[Federal Register Volume 68, Number 82 (Tuesday, April 29, 2003)]
[Notices]
[Pages 22744-22759]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-10396]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to 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 April 18, 2003, through May 1, 2003. The
last biweekly notice was published on April 15, 2003, (68 FR 18269).
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 May 29, 2003, 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
[[Page 22745]]
leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of continuing disruptions in delivery of mail to United States
Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the request for
hearing and petition for leave to intervene should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and because of continuing disruptions in
delivery of mail to United States Government offices, it is requested
that copies be transmitted either by means of facsimile transmission to
301-415-3725 or by e-mail to [email protected]. A copy of the
request for hearing and petition for leave to intervene should also be
sent to the attorney for the licensee.
Non-timely 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].
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: March 20, 2003.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) and the licensing basis in the
Updated Safety Analysis Report (UFSAR) to support installation of a
passive low-pressure injection (LPI) cross connect inside containment.
The proposed changes to the TS would add requirements for the passive
LPI cross connect and eliminate requirements associated with the
capability to cross connect by manual operator action the trains
outside containment. The proposed changes to the UFSAR would revise the
licensing basis for a portion of the core flood and LPI/Decay Heat
Removal (DHR) piping to allow the exclusion of dynamic effects
associated with postulated pipe rupture of that piping by application
of leak-before-break technology for Unit 1. The proposed changes to the
UFSAR would also revise the licensing basis for selected portions of
the LPI/DHR piping to adopt the design requirements of Standard Review
Plan Section 3.6.2, Branch Technical Position MEB 3-1.
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
[[Page 22746]]
consideration, which is presented below: Pursuant to 10 CFR 50.91, Duke
Power Company (Duke) has made the determination that this amendment
request involves a No Significant Hazards Consideration by applying the
standards established by the NRC regulations in 10 CFR 50.92. This
ensures 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: The proposed LAR
[Licence Amendment Request] modifies the Technical Specifications
[(TS)] to incorporate new TS requirements associated with the new Low
Pressure Injection (LPI) System configuration and eliminate TS
requirements associated with the old LPI configuration. The proposed
LAR also modifies the licensing basis to adopt Standard Review Plan
(SRP) 3.6.2 Branch Technical Position (BTP) MEB 3-1 requirements for
selected portions of LPI piping and to credit Leak-Before-Break (LBB)
to allow the dynamic effects associated with postulated pipe rupture of
selected portions of the LPI/Core Flood (CF) piping to be excluded from
the design basis. The proposed design allowances for these selected
portions of piping continue to allow the LPI system design to meet GDC
[General Design Criterion] 4 requirements related to environmental and
dynamic effects. The proposed LAR will continue to ensure that ONS
[Oconee Nuclear Station] can meet design basis requirements associated
with the LPI safety function. The LPI System provides a means for
delivering a large volume of borated water to the reactor core
following postulated large pipe breaks in the Reactor Coolant System.
The planned modification adds a passive crossover connection between
the two LPI injection lines inside containment, along with necessary
check valves and flow orifices that will eliminate the need for time-
critical operator actions to manually open the LPI discharge header
outside containment. The new components will have the same pressure,
seismic, and quality group qualifications as the existing components in
the LPI system. The addition of the crossover line will enhance the
ability of the control room operator to mitigate the consequences of
specific events for which LPI is credited. Therefore, the proposed LAR
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The LPI system is also relied on to cool the reactor core during
unit shutdown. Hydraulic analyses have demonstrated that adequate LPI
flow is available for normal shutdown cooling with the new LPI piping
configuration.
2. Create the possibility of a new or different kind of accident
from any kind of accident previously evaluated: The proposed LAR
modifies the Technical Specification to incorporate new TS requirements
associated with the new LPI System configuration and eliminate TS
requirements associated with the old LPI System configuration. The
proposed LAR also modifies the licensing basis to adopt MEB 3-1
requirement for selected portions of LPI piping and to credit LBB to
allow the dynamic effects associated with postulated pipe rupture of
selected portions of the LPI/Core Flood (CF) piping to be excluded from
the design basis. The proposed design allowances for these selected
portions of piping continue to allow the LPI system design to meet GDC
4 requirements related to environmental and dynamic effects. The LPI
and Core Flood systems affected by implementing the proposed changes to
the TS are not assumed to initiate design basis accidents. The systems
affected by the changes are used to mitigate the consequences of an
accident that has already occurred. The proposed TS and licensing basis
changes do not affect the mitigating function of these systems.
