). The NRC staff's
analysis is presented below:
The first standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
increase in the probability or consequences of an accident previously
evaluated. The proposed changes will be made in a manner such that the
current requirements are maintained for the core spray system. The
source of core spray water was not considered as a precursor of any
previously analyzed and evaluated accident. No hardware design change
is involved with the proposed amendment. Thus, the proposed amendment
would create no adverse effect on the functional performance of any
plant structure, system, or component (SSC). All SSCs will continue to
perform their design functions with no decrease in their capabilities
to mitigate the previously analyzed consequences of postulated
accidents. Accordingly, the revised specifications will lead to no
increase in the consequences of an accident previously evaluated, and
no increase of the probability of an accident previously evaluated.
The second standard requires that operation of the unit in
accordance with the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated. The proposed amendment is not the result of a hardware
design change, nor does it lead to the need for a hardware design
change. There is no change in the methods the unit is operated. As a
result, all SSCs will continue to perform as previously analyzed by the
licensee, and previously evaluated and accepted by the NRC staff.
Therefore, the proposed amendment will not create the possibility of a
new or different kind of accident from any previously evaluated.
The third standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
reduction in a margin of safety. Since the licensee did not propose to
exceed or alter a design basis or safety limit, and did not propose to
operate any component in a less conservative manner, the proposed
amendment will not affect in any way the performance characteristics
and intended functions of any SSC. Therefore, the proposed amendment
does not involve a significant reduction in a margin of safety.
Based on the NRC staff's analysis, 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: Kevin P. Gallen, Morgan, Lewis & Bockius,
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Richard J. Laufer.
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: December 23, 2003.
Brief description of amendments: The licensee proposed to revise
various parts of the Technical Specifications (TSs) to allow entry into
a mode or other specified condition in the applicability of a
specification while in a condition statement and the associated
required actions of the TSs, provided the licensee
[[Page 2739]]
performs a risk assessment and manages risk consistent with the program
in place for complying with the requirements of 10 CFR 50.65(a)(4).
Specifically, TS 3.0, ``Limiting Conditions for Operation (General),''
as well as other portions of the TSs (i.e., Sections 3.4, 3.7, and 3.8)
referencing TS 3.0, will be revised.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a
notice of opportunity for comment in the Federal Register on August 2,
2002 (67 FR 50475), on possible amendments concerning TSTF-359,
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 April 4, 2003 (68 FR 16579). In
its application for amendment, the licensee affirmed the applicability
of the following NSHC determination.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee presented
an analysis of NSHC by endorsing the model NSHC published in 68 FR
16579 (reproduced 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 allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed by proposed LCO 3.0.4, are no different than the
consequences of an accident while entering and relying on the
required actions while starting in a condition of applicability of
the TS. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
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).
Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. Thus, this change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in [a] Margin of Safety
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. The TS
allow operation of the plant without the full complement of
equipment through the conditions for not meeting the TS LCO. The
risk associated with this allowance is managed by the imposition of
required actions that must be performed within the prescribed
completion times. The net effect of being in a TS condition on the
margin of safety is not considered significant. The proposed change
does not alter the required actions or completion times of the TS.
The proposed change allows TS conditions to be entered, and the
associated required actions and completion times to be used in new
circumstances. This use is predicated upon the licensee's
performance of a risk assessment and the management of plant risk.
The change also eliminates current allowances for utilizing required
actions and completion times in similar circumstances, without
assessing and managing risk. The net change to the margin of safety
is insignificant. Therefore, 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: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Richard J. Laufer.
Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-317, Calvert
Cliffs Nuclear Plower Plant, Unit No. 2, Calvert County, Maryland
Date of amendment request: September 30, 2003.
Description of amendment request: The proposed amendment would
increase the maximum enrichment limit of the fuel assemblies that can
be stored in the Unit 2 spent fuel pool by taking credit for soluble
boron, burnup and configuration control in maintaining acceptable
margins of subcriticality.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed change will increase the maximum enrichment limit
of the fuel assemblies that can be stored in the Unit 2 spent fuel
pool (SFP) by taking credit for soluble boron, burnup and
configuration control in maintaining acceptable margins of
subcriticality. The proposed change will modify Technical
Specification 4.3.1 ``Criticality,'' add Technical Specification
3.7.16, ``Spent Fuel Pool Boron Concentration'' and add Technical
Specification 3.7.17 ``Spent Fuel Pool Storage.'' The postulated
accidents for the SFP are basically four types; (1) dropped fuel
assembly on top of the storage rack, (2) a misloading accident, (3)
an abnormal location of a fuel assembly, and (4) loss-of-normal
cooling to the SFP.
There is no increase in the probability of a fuel assembly drop
accident in the SFP when considering the higher enriched fuel or the
presence of soluble boron in the SFP water. Dropping a fuel assembly
on top of the SFP storage racks is not credible at Calvert Cliffs
due to the design of the spent fuel handling machine and the height
of the SFP storage racks. The handling of fuel assemblies has always
been performed in borated water and will not change as a result of
crediting soluble boron in the SFP criticality analysis. The
proposed change does not change the general design or
characteristics of the fuel assemblies. Therefore, the proposed
change does not increase the probability of a fuel assembly drop
accident.
There is no increase in the probability of the accidental
misloading of irradiated fuel assemblies into the SFP storage racks
when considering the higher enriched fuel or the presence of soluble
boron in the SFP water for criticality control. Fuel assembly
placement will continue to be controlled pursuant to approved fuel
handling procedures.
Due to the design of the SFP storage racks, an abnormal
placement of a fuel assembly into the SFP storage racks is not
possible. Also, the design of the SFP prevents an inadvertent
placement of a fuel assembly between the outer most storage cell and
the pool wall. The proposed change does not make any change to the
design of SFP. Therefore, there is no increase in the probability of
abnormal placement of a fuel assembly into the SFP storage racks.
The proposed change will not result in any changes to the SFP
cooling system, and the fuel assembly design and characteristics are
not changed by an increase in fuel enrichment. Therefore, there is
no increase in the probability of a loss of SFP cooling. Also, since
a high concentration of soluble boron has always been maintained in
the SFP water, there is no increase in the probability of the loss
of normal cooling to the SFP water considering the presence of
soluble boron in the pool water for criticality control.
[[Page 2740]]
There is no increase in the consequences of an accidental drop,
accidental misloading, or abnormal placement of a maximum enriched
fuel assembly into the SFP storage racks, because the criticality
analysis demonstrates that the pool will remain subcritical
following either event. The Technical Specification limit for SFP
boron concentration will ensure that an adequate SFP boron
concentration will be maintained.