Consequently, these changes do not alter the nature of events
postulated in the Safety Analysis Report nor do they introduce any
unique precursor mechanisms. Therefore, the proposed amendment will 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 proposed TS and licensing basis changes do not unfavorably
affect any plant safety limits, set points, or design parameters. The
changes also do not unfavorably affect the fuel, fuel cladding, RCS, or
containment integrity. Therefore, the proposed TS and licensing basis
changes, which adds TS requirements and adopts new design allowances
associated with the passive LPI cross connect modification, do not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC 20005.
NRC Section Chief: John A. Nakoski.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: March 19, 2003.
Description of amendment request: The proposed amendment deletes
requirements from the technical specifications (TS) and other elements
of the licensing bases to maintain a Post Accident Sampling System
(PASS). Licensees were generally required to implement PASS upgrades as
described in NUREG-0737, ``Clarification of TMI [Three Mile Island
Nuclear Station] Action Plan Requirements,'' and Regulatory Guide 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI, Unit 2 (TMI-2). Requirements
related to PASS were imposed by Order for many facilities and were
added to or included in the TS for nuclear power reactors currently
licensed to operate. Lessons learned and improvements implemented over
the last 20 years have shown that the information obtained from PASS
can be readily obtained through other means or is of little use in the
assessment and mitigation of accident conditions.
The changes are based on NRC-approved Technical Specification Task
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
413, ``Elimination of Requirements for a Post Accident Sampling System
(PASS).'' The U.S. Nuclear Regulatory Commission (NRC) staff issued a
notice of opportunity for comment in the Federal Register on December
27, 2001 (66 FR 66949), on possible amendments concerning TSTF-413,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on March 20, 2002 (67 FR 13027).
The licensee affirmed the applicability of the following NSHC
determination in its application dated March 19, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an
[[Page 22747]]
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 PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result of
the TMI-2 accident. The specific intent of the PASS was to provide a
system that has the capability to obtain and analyze samples of plant
fluids containing potentially high levels of radioactivity, without
exceeding plant personnel radiation exposure limits. Analytical results
of these samples would be used largely for verification purposes in
aiding the plant staff in assessing the extent of core damage and
subsequent offsite radiological dose projections. The system was not
intended to and does not serve a function for preventing accidents and
its elimination would not affect the probability of accidents
previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual benefit
to post accident mitigation. Past experience has indicated that there
exists in-plant instrumentation and methodologies available in lieu of
a PASS for collecting and assimilating information needed to assess
core damage following an accident. Furthermore, the implementation of
Severe Accident Management Guidance (SAMG) emphasizes accident
management strategies based on in-plant instruments. These strategies
provide guidance to the plant staff for mitigation and recovery from a
severe accident. Based on current severe accident management strategies
and guidelines, it is determined that the PASS provides little benefit
to the plant staff in coping with an accident.
The regulatory requirements for the PASS can be eliminated 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. The elimination of the PASS will not prevent an
accident management strategy that meets the initial intent of the post-
TMI-2 accident guidance 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 PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in 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 PASS related requirements will not result in any
failure mode not previously analyzed. The PASS was intended to allow
for verification of the extent of reactor core damage and also to
provide an input to offsite dose projection calculations. The PASS is
not considered an accident precursor, nor does its existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radioisotopes 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 PASS, 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. Methodologies that are not reliant on
PASS are designed 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 use of a PASS is redundant and does not provide quick
recognition of core events or rapid response to events in progress. The
intent of the requirements established as a result of the TMI-2
accident can be adequately met without reliance on a PASS.
Therefore, this change does not involve a significant reduction in
the margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.
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: March 26, 2003.
Description of amendment request: The amendment request proposes to
adopt the Boiling Water Reactor Vessel and Internals Project integrated
surveillance program (BWRVIP ISP) as the basis for demonstrating
compliance with the requirements of Appendix H to Title 10 of the Code
of Federal Regulations Part 50 (10 CFR 50), ``Reactor Vessel Material
Surveillance Program Requirements'' and delete Technical Specification
(TS) 4.6.A.5. The licensee also proposes to update the pressure-
temperature (P-T) curves through the end of the current operating
license by revising TS Figures 3.6.1, 3.6.2, and 3.6.3.
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's review is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Brittle fracture of the reactor pressure vessel (RPV) is not a
postulated or evaluated design basis accident. No evaluations of other
postulated accidents are affected by this proposed change. Because the
applicable regulatory requirements continue to be met, the change does
not significantly increase the probability of any accident previously
evaluated.
Also, the change will not alter any assumptions previously made in
evaluating the radiological consequences of accidents.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Create the possibility for a new or different kind of accident
from any previously evaluated.
The proposed change does not involve a modification of the design
of plant structures, systems, or components. The change will not impact
the manner in which the plant is operated and will not degrade the
reliability of structures, systems, or components important to safety
as equipment protection features will not be deleted or modified,
equipment redundancy or independence will not
[[Page 22748]]
be reduced, supporting system performance will not be affected, and no
severe testing of equipment will be imposed. No new failure modes or
mechanisms will be introduced as a result of this proposed change.
Therefore, the changes to the material surveillance program and
pressure-temperature limits that compose this proposed change do not
create the possibility of a new or different kind of accident than
those previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined. There is no change or impact on any safety
analysis assumption or in any other parameter affecting the course of
an accident analysis supporting the Bases of any Technical
Specification. The proposed change does not involve any increase in
calculated off-site dose consequences.
Therefore, this change does not involve a significant reduction in
a margin of safety.
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: James W. Clifford.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: February 3, 2003.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) 3/4.7.1.4, ``Turbine Cycle--
Specific Activity,'' and its associated bases. With the exception of TS
4.0.4, wording similar to that presented in the improved Standard
Technical Specifications will be adopted. The amendment request
proposes an exception to the requirements of TS 4.0.4 when entering
MODE 4, along with conditions for when the surveillance requirement
must be satisfied in MODE 4. Additionally there are editorial changes
to the TS Index reflecting the proposed revision.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR) 50.91(a), the licensee has provided its analysis
of the issue of no significant hazards consideration, which is
presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes, in part, modify the modes of applicability by
stating that TS 4.0.4 is not applicable for Mode 4 entry. For the
surveillance requirement, the change specifies the conditions in Mode 4
that are necessary to obtain a representative sample from the steam
generators. Analyzed events are assumed to be initiated by the failure
of plant structures, systems or components. The level of specific
activity contained in the reactor coolant is germane to the
consequences of an accident and is not related in any way to the
probability of failure of a plant structure, system or component which
would result in the occurrence of an unanalyzed event. Because the
probability of failure of plant equipment is not affected, there is no
impact on the probability of occurrence of a previously analyzed
accident.
The consequences of a previously analyzed event are dependent on
the initial conditions assumed for the analysis, and the availability
and successful functioning of the equipment assumed to operate in
response to the analyzed event. The proposed changes do not alter the
initial conditions assumed in the analysis of interest. The plant
parameters assumed for the analyses are maintained within assumed
limits through compliance with the Technical Specifications and plant
procedures. Additionally, the proposed changes do not impose any new
safety analyses limits. Any deviation from the allowable activity
limits will require the plant to be placed in a condition where the
specification does not apply. Therefore, the proposed changes do not
involve a significant increase in the consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new or
different kind of accident from any previously evaluated.
The proposed changes do not involve a physical alteration of the
plant. No new equipment is being introduced, and installed equipment is
not being operated in a new or different manner. There is no change
being made to the parameters within which the plant is operated, or to
the setpoints at which protective or mitigative actions are initiated.
No alteration in the procedures that ensure the plant remains within
analyzed limits is being proposed, and no change is being made to the
procedures relied upon to respond to an off-normal event. As such, no
new failure modes are being introduced. These changes have no physical
effect on any plant equipment. Therefore, the changes do not create the
possibility of a new of different kind of accident from any previously
evaluated.