There is no increase in the consequences of a loss-of-normal SFP
cooling because the Technical Specification boron concentration
provides significant negative reactivity. Loss of the SFP water via
boiling will not result in a loss of soluble boron, since the
soluble boron is not volatile. Therefore, loss of SFP cooling
system, without makeup flow, is not a mechanism for boron dilution.
Even in the unlikely event that soluble boron in the SFP is
completely diluted via unborated makeup flow, a pool completely
filled with maximum enriched unburned assemblies will remain
subcritical by a design margin that meets the requirements of 10 CFR
50.68.
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 will increase the maximum enrichment limit
of the fuel assemblies that can be stored in the Unit 2 SFP by
taking credit for soluble boron, burnup and configuration control in
maintaining acceptable margins of subcriticality. Increasing the
maximum enrichment limit does not create a new type of criticality
accident.
Soluble boron has been maintained in the SFP water and is
currently required by procedures. Therefore, crediting soluble boron
in the SFP criticality analysis will have no effect on normal pool
operation and maintenance. Crediting soluble boron will only result
in increased sampling to verify the boron concentration in
accordance with the proposed Technical Specification Surveillance
Requirement. This increased sampling will not create the possibility
of a new or different kind of accident.
A dilution of the SFP soluble boron has always been a
possibility. However, the boron dilution event previously had no
consequences, since boron was not previously credited in the
accident analysis. The initiating events that were considered for
having the potential to cause dilution of the boron in the SFP to a
level below that credited in the criticality analyses fall into
three categories: dilution by flooding, dilution by loss-of-coolant
induced makeup, and dilution by loss-of-cooling system induced
makeup. The SFP dilution analysis demonstrates that a dilution event
that could increase k-effective in the SFP to greater than 0.95 is
not a credible event. It is not credible that dilution could occur
for the required length of time without operator notice, since this
event would activate the high level alarm and initiate Auxiliary
Building flooding. In addition, in excess of 1,043,000 gallons of
unborated water must be added to the SFP to reach the minimum
soluble boron concentration. This is more water volume than is
contained in both pretreated water storage tanks and also more water
volume than is contained in the demineralized water storage tank and
both condensate storage tanks combined. Even in the unlikely event
that soluble boron in the SFP is completely diluted, the SFP will
remain subcritical by a design margin that meets the requirements of
10 CFR 50.68.
Burned assemblies have been stored in the SFP for many cycles.
Therefore, crediting burnup in the SFP criticality analysis will
have no effect on normal pool operation and maintenance. Fuel
assembly placement, although more complex, will continue to be
controlled pursuant to approved fuel handling procedures and in
accordance with Technical Specification spent fuel rack storage
configuration limitations.
The proposed change will not result in any other change in the
plant configuration or equipment design. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The Technical Specification changes proposed by this license
amendment request will provide an adequate safety margin to ensure
that the stored fuel assembly array of maximum enriched fuel will
always remain subcritical. Those limits are based on a plant
specific criticality analysis performed for the Calvert Cliffs Unit
2 SFP, that include technically supported margins.
Soluble boron is used to provide subcritical margin such that
the SFP k-effective is maintained less than or equal to 0.95. Since
k-effective is less than or equal to 0.95, the current margin of
safety is maintained. In addition, while the criticality analysis
utilized credit for soluble boron, the fuel in the SFP rack will
remain subcritical with no soluble boron with a 95 percent
probability at a 95 percent confidence level as required by 10 CFR
50.68. This substantial reduction in the SFP soluble boron
concentration was evaluated and shown not to be credible.
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
amendments request involves no significant hazards consideration.
Attorney for licensee: James M. Petro, Jr., Esquire, Counsel,
Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor,
Baltimore, MD 21202.
NRC Section Chief: Richard J. Laufer.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: October 16, 2003.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3.4.9 to change the minium
pressurizer (PZR) heater capacity from 126 to 400 kW to correct a non-
conservative TS associated with a PZR design basis deficiency.
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. Involve a significant increase in the probability or
consequences of an accident previously evaluated:
No. The proposed changes revise the minimum PZR [pressurizer]
heater capacity required and capable of being powered from an
emergency power supply source. UFSAR [Updated Final Safety Analysis
Report] do not take credit for PZR heater operation; however, an
implicit initial condition assumption of the safety analyses is that
RCS [Reactor Coolant System] is operating at normal pressure.
Assurance of this assumption is enhanced due to these proposed
changes. Consequently, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any kind of accident previously evaluated:
No. These changes correct a non-conservative value from the TS
[technical specification] and are necessary to assure RCS pressure
control and adequate natural circulation cooling. The available
heater capacity being powered from an emergency power supply is
approximately 1000 kW for the most restrictive unit which exceeds
the proposed 400 kW minimum capacity required by TS. The proposed
changes help ensure that the RCS is operating at normal pressure
which is an implicit initial assumption used in several UFSAR
described safety analyses. Consequently, these changes do not create
the possibility of a new or different kind of accident from any kind
of accident previously evaluated.
3. Involve a significant reduction in a margin of safety:
No. The proposed change does not adversely affect any plant
safety limits, set points, or design parameters. The change also
does not adversely affect the fuel, fuel cladding, RCS, or
containment integrity. Therefore, the proposed changes do not
involve a 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
[[Page 2741]]
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC 20005.
NRC Section Chief: John A. Nakoski.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: December 5, 2003.
Description of amendment request: The proposed amendment would
revise the Safety Limit Minimum Critical Power Ratio (SLMCPR) values in
Technical Specification 1.1.A.1 to incorporate the results of the
cycle-specific core reload analysis for Vermont Yankee Nuclear Power
Station Cycle 24 operation.
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 operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment, will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The basis of the Safety Limit Minimum Critical Power Ratio
(SLMCPR) is to ensure no mechanis0tic fuel damage is calculated to
occur if the limit is not violated. The new SLMCPR values preserve
the existing margin to transition boiling and probability of fuel
damage is not increased. The derivation of the revised SLMCPR for
Vermont Yankee for incorporation into the Technical Specifications,
and its use to determine plant and cycle-specific thermal limits,
have been performed using NRC [U.S. Nuclear Regulatory Commission]
approved methods. These plant-specific calculations are performed
each operating cycle and if necessary, will require future changes
to these values based upon revised core designs. The revised SLMCPR
values do not change the method of operating the plant and have no
effect on the probability of an accident initiating event or
transient.
Based on the above, Vermont Yankee has concluded that the
proposed change will not result in a significant increase in the
probability or consequences of an accident previously evaluated.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment, will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed changes result only from a specific analysis for
the Vermont Yankee core reload design. These changes do not involve
any new or different methods for operating the facility. No new
initiating events or transients result from these changes.