3. The proposed changes do not involve a significant reduction in a
margin of safety.
The margin of safety is established through equipment design,
limitations on operating parameters, and the setpoints at which
automatic actions are initiated. No equipment design features are
impacted by these changes, no operating parameters are revised, and no
changes are proposed to the actuation setpoints. The limit on secondary
coolant Dose Equivalent Iodine remains at the current value of 0.1
microcuries per gram. Therefore, the proposed changes do not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis, and based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. S. Ross, Florida Power & Light Company,
PO Box 14000, Juno Beach, FL 33408-0420.
NRC Section Chief: James W. Clifford.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station, Unit 1 (NMP1), Oswego County, New York
Date of amendment request: October 7, 2002, as supplemented on
March 24, 2003.
Description of amendment request: The licensee's October 7, 2002,
application proposed to add Specification 4.0.3 to address missed
surveillances to the Technical Specifications (TSs). This new
specification specifies an initial 24-hour delay period for performing
a missed surveillance prescribed by Specification 3.0.3. Specification
4.0.3 will also require: ``A risk evaluation shall be performed for any
surveillance delayed greater than 24 hours and the risk impact shall be
managed.'' In addition, the licensee proposed to add wording to each of
the following existing specifications such that the new Specification
4.0.3 would apply to them: Specification 6.16, 6.17, 6.18, and 6.19. On
November 12, 2002, the Nuclear Regulatory Commission (NRC) staff
published a proposed no significant hazards consideration determination
and opportunity for a
[[Page 22749]]
hearing (67 FR 68739) for the October 7, 2002, application.
As a result of the NRC staff comments, the licensee supplemented
the application by a letter dated March 24, 2003. The supplement adds
new requirements related to the use and application of the surveillance
requirements (SRs) currently included in the TSs.
These new explicit SR applicability requirements would supersede
the more general current requirements. The proposed new requirements
reflect the current practices at NMP1, and as such, do not change any
existing method of plant 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 for the March 24, 2003, supplement, which is presented
below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
Adoption of new administrative requirements related to the proper
use of the surveillance requirements currently included in the NMP1 TSs
do not affect any accident initiator, and as such, will have no effect
on the probability of an accident. The proposed changes do not involve
physical changes to the plant or introduce any new modes of operation.
Accordingly, continued assurance is provided that the process
variables, structures, systems, and components are maintained such that
there will be no degradation of any fission product barrier which could
increase the radiological consequences of an accident. Therefore, the
proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Adoption of new administrative requirements related to the proper
use of the surveillance requirements currently included in the NMP1 TSs
will have no adverse effect on the design or assumed accident
performance of any structure, system, or component, or introduce any
new modes of system operation or failure modes. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes add new administrative requirements related to
the proper use of the surveillance requirements currently included in
the NMP1 TSs. The addition of requirements will make application of the
surveillance requirements more restrictive than currently required by
the TSs. Accordingly, 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
supplement of March 24, 2003, 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.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: March 27, 2003.
Description of amendment request: The proposed amendment would
revise Technical Specification Surveillance Requirement 3.1.4.1, ``Rod
Group Alignment Limits, to change the allowable alignment limits of
individual rods in Mode 1 when greater than 85-percent power.
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 Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant increase in
the probability or consequences of any accident previously evaluated.
This proposed change does not cause an increase in the
probabilities of any accidents previously evaluated because the change
will not cause an increase in the probability of any initiating events
for accidents previously evaluated.
The consequences of the accidents previously evaluated in the PBNP
[Point Beach Nuclear Plant] Final Safety Analysis Report (FSAR) are
determined by the results of analyses that are based on initial
conditions of the plant, the type of accident, transient response of
the plant, and the operation and failure of equipment and systems.
Based on the analyses documented in WCAP-15432, Revision 2
[``Conditional Extension of the Rod Misalignment Technical
Specification for Point Beach Units 1 and 2, (proprietary)'' dated
April 2001], all pertinent licensing-basis acceptance criteria have
been met and the margin of safety, as defined in the Technical
Specification Bases, is not significantly reduced in any of the Point
Beach licensing basis accident analyses due to the subject change.
Therefore, the probability of an accident previously evaluated has not
significantly increased. Because design limitations continue to be met
and the integrity of the reactor coolant system pressure boundary is
not challenged, the assumptions employed in the calculation of the
offsite radiological doses remain valid. Neither rod position
indication nor the limits on allowed rod position deviation is an
accident initiator or precursor. Therefore, the consequences of an
accident previously evaluated will not be significantly increased.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a new or different kind of
accident from any accident previously evaluated.
The changes described in the proposed amendment are supported by
the analyses provided in the submittal [the March 27, 2003,
application]. The evaluation of the effects of the proposed changes
indicates that all design standards and applicable safety criteria
limits are met. These changes therefore do not cause the initiation of
any new or different accident nor create any new failure mechanisms.
Equipment important to safety will continue to operate as designed.
The proposed change does not result in any event previously deemed
incredible being made credible. The change does not result in more
adverse conditions or result in any increase in the challenges to
safety systems. Therefore, operation of the Point Beach Nuclear Plant
in accordance with the proposed amendment will not create the
possibility of a new or different type of accident from any accident
previously evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant reduction in a
margin of safety.
Based on the analyses documented in WCAP-15432, Revision 2, all
pertinent licensing-basis acceptance criteria have been met and the
margin of safety, as defined in the Technical Specification
[[Page 22750]]
Bases, is not significantly reduced in any of the Point Beach licensing
basis accident analyses based on the subject changes to safety analyses
input parameter values. There are no new or significant changes to the
initial conditions contributing to accident severity or consequences.
Since the analyses in the accompanying submittals [March 27, 2003,
application and WCAP-15432] demonstrate that all applicable acceptance
criteria continue to be met, the subject operating conditions will not
involve a significant reduction in a margin of safety at Point Beach.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: March 25, 2003.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3.2.1 and TS 3.2.3 for
implementation of relaxed axial offset control of the reactor cores,
relocate selected operating parameters from TS 2.0 and TS 3.3.1 to the
Core Operating Limits Report (COLR), revise the Pressurizer Pressure-
Low Allowable Value, and revise the appropriate references in TS 5.6.5
to the NRC-approved methodologies which support relocation of operating
parameters to the COLR.
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:
Group 1--Implementation of Relaxed Axial Offset Control
A. TS 3.2.1, Heat Flux Hot Channel Factor--FQ(Z) and
Bases: Modification of Required Actions and Completion Time if
FWQ(Z) is not within its limit and update Bases.
B. TS 3.2.3, Axial Flux Difference (AFD) and Bases: Modification
of Limiting Conditions for Operation, Actions and Surveillance
Requirements and revision of the Bases.
This license amendment request proposes to revise the Technical
Specifications to implement the relaxed axial offset control
methodology to address the heat flux hot channel factor and axial
flux difference limits.
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
This license amendment request proposes to revise the Technical
Specifications to implement the relaxed axial offset control
methodology to address the heat flux hot channel factor and axial
flux difference limits. The revised Technical Specifications and
parameter changes associated with relaxed axial offset control
assure that the limiting safety analysis inputs (such as, heat flux
hot channel factor and axial flux difference limits) are not
exceeded. The bounding power distribution transient factor values,
W(Z), and the axial flux difference limits that are documented in
the Core Operating Limits Report will be determined by NRC approved
analytical methods and will be validated as part of the cycle
specific reload evaluation process.
Heat flux hot channel factors and axial flux difference limits
are not assumed accident initiators. Therefore, the relaxed axial
offset control related Technical Specification changes do not
involve a significant increase in the probability of an accident.
Likewise, operation of the plant within the proposed Technical
Specification controls and limits assures that safety analysis
assumptions are met, thus, if an accident were to occur, the
consequences would continue to be bounded by the accident analyses.
Therefore, the relaxed axial offset control related technical
specification changes do not involve a significant increase in the
consequences of an accident.