Based on the above, Vermont Yankee has concluded that the
proposed change will not create the possibility of a new or
different kind of accident from those previously evaluated.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment, will not involve a
significant reduction in a margin of safety.
The new SLMCPR is calculated using NRC approved methods with
plant and cycle specific parameters for the current core design. The
SLMCPR value remains high enough to ensure that greater than 99.9%
of all fuel rods in the core will avoid transition boiling if the
limit is not violated, thereby preserving the fuel cladding
integrity. The operating MCPR limit is set appropriately above the
safety limit value to ensure adequate margin when the cycle specific
transients are evaluated. Accordingly, the margin of safety is
maintained with the revised values.
As a result, Vermont Yankee has determined that the proposed
change will not result in a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: Darrell J. Roberts, Acting.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: December 19, 2003.
Description of Amendment Request: The proposed amendment deletes
requirements from the Technical Specifications (TS) to maintain
hydrogen recombiners and hydrogen monitors.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 2, 2002 (67 FR 50374), on possible
amendments to eliminate the hydrogen recombiners from TS, including a
model safety evaluation and model no significant hazards consideration
(NSHC) determination, using the Consolidated Line Item Improvement
Process (CLIIP). The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on September 25, 2003 (68 FR
55416). The licensee affirmed the applicability of the model NSHC
determination in its application dated December 19, 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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen monitors are no longer required to mitigate design-basis
accidents and, therefore, the hydrogen monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
RG [Regulatory Guide] 1.97 Category 1, is intended for key variables
that most directly indicate the accomplishment of a safety function
for design-basis accident events. The hydrogen monitors no longer
meet the definition of Category 1 in RG 1.97. As part of the
rulemaking to revise 10 CFR 50.44, the Commission found that
Category 3, as defined in RG 1.97, is an appropriate categorization
for the hydrogen monitors because the monitors are required to
diagnose the course of beyond design-basis accidents.
The regulatory requirements for the hydrogen monitors can be
relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3, and removal
of the hydrogen monitors from TS will not prevent an accident
management strategy through the use of the SAMGs, the emergency plan
(EP), the emergency operating procedures (EOP), and site survey
monitoring that support modification of emergency plan protective
action recommendations (PARs).
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of
[[Page 2742]]
the hydrogen monitor requirements, including removal of these
requirements from TS, does not involve a significant increase in the
probability or the consequences of any 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 elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, will not result in any failure mode
not previously analyzed. The hydrogen recombiner and hydrogen
monitor equipment was intended to mitigate a design-basis hydrogen
release. The hydrogen recombiner and hydrogen monitor equipment are
not considered accident precursors, nor does their existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within
the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, in light of existing plant equipment,
instrumentation, procedures, and programs that provide effective
mitigation of and recovery from reactor accidents, results in a
neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3--hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI, Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Therefore, this change does not involve a significant reduction
in the margin of safety. Removal of hydrogen monitoring from TS will
not result in a significant reduction in their functionality,
reliability, and availability.
Based upon the reasoning presented above, the requested change does
not involve a significant hazards consideration. Therefore, the NRC
staff proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Exelon Generation Company, LLC, Docket No. STN 50-454, Byron Station,
Unit 1, Ogle County, Illinois
Date of amendment request: December 5, 2003.
Description of amendment request: The proposed amendment would
allow irradiation of two lead test assemblies (LTAs) and two
``standard'' Westinghouse 17x17 VANTAGE+ZIRLOTM assemblies
beyond the current fuel rod-average licensing basis burnup value of
60,000 MWD/MTU up to 65,000 MWD/MTU during the current operating cycle
(B1C13). Irradiation of these four assemblies is intended to confirm
the acceptable use of the ZIRLOTM alloys to a discharge
burnup level exceeding the current licensing basis.
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.
Fuel rod defects or failures are not considered as initiators
for any previously analyzed accident; therefore the requested
license amendment will have no effect on the probability of any
previously evaluated accident. In addition, NRC-approved
methodologies and technical reports have been used in the B1C13
specific reload safety evaluation to confirm that the fuel rod
design limits will be met; therefore, increasing the burnup limit of
the specified fuel assemblies to the requested value will not
increase the consequences of any previously analyzed accident.
The regular ZIRLOTM and ZIRLOTM (LT-1)
high burnup fuel rods will continue to satisfy the specified
acceptable fuel design limits (SAFDLs) specified in NRC-approved
Westinghouse topical reports. The clad integrity of the
ZIRLOTM and ZIRLOTM (LT-1) high burnup rods
will be maintained as the subject fuel assemblies will be placed in
less than limiting core locations and will continue to meet the
safety parameter requirements. The acceptability of using the
ZIRLOTM and ZIRLOTM (LT-1) high burnup rods
has been evaluated and confirmed in the B1C13 Reload Safety
Evaluation supported by the Westinghouse LTA Report, ``Byron Unit 1
Cycle 13 LTA Report,'' dated August 2003.
It has been shown in WCAP-12610-P-A, that even though there are
variations in core inventories of isotopes due to extended burnup up
to 75,000 MWD/MTU, there are no significant increases of isotopes
that are major contributors to accident doses. It is worthy to note
that, at higher burnups, there is actually a reduction in certain
isotopes that are major dose contributors under accident situations
(e.g., Kr-88). With only a limited number of ZIRLOTM and
ZIRLOTM (LT-1) high burnup rods in the entire core, any
variation of isotopes will be extremely small. Thus, the radiation
dose limitations of 10 CFR 100, ``Reactor Site Criteria,'' will not
be exceeded.
Based on the above discussion, it is concluded that 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 to increase the current fuel rod-average
burnup limit does not involve the use or installation of new
equipment and all currently installed equipment will not be operated
in a new or different manner. No new or different system
interactions are created and no new processes are introduced. The
proposed change will not introduce any new failure mechanisms,
malfunctions, or accident initiators not already considered in the
design and licensing bases.
Based on this evaluation, the proposed change does 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 change to increase the current fuel rod-average
burnup limit of 60,000 MWD/MTU up to 65,000 MWD/MTU during B1C13
will cause the following fuel rod design criteria to become more
limiting: Fuel rod growth, clad fatigue, rod internal pressure and
cladding corrosion. However, the regular ZIRLOTM and
ZIRLOTM (LT-1) high burnup fuel rods will continue to
satisfy the SAFDLs specified in NRC-approved Westinghouse topical
reports as noted above. The clad integrity of the ZIRLOTM
and ZIRLOTM (LT-1) high burnup rods and the appropriate
margin to safety will be maintained as the subject fuel assemblies
will be placed in less than limiting core locations and will
continue to meet the safety parameter requirements. The
acceptability of using the ZIRLOTM and ZIRLOTM
(LT-1) high burnup rods has been evaluated and confirmed in the
B1C13 Reload Safety Evaluation supported by the Westinghouse LTA
Report, ``Byron Unit 1 Cycle 13 LTA Report,'' dated August 2003.
Based on the above evaluation, 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.
[[Page 2743]]
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
Exelon Generation Company, LLC, Docket No. 50-265, Quad Cities Nuclear
Power Station, Unit 2, Rock Island County, Illinois Date of amendment
request:
Date of amendment request: November 14, 2003, as supplemented by
letter dated December 23, 2003.
Description of amendment request: The proposed amendment would
revise the values and wording of the technical specifications safety
limit minimum critical power ratio (SLMCPR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The 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. Limits have been established consistent with NRC
approved methods to ensure that fuel performance during normal,
transient, and accident conditions is acceptable. The proposed
change conservatively establishes the SLMCPR for Quad Cities Nuclear
Power Station (QCNPS), Unit 2, Cycle 18 such that the fuel is
protected during normal operation and during any plant transients or
anticipated operational occurrences (AOOs).
Changing the SLMCPR does not increase the probability of an
evaluated accident. The change does not require any physical plant
modifications, physically affect any plant components, or entail
changes in plant operation. Therefore, no individual precursors of
an accident are affected.
The proposed change revises the SLMCPR to protect the fuel
during normal operation as well as during any transients or
anticipated operational occurrences. Operational limits will be
established based on the proposed SLMCPR to ensure that the SLMCPR
is not violated during all modes of operation. This will ensure that
the fuel design safety criterion (i.e., that at least 99.9% of the
fuel rods do not experience transition boiling during normal
operation and anticipated operational occurrences) is met. Since the
proposed change does not affect operability of plant systems
designed to mitigate any consequences of accidents, 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. Does the proposed change 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 creating one or more new precursors of that
accident. New accident precursors may be created by modifications of
plant configuration, including changes in allowable modes of
operation. The proposed change does not involve any plant
configuration modifications or changes to allowable modes of
operation. The proposed change to the SLMCPR assures that safety
criteria are maintained for QCNPS, Unit 2, Cycle 18.
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?
The SLMCPR provides a margin of safety by ensuring that at least
99.9% of the fuel rods do not experience transition boiling during
normal operation and AOOs if the MCPR limit is not violated. The
proposed change will ensure the appropriate level of fuel
protection. Additionally, operational limits will be established
based on the proposed SLMCPR to ensure that the SLMCPR is not
violated during all modes of operation. This will ensure that the
fuel design safety criteria (i.e., that no more than 0.1% of the
rods are expected to be in boiling transition if the MCPR limit is
not violated) are met.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The supplemental letter contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The 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, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
Florida Power and Light Company, Docket No. 50-389, St. Lucie Plant,
Unit No. 2, St. Lucie County, Florida
Date of amendment request: December 2, 2003.
Description of amendment request: The proposed amendment would
revise the Technical Specifications to allow a reduction in the minimum
reactor coolant system flow, corresponding to an increase in the steam
generator tube plugging limit from 15 percent to 30 percent.
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) Operation of the facility in accordance with the proposed
amendment would 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 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,
PO Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Allen G. Howe.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of amendment request: October 8, 2003.
Description of amendment request: The proposed amendment is to
revise Technical Specifications (TS) 4.2.b.3.a, ``Inspection
Frequency,'' for the Kewaunee Nuclear Power Plant (KNPP). The proposed
one-time change would revise the steam generator (SG) inspection
interval requirements in TS for KNPP to allow a 40-month inspection
interval after one SG inspection.
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. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
The proposed one-time change revises the Steam Generator (SG)
inspection interval requirements in Technical Specifications (TS)
4.2.b.3.a, following the Kewaunee Nuclear Plant, spring 2003
refueling outage, to allow a 40-month inspection frequency after one
inspection, rather than after two consecutive inspections results
that are within the C-1 category.
The proposed on-time extension of the SG tube in-service
inspection interval does not involve changing any structure, system,
or component, or affect reactor operations. It is not an initiator
of an accident and does not
[[Page 2744]]
change any existing safety analysis previously analyzed in the
Kewaunee Updated Safety Analysis Report (USAR). As such, the
proposed changes do not involve a significant increase in the
probability of an accident previously evaluated.
Since the proposed change does not alter the plant design, there
is no direct increase in SG leakage. Industry experience indicates
that the probability of increased SG tube degradation would be very
low. Additionally, steps described below will further minimize the
risk associated with this extension. For example, the scope of
inspections performed during the last KNPP refueling outage (i.e.,
the first refueling outage following Steam generator replacement
(SGR) exceeded the TS requirements for the first two refuleing
outages after SGR. That is, more tubes were inspected than were
required by TS (i.e., 100 percent inspection was performed).
Currently, KNPP does not have an active SG damage mechanism, and
will meet the current industry examination guidelines without
performing additional SG inspections until the spring 2006 refueling
outage. Additionally, as part of our SG Tube Surveillance Program,
both a Condition Monitoring Assessment and an Operational Assessment
are performed after each inspection and compared to the Nuclear
Energy Institute (NEI) 97-06, ``Steam Generator Program
Guidelines,'' performance criteria. The results of the Condition
Monitoring Assessment demonstrated that all performance criteria
were met during the KNPP spring 2003 refueling outage, and the
results of the Operational Assessment show that all performance
criteria will be met over the proposed operating period. Considering
these actions, along with improved SG design and reliability of
Westinghouse replacement SGs, extending the SG tube inspection
frequency does not involve a significant increase in the
consequences of an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated?
The proposed change revises the SG inspection frequency
requirements in TS 4.2.b.3.a, to allow a 40-month inspection
interval after one inspection, rather than after two consecutive
inspections with inspection results within the C-1 category.
The proposed change will not alter any plant design basis or
postulated accident resulting from potential SG tube degradation.
The scope of inspections (i.e., 100 percent) performed during the
last KNPP refueling outage (i.e., the first refueling outage
following SG replacement) significantly exceeded the TS requirements
for the scope of the first two refueling outages after SG
replacement.
Primary to secondary leakage that may be experienced during all
plant conditions is expected to remain within current accident
analysis assumptions. The proposed change does not affect the design
of the SGs, the method of SG operation, or reactor coolant chemistry
controls. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. The
proposed change involves a one-time extension to the SG tube in-
service inspection frequency, and therefore will not give rise to
new failure modes. In addition, the proposed change does not impact
any other plant system or components.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety?
The SG tubes are an integral part of the Reactor coolant System
(RCS) pressure boundary that are relied upon to maintain the RCS
pressure and inventory. The SG tubes isolate the radioactive fission
products in the reactor coolant from the secondary system. The
safety function of the SG is maintained by ensuring integrity of the
SG tubes. In addition, the SG tubes comprise the heat transfer
surface between the primary and secondary systems such that residual
heat can be removed from the primary system.