The relaxed axial offset control related technical specification
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
This proposed change does not involve a physical alteration of
the plant; that is, no new or different type of equipment will be
installed. This proposed change does not introduce any new mode of
plant operation or change the methods governing normal plant
operation. No new failure mode has been created and no new equipment
performance burdens are imposed. Therefore the possibility of a new
or different kind of accident from those previously analyzed has not
been created.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
This license amendment request proposes to revise the Technical
Specifications to implement the relaxed axial offset control
methodology to address the heat flux hot channel factor and axial
flux difference limits. The supporting Technical Specification
limits are defined by NRC approved analytical methods which are
performed to conservatively bound the operating conditions defined
by the Technical Specifications and to demonstrate meeting the
regulatory acceptance limits. The heat flux hot channel factor
licensed safety margins are maintained. The heat flux hot channel
factor conforms to plant design bases and limits actual plant
operation within analyzed and licensed boundaries. The relaxed axial
offset control methodology has been demonstrated to ensure that core
heat flux hot channel factors will remain below accident analysis
limits. The margin of safety provided by the analyses in accordance
with the acceptance limits is maintained and not reduced. Thus, the
implementation of relaxed axial offset control at Prairie Island
does not involve a significant reduction in a margin of safety.
Group 2--Relocation of Technical Specifications Safety Limits Figure
and Overtemperature Delta-T and Overpower Delta-T Parameter Values to
the Core Operating Limits Report, and Miscellaneous Administrative
Changes
A. TS 2.1.1, ``Reactor Core SLs [Safety Limits]'' and Bases:
Relocate the safety limits Figure to the Core Operating Limits
Report, update TS 2.1.1 and Bases.
B. TS 3.3.1, Table 3.3.1-1 (Pages 2, 7 and 8), ``Reactor Trip
System Instrumentation'', Overpower Delta-T Trip Function, and
Overtemperature Delta-T and Overpower Delta-T parameter values:
Delete SR [Surveillance Requirement] 3.3.1.3, SR 3.3.1.6, and remove
f(DI) from Overpower Delta-T Trip Function, relocate overtemperature
delta-T and overpower delta-T parameter values and revise the Bases.
C. TS 5.6.5, Core Operating Limits Report (COLR): Additions to
document Technical Specifications with limits in the Core Operating
Limits Report and the analytical methods used to determine the
values for relocated safety limits and overtemperature delta-T and
overpower delta-T parameters and miscellaneous administrative
changes.
This license amendment request proposes to relocate the safety
limits and overtemperature delta-T and overpower delta-T parameter
values to the Core Operating Limits Report. Relocation of these
limits and parameter values to the Core Operating Limits Report
allows them to be changed under licensee controls. This license
amendment also proposes to include, in the Technical Specifications
administrative controls section, the appropriate references to the
NRC approved methodologies which will be used to determine the
safety limits and overtemperature delta-T and overpower delta-T
parameter values. These changes are acceptable because the values
used to operate the Prairie Island plant will be determined using
NRC approved methods and these changes are consistent with the
guidance of the industry standard Technical Specifications, NUREG-
1431, Revision 2, ``Standard Technical Specifications Westinghouse
Plants''. This license amendment request also proposes to delete
references to an NRC Safety Evaluation and make some editorial
corrections in the Technical Specifications administrative controls
section. These changes are
[[Page 22751]]
acceptable since they are administrative and do not affect plant
operation.
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
This license amendment request proposes to relocate the safety
limits and overtemperature delta-T and overpower delta-T parameter
values to the Core Operating Limits Report and to include, in the
Technical Specifications administrative controls section, the
appropriate references to the NRC approved methodologies which
support determination of these limits and parameter values. The
safety limits and overtemperature delta-T and overpower delta-T
parameter values that are documented in the Core Operating Limits
Report will be determined by NRC approved analytical methods and
will be validated as part of the cycle specific reload evaluation
process.
Safety limits are not assumed accident initiators. Thus
relocation of the safety limits does not involve a significant
increase in the probability of an accident. Overtemperature delta-T
and overpower delta-T parameter values are inputs to the reactor
trip system which is provided to mitigate the consequences of an
accident. The reactor trip system is not an accident initiator and
therefore, changes to input values do not increase the probability
of an accident.