SG tube integrity is a function of the design, environment, and
current physical condition. Extending the SG tube inservice
inspection frequency by one operating cycle will not alter the
function or design of the SG. SG inspections conducted during the
first refueling outage following SG replacement demonstrated that
the SGs do not have an active damage mechanism, and the scope of
those inspections significantly exceeded those required by the TS.
These inspection results were comparable to similar inspection
results for similar replacement SGs installed at other plants, and
subsequent inspections at those plants yielded results that support
this extension request. The improved design of the replacement SGs
also provides reasonable assurance that significant tube degradation
is not likely to occur over the proposed operating period.
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: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket No. 50-282, Prairie Island
Nuclear Generating Plant, Unit 1, Goodhue County, Minnesota
Date of amendment request: August 27, 2003, as supplemented
December 16, 2003.
Description of amendment request: The proposed amendment would
revise Technical Specification 5.5.14, ``Containment Leakage Rate
Testing Program,'' to allow Unit 1 to be excepted from the requirements
of Regulatory Guide 1.163, for post-modification integrated leakage
rate testing associated with steam generator replacement.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change would provide the Prairie Island Nuclear
Generating Plant an exception from performing a required containment
integrated leak rate test following the replacement of the steam
generators in Unit 1.
Integrated leak rate tests are performed to assure the leak-
tightness of the primary containment boundary system, and as such
they are not accident initiators. Therefore, not performing an
integrated leak rate test will not affect the probability of an
accident previously evaluated.
The intent of post-modification integrated leak rate testing
requirements is to assure the leak-tight integrity of the area
affected by the modification. For the Unit 1 steam generator
replacement modification, this intent will be satisfied by
performing the American Society of Mechanical Engineers code
required inspections and tests. Since the leak-tightness integrity
of the primary containment boundary affected by replacement of the
steam generators will be assured, there is no change in the primary
containment boundary's ability to confine radioactive materials
during an accident.
Therefore adding a Technical Specification requirement that
provides an exception for Unit 1 from the steam generator
replacement post-modification integrated leak rate testing
requirements does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change would provide the Prairie Island Nuclear
Generating Plant an exception from performing a required containment
integrated leak rate test following the replacement of the steam
generators in Unit 1.
Providing an exception from performing a test does not involve a
physical change to the plant nor does it change the operation of the
plant. Thus it cannot introduce a new failure mode.
Therefore adding a Technical Specification requirement that
provides an exception for Unit 1 from the steam generator
replacement post-modification integrated leak rate testing
requirements does not create the possibility of a new or different
kind of accident from any previously evaluated.
[[Page 2745]]
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change would provide the Prairie Island Nuclear
Generating Plant an exception from performing a required containment
integrated leak rate test following the replacement of the steam
generators in Unit 1.
The intent of post-modification integrated leak rate testing
requirements is to assure the leak-tight integrity of the area
affected by the modification. This intent will be satisfied by
performing American Society of Mechanical Engineers code required
inspections and tests. The acceptance criterion for American Society
of Mechanical Engineers code system pressure testing for the base
metal and welds is no leakage. In addition, the test pressure for
the system pressure test will be several times that required during
an integrated leak rate test. Since the leak-tight integrity of the
primary containment boundary affected by replacement of the steam
generators will be assured, there is no change in the primary
containment boundary's ability to confine radioactive materials
during an accident.
Therefore, adding a Technical Specification requirement that
provides an exception for Unit 1 from the steam generator
replacement post-modification integrated leak rate testing
requirements does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: L. Raghavan.
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 22, 2003.
Description of amendment request: The proposed amendment would
revise the Unit 1 and 2 Technical Specifications (TSs) by adding TS
3.3.1.3, ``Oscillation Power Range Monitor (OPRM) Instrumentation,''
and revising TS 3.4.1, ``Recirculation Loops Operating,'' and TS 5.6.5,
``Core Operating Limits Report,'' to remove specifications and
information related to current stability specifications which will no
longer be needed.
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 OPRM most directly affects the APRM [average power range
monitor] and LPRM [local power range monitor] portions of the Power
Range Neutron Monitoring system. Its installation does not affect
the operation of these sub-systems. None of the accidents or
equipment malfunctions affected by these sub-systems are affected by
the presence or operation of the OPRM. The APRM channels provide the
primary indication of neutron flux within the core and respond
almost instantaneously to neutron flux changes. The APRM Fixed
Neutron Flux-High function is capable of generating a trip signal to
prevent fuel damage or excessive reactor pressure. For the ASME
[American Society of Mechanical Engineers] overpressurization
protection analysis in FSAR [Final Safety Analysis Report] Chapter
5, the APRM Fixed Neutron Flux-High function is assumed to terminate
the main steam isolation valve closure event. The high flux trip,
along with the safety/relief valves, limits the peak reactor
pressure to less than the ASME Code limits. The control rod drop
accident (CRDA) analysis in Chapter 15 takes credit for the APRM
Fixed Neutron Flux-High function to terminate the CRDA. The
Recirculation Flow Controller Failure event (pump runup) is also
terminated by the high neutron flux trip. The APRM Fixed Neutron
Flux-High function is required to be OPERABLE in MODE 1 where the
potential consequences of the analyzed transients could result in
the Safety Limits (e.g., MCPR [minimum critical power ratio] and
Reactor pressure) being exceeded.
The installation of the OPRM equipment does not increase the
consequences of a malfunction of equipment important to safety. The
APRM and RPS [Reactor Protection System] systems are designed to
fail in a tripped (fail safe) condition; the OPRM will have no
affect on the consequences of the failure of either system. An
inoperative trip signal is received by the RPS any time an APRM mode
switch is moved to any position other than Operate, an APRM module
is unplugged, the electronic operating voltage is low, or the APRM
has too few LPRM inputs. These functions are not specifically
credited in the accident analysis, but are retained for the RPS as
required by the NRC approved licensing basis.
The OPRM allows operation under operating conditions presently
restricted by the current Technical Specifications by providing
automatic suppression functions in the area of concern in the event
an instability occurs. The consequences of any accident or equipment
malfunction are not increased by operating under those conditions.
Although protected by the OPRM from thermal-hydraulic core
instabilities above 30% core power, operation under natural core
circulation conditions is not allowed. No accidents or transients of
a type not analyzed in the FSAR are created by operating under these
conditions with the protection of the OPRM system.
This change does not increase the probability of an accident as
previously evaluated. The OPRM is designed and installed to not
degrade the existing APRM, LRPM, and RPS systems. These systems will
still perform all of their intended functions. The new equipment is
tested and installed to the same or more restrictive environmental
and seismic envelopes as the existing systems. The new equipment has
been designed and tested to electromagnetic interference (EMI)
requirements which assure correct operation of the existing
equipment. The new system has been designed to single failure
criteria and is electrically isolated from equipment of different
electrical divisions and from non-1E equipment. The electrical
loading is within the capability of the existing power sources and
the heat loads are within the capability of existing cooling
systems. The OPRM allows operation under operating conditions
presently forbidden or restricted by the current Technical
Specifications. No other transient or accident analysis assumes
these operating restrictions.
Therefore, this 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.
This proposal does not create the possibility of a new or
different type of accident from any accident previously evaluated.
The OPRM system is a monitoring and accident mitigation system that
cannot create the possibility for an accident not previously
evaluated.
The OPRM will allow operation in conditions restricted by the
current Technical Specifications. Although protected by the OPRM
from thermal-hydraulic core instabilities above 30% core power,
operation under natural circulation conditions is not allowed. No
accidents or transients of a type not analyzed in the FSAR are
created by operating under these conditions with the protection of
the OPRM system. No new failure modes of either the new OPRM
equipment or of the existing APRM equipment have been introduced.
Quality software design, testing, implementation and module self-
health testing provides assurance that no new equipment malfunctions
due to software errors are created. The possibility of an accident
of a new or different type than any evaluated previouly is not
created.
The new OPRM equipment is designed and installed to the same
system requirements as the existing APRM equipment and is designed
and tested to have no impact on the existing functions of the APRM
system. Appropriate isolation is provided where new interconnections
between redundant separation groups are formed. The OPRM modules
have been designed and tested to assure that no new failure modes
have been introduced.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
[[Page 2746]]
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
There has been no reduction in the margin of safety as defined
in the basis for the Technical Specifications. The OPRM system does
not negatively impact the existing APRM system. As a result, the
margins in the Technical Specifications for the APRM system are not
impacted by this addition.
Current operation under the ICAs [interim corrective actions]
provides an acceptable margin of safety in the event of an
instability event as the result of preventive actions and Technical
Specification controlled response by the control room operators. The
OPRM system provides an increase in the reliability of the
protection of the margin of safety by providing automatic protection
of the MCPR safety limit, while the protection burden is
significantly reduced for the control room operators. This
protection is demonstrated as described above, and in the NRC
reviewed and approved Topical Reports NEDO-32465-A and CENPD-400-P-
A.
Replacement of the ICA operating restrictions from Technical
Specifications with the OPRM system does not affect the margin of
safety associated with any other system or fuel design parameter.
Therefore, this change does not involve a 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: 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.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: November 17, 2003.
Description of amendment request: The proposed change would revise
the Technical Specifications to delete the primary containment
isolation valves and instrumentation associated with the permanent
removal of the reactor vessel head spray piping.
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 not involve a significant increase in the probability or
consequences of an accident previously analyzed?
Response: No.
The proposed changes to Technical Specification Tables 3.3.2-1,
3.3.7.4-2, 3.4.3.2-1, and 3.6.3-1 do not involve a change in
structures, systems, or components that would affect the probability
or consequences of any accident previously evaluated in the Hope
Creek Updated Final Safety Analysis Report.
The proposed changes involve eliminating piping and valves
associated with the reactor head spray. The reactor head spray
system was initially provided to cool down the steam dryer and
separator during shutdown. The head spray system is not credited for
the prevention or mitigation of any accident. Therefore, neither the
offsite or control room radiological consequences are affected. The
head spray piping removal and addition of a bolted flange on the
reactor coolant pressure boundary enhances plant safety by
eliminating a source of pipe whip and potential leakage. In
addition, the drywell penetration will be capped and welded closed.
This will maintain primary containment integrity and will be
periodically tested in conjunction with the containment integrated
leak rate test.
Therefore, as discussed above, this modification does not
involve a significant increase in the probability or consequences
from any accident previously analyzed.
2. Does not create the possibility of a new or different kind of
accident from any accident previously analyzed?
Response: No.
The proposed changes to Technical Specification Tables 3.3.2-1,
3.3.7.4-2, 3.4.3.2-1, and 3.6.3-1 do not involve a change in
structures, systems, or components that would create a new or
different kind of accident from any accident previously evaluated in
the Hope Creek Updated Final Safety Analysis Report.
The proposed change to eliminate the head spray piping and the
addition of a bolted flange on the reactor coolant pressure boundary
enhances plant safety by eliminating a source of pipe whip and
potential leakage. In addition, the drywell penetration will be
capped and welded closed. This will maintain primary containment
integrity and will be tested in conjunction with the containment
integrated leak rate test.
Therefore, as discussed above, this modification does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does not involve a significant reduction in the margin of
safety?
Response: No.
The proposed change to delete the head spray valves from Tables
3.3.2-1, 3.3.7.4-2, 3.4.3.2-1, and 3.6.3-1 does not reduce any
margin of safety as defined in the Technical Specifications or
Bases. The bolted flange that will be installed on the head spray
penetration will maintain the integrity of the reactor coolant
pressure boundary. This flange would then be tested as part of the
reactor pressure vessel hydrostatic test. In addition, the drywell
penetration will be capped and welded closed. This will maintain
primary containment integrity and will be tested as part of the
containment integrated leak rate test.
Accordingly, based on the above, the proposed change does not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, PO Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: Darrell Roberts, Acting.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of amendment request: October 13, 2003.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) limiting conditions for
operation 3.8.4, 3.8.5, and 3.8.6, on direct current sources, operating
and shutdown, and battery cell parameters. The proposed amendments
creates TS 5.5.19, for a battery monitoring and maintenance program.
The bases are revised to be consistent with these changes. The proposed
amendments are based on Technical Specification Task Force (TSTF)
Traveler, TSTF-360, Revision 1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed changes increase the Completion Time for an
inoperable battery, relocate preventative maintenance requirements
to licensee controlled programs, and generally restructure the TS
[technical specification] requirements for DC [direct current]
sources. The revised requirements will allow licensed operators to
focus their attention on battery parameters that are indicative of
battery operability as opposed to preventative maintenance issues.
The increased Completion Time for an inoperable battery will allow
corrective maintenance to be accomplished via a more orderly and
effective work process. It will also minimize the potential for an
additional shutdown/restart transient to comply with the TS in order
to accomplish the required maintenance. The DC sources are not
initiators to any analyzed accident sequence. Operation in
accordance with the proposed TS will continue to ensure that the DC
sources remain capable of performing their safety function and that
all analyzed accidents will continue to be mitigated as previously
analyzed.
[[Page 2747]]
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
No. The proposed changes do not introduce any new equipment,
create new failure modes for existing equipment, or create any new
limiting single failures. Plant operation will not be altered, and
all safety functions previously addressed in accident analyses will
continue to be performed.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
No. The proposed changes will not adversely affect operation of
plant equipment--principally the four Class 1E DC sources and the
equipment supported by them. The changes aimed at restructuring the
TS requirements for DC sources will have the effect of reducing the
burden on licensed operators by focusing the TS requirements on
conditions that impair DC source operability. Requirements related
to preventive maintenance will be addressed via new Specification
5.5.19 and the plant maintenance program. Margin to the battery
operability requirements will continue to be maintained at current
levels in accordance with IEEE-450. The extended Completion Time for
an inoperable battery has been shown to have a negligible impact on
plant risk using the criteria of Regulatory Guides 1.174 and 1.177.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
Nations Bank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Section Chief: John A. Nakoski.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of amendment request: December 15, 2003.
Description of amendment request: The proposed amendments would
revise Technical Specifications surveillance requirement (SR) 3.3.1.2
for the nuclear instrumentation system power range daily surveillance
when operating above 15-percent rated thermal power. In addition, the
format of SR 3.3.1.3 is being revised to be consistent with the format
of the proposed change to SR 3.3.1.2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change to SR [surveillance requirement] 3.3.1.2
does not significantly increase the probability or consequences of
an accident previously evaluated in the FSAR [Final Safety Analysis
Report]. This modification does not directly initiate an accident.
The consequences of accidents previously evaluated in the FSAR are
not adversely affected by this proposed change because the change to
the NIS [nuclear instrumentation system] Power Range channel
adjustment requirement ensures the conservative response of the
channel even at part power levels. The proposed change to SR 3.3.1.3
is to change the format consistent with the format of the proposed
change to SR 3.3.1.2.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change to SR 3.3.1.2 does not create the
possibility of a new or different kind of accident than any accident
already evaluated in the FSAR. No new accident scenarios, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed change. The proposed Technical Specifications change
does not challenge the performance or integrity of any safety-
related systems. The proposed change to SR 3.3.1.3 is to change the
format to be consistent with the format of the proposed change to SR
3.3.1.2.
3. Does the proposed change involve a significant reduction in
the margin of safety?
The proposed change to SR 3.3.1.2 does not involve a significant
reduction in a margin of safety. The proposed change does require a
revision to the criterion for implementation of Power Range channel
adjustment based on secondary power calorimetric calculation;
however, the change does not eliminate any RTS [reactor trip system]
surveillances or alter the frequency of surveillances required by
the Technical Specifications. The revision to the criterion for
implementation of the daily surveillance will have a conservative
effect on the performance of the NIS Power Range channel,
particularly at part power after normalization at 100% RTP [rated
thermal power] conditions. The nominal trip setpoints specified by
the Technical Specifications and the safety analysis limits assumed
in the transient and accident analysis are unchanged. The margin of
safety associated with the acceptance criteria for any accident is
unchanged. The proposed change to SR 3.3.1.3 is to change the format
to be consistent with the format of the proposed change to SR
3.3.1.2.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Section Chief: John A. Nakoski.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: December 1, 2003.
Description of amendment request: The proposed amendments would
revise Technical Specifications Section 5.5.12, ``Primary Leakage Rate
Testing Program,'' to change the peak calculated post accident primary
containment internal pressure to support a 10 psi increase in the
nominal Unit 1 and 2 reactor operating pressure.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change to TS [technical specification] section
5.5.12, ``Primary Containment Leakage Rate Testing Program'',
involves an increase to the peak post accident primary containment
pressure. It does not involve physical changes to the primary
containment structure itself, nor to any of its support systems and
components, nor does it involve changes to any other systems and
components designed for the prevention of previously analyzed
events. Consequently, the proposed amendment does not involve a
significant increase in the probability of occurrence of a
previously evaluated event.
The increase in operating pressure for the Hatch reactors from
1035 psig to 1045 psig results in an increase to the peak post-
accident primary containment internal pressure. This pressure
increases from 50.5 to 50.8 psig for Unit 1 and from 46.9 to 47.3
psig for Unit 2. This is a very small increase with respect to the
Unit 1 and 2 primary containment design pressure of 56 psig and with
the maximum code allowable pressure of 62 psig. The primary
containment thus remains capable of withstanding the post accident
pressure and thus the consequences of a previously evaluated event
are not increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The primary containment boundary will not be altered by the
proposed change to
[[Page 2748]]
Technical Specifications sections 5.5.12, Primary Containment
Leakage Rate Testing Program. Furthermore, the primary containment
will function as presently described in the Updated Final Safety
Analysis Report and will be subject to the same structural and
functional requirements. The containment will be operated,
maintained and surveilled as before, with the exception of the
increased peak post accident pressure, which changes the post
accident test pressure acceptance criteria. As a result, no new
modes of operation are introduced by this Technical Specifications
change and therefore, the possibility of a new or different kind of
accident from any previously evaluated is not created.
3. Does the proposed change involve a significant decrease in
the margin of safety?
The change in the analyzed peak post accident containment
pressure will require that the containment be tested to ensure that
it meets leakage acceptance criteria at the new pressures of 50.8
psig and 47.3 psig for Units 1 and 2 respectively. Therefore, the
primary containment's ability to sustain the slightly higher
pressures will be verified during leak rate testing at the required
intervals.
The Unit 1 peak pressure increases from 50.5 to 50.8 psig and
the Unit 2 pressure increases from 46.9 to 47.3 psig. The primary
containment design pressure is 56 psig for both units and the
maximum code allowable pressure is 62 psig. Therefore, the margin to
the design and maximum code allowable pressures has not been
significantly affected. As a result, this proposed Technical
Specifications change does not significantly reduce the margin of
safety associated with the primary containment function.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: John A. Nakoski.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: May 12, 2003, as revised by letter dated
December 5, 2003.
Brief description of amendment request: By letter dated December 5,
2003, Entergy submitted a revised application for amendment to
Technical Specification (TS) 3.3.6.1, ``Primary Containment and Drywell
Isolation Instrumentation,'' to add a provision to the APPLICABILITY
function that will eliminate the requirement that the Residual Heat
Removal System Isolation, Reactor Vessel Water Level-Low, Level 3, be
OPERABLE under certain conditions during refueling outages.
Specifically, the proposed change requested in the original application
dated May 12, 2003, would remove the requirement for this isolation
function, specified in Table 3.3.6.1-1, when the upper containment
reactor cavity is at the High Water Level condition specified in TS
3.5.2, ``Emergency Core Cooling Systems Shutdown.'' The revised
application adds a new surveillance requirement (SR) 3.3.6.1.9 to
verify that the water level in the upper containment pool is greater
than or equal to 22 feet 8 inches above the reactor pressure vessel
flange every four hours, and adds a footnote to Table 3.3.6.1-1, Item
5.b, for MODE 5 that states that the function is not required when the
upper containment reactor cavity and transfer canal gates are removed
and SR 3.3.6.1.9 is met. The proposed SR and footnote are only
applicable in MODE 5. The May 12, 2003, application was previously
noticed in the Federal Register on June 10, 2003 (68 FR 34665).
Date of publication of individual notice in Federal Register:
December 15, 2003 (68 FR 69726).
Expiration date of individual notice: January 14, 2004.
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 email to [email protected].
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: July 14, 2003, as supplemented
by letter dated October 1, 2003.
[[Page 2749]]
Brief description of amendments: The amendments extend from 1 hour
to 24 hours the completion time for Condition B of Technical
Specification 3.5.1, which defines requirements for the restoration of
an emergency core cooling system accumulator when it has been declared
inoperable for a reason other than boron concentration.
Date of issuance: December 23, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance December 23, 2003.
Amendment Nos.: 211, 205, 218, and 200.
Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9, and
NPF-17: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 14, 2003 (68 FR
59214).
The supplement dated October 1, 2003, provided clarifying
information that did not change the scope of the July 14, 2003,
application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 23, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of application for amendment: January 9, 2003.
Brief description of amendment: The proposed Technical
Specification (TS) amendment request changes the definition of a Logic
System Functional Test, deletes the definition of a Simulated Automatic
Actuation, clarifies Surveillance Requirement 4.5.G.1.a regarding
simulated automatic actuation testing, and revises associated TS Bases.
Date of Issuance: December 23, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 216.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5674).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated December 23, 2003.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: October 24, 2003.
Brief description of amendment: The amendment revises TS 3.1.8,
``Scram Discharge Volume (SDV) Vent and Drain Valves,'' for the
condition of having one or more SDV vent or drain lines with one valve
inoperable.
Date of issuance: December 30, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No: 161.
Facility Operating License No. NPF-29: The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 25, 2003 (68
FR 66135).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 30, 2003.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: May 8, 2003, as supplemented by
letter dated October 24, 2003.
Brief description of amendment: The amendment changes Technical
Specification (TS) 3.3.6.1, ``Primary Containment and Drywell Isolation
Instrumentation,'' to add a note allowing intermittent opening of
penetration flow paths, under administrative control, that are isolated
to comply with TS ACTIONS and to revise the operability requirement for
the Reactor Core Isolation Cooling (RCIC) steam supply line low
pressure isolation instrumentation to be consistent with the RCIC
system operability requirements.
Date of issuance: January 8, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No: 162.
Facility Operating License No. NPF-29: The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: June 10, 2003 (68 FR
34664).
The October 24, 2003, supplemental letter provided clarifying
information that did not change the scope of the original Federal
Register notice or the original no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 8, 2004.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: September 3, 2003.
Brief description of amendment: The amendment modified Technical
Specification (TS) requirements for mode change limitations to adopt
the TS Task Force (TSTF) change TSTF-359, ``Increase Flexibility in
Mode Restraints.''
Date of issuance: January 5, 2004.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 109.
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 30, 2003 (68
FR 56345).
The staff's related evaluation of the amendment is contained in a
Safety Evaluation dated January 5, 2004.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station,
Unit No. 2, Salem County, New Jersey
Date of application for amendment: July 1, 2003, as supported by
letter dated June 16, 2003, and supplemented on November 11, 2003.
Brief description of amendment: The requested changes revise
License Condition 2.C.(10) to document changes to the Salem Post-Fire
Safe Shutdown (SSD) strategy for Fire Areas 2-FA-AB-64B, 2-FA-AB-84B,
and 2-FA-AB-84C. The licensee requested changes to the SSD as a result
of recent plant modifications implemented in response to the resolution
of Electrical Raceway Fire Barrier System issues at Salem.
Date of issuance: January 7, 2004.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 242.
Facility Operating License No. DPR-75: This amendment revised the
Facility Operating License.
Date of initial notice in Federal Register: July 16, 2003 (68 FR
42134). The supporting and supplemental
[[Page 2750]]
letters dated June 16, and November 11, 2003, contained clarifying
information that did not change the NRC staff's proposed finding of no
significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 7, 2004.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: May 22, 2003.
Brief description of amendments: The amendments revise Technical
Specification (TS) 3/4.3.2, ``Engineering Safety Features Actuation
System Instrumentation,'' and TS 3/4.9.9, ``Refueling Operations--
Containment Ventilation Isolation System,'' governing radiation
monitoring instrumentation, to relax restrictions on containment purge
valve operation.
Date of issuance: January 5, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 160 and 150.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 14, 2003 (68 FR
59221).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 5, 2004.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant, Units 2 and 3, Limestone County, Alabama
Date of application for amendments: April 14, 2003, as supplemented
by letters dated September 5 and November 7, 2003.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3.3.4.1, ``End-Of-Cycle Recirculation Pump Trip
(EOC[dash]RPT) Instrumentation,'' and TS 3.7.5, ``Main Turbine Bypass
System,'' to reference additional core limits adjustment factors for
linear heat generation rate for equipment out-of-service conditions.
Also, Section b of TS 5.65. ``Core Operating Limits Report (COLR),''
was revised to add references to the Framatome Advanced Nuclear Power
analytical methods what will be used in the upcoming fuel cycles to
determine core operating limits.
Date of issuance: December 30, 2003.
Effective date: Date of issuance, to be implemented within 60 days
from the completion of Unit 3 Spring 2004 and Unit 2 Spring 2005
refueling outages.
Amendment Nos.: 287 & 245.
Facility Operating License Nos. DPR-52, and DPR-68. Amendments
revised the TSs.
Date of initial notice in Federal Register: May 27, 2003 (68 FR
28858).
TVA's supplemental letters provided clarifying information that did
not expand the scope of the original application or change the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 30, 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 Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
[[Page 2751]]
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 email to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By February 19, 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 27, 2003 as supplemented by
letter dated December 27 and two letters dated December 28, 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 (SDG) 22 from
21 days to 113 days as a one-time change for the purpose of making
repairs to SDG 22.
Date of issuance: December 30, 2003.
Effective date: December 30, 2003.
Amendment No.: 149.
Facility Operating License No. NPF-80: Amendment revised the
Technical Specifications.
[[Page 2752]]
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 30,
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 13th day of January 2004.
For the Nuclear Regulatory Commission.
Eric J. Leeds,
Deputy Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 04-1104 Filed 1-16-04; 8:45 am]
BILLING CODE 7590-01-P