Safety limits define bounding values within which plant
operation will not initiate an accident condition. Safety limits
relocated to the Core Operating Limits Report and determined by use
of NRC approved methodologies will continue to determine the safe
limits of plant operation, thus this change does not involve a
significant increase in the consequences of an accident. The reactor
trip system, with inputs from the overtemperature delta-T and
overpower delta-T trip functions, mitigates the consequences of
accidents.
The overtemperature delta-T and overpower delta-T trip parameter
values are determined to assure that the design limit departure from
nucleate boiling ratio is met and fuel integrity is maintained.
Overtemperature delta-T and overpower delta-T trip parameters
relocated to the Core Operating Limits Report and values determined
by use of NRC approved methodologies will continue to determine the
inputs for these trip functions which mitigate the design basis
accident consequences, thus this change does not involve a
significant increase in the consequences of an accident.
Addition of references to NRC approved methodologies in the
Technical Specifications administrative controls section is an
administrative change which does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed miscellaneous administrative changes in the
Technical Specifications administrative controls section do not
affect plant operation and therefore do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
As discussed above, these proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
The malfunction of safety related equipment, assumed to be
operable in the accident analyses, would not be impacted as a result
of the proposed technical specification changes. No new failure mode
has been created and no new equipment performance burdens are
imposed. Therefore the possibility of a new or different kind of
accident from those previously analyzed has not been created. The
proposed administrative changes do not create the possibility of a
new or different kind of accident from those previously analyzed.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
This license amendment request proposes to relocate the safety
limits and overtemperature delta-T and overpower delta-T parameter
values to the Core Operating Limits Report and to include, in the
Technical Specifications administrative controls section, the
appropriate references to the NRC approved methodologies which
support determination of these limits and parameter values. This
proposed change also allows these relocated limits and parameter
values to be changed under licensee controls. Safety limits in the
Core Operating Limits Report will be determined by use of NRC
approved methodologies and will continue to determine the safe
limits of plant operation. Overtemperature delta-T and overpower
delta-T trip parameter values in the Core Operating Limits Report
will be determined by use of NRC approved methodologies and will
continue to determine the inputs for these trip functions which
mitigate design basis accidents. The Safety Limits licensed safety
margins are maintained. The Safety Limits conform to plant design
bases and limit actual plant operation within analyzed and licensed
boundaries. The methodology described in WCAP-8745, along with the
low pressurizer pressure allowable value, ensures that the
overtemperature delta-T and overpower delta-T trips will protect
against fuel centerline melting and departure from nucleate boiling
during Condition II events. Thus, these changes do not involve a
significant reduction in the margin of safety.
This license amendment request proposes to delete references to
an NRC Safety Evaluation and make some editorial corrections in the
Technical Specifications administrative controls section. These
changes are administrative and thus do not involve a significant
reduction in the margin of safety.
Group 3--Revision of Pressurizer Pressure-Low reactor trip Allowable
Value
TS 3.3.1, Table 3.3.1-1 (Page 2), ``Reactor Trip System
Instrumentation'', Function 8.a, Pressurizer Pressure-Low: Increase
Pressurizer Pressure-Low Allowable Value.
This license amendment request proposes to increase the
Allowable Value defined in Table 3.3.1-1 for the Pressurizer
Pressure-Low reactor trip.
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
This license amendment request proposes to increase the
Allowable Value defined in Table 3.3.1-1 for the Pressurizer
Pressure-Low reactor trip. Pressurizer Pressure-Low reactor trip is
an input to the reactor trip system which is provided to mitigate
the consequences of an accident. The reactor trip system is not an
accident initiator and therefore, changes to the Pressurizer
Pressure-Low Allowable Value do not involve an increase in the
probability of an accident.
The Pressurizer Pressure-Low Allowable Value is being increased
which is a conservative change. The increase in the Pressurizer
Pressure-Low reactor trip Allowable Value will assure that the
overtemperature delta-T and overpower delta-T reactor trip
functions, with values determined in accordance with NRC approved
methodologies, provide protection against fuel centerline melting
and departure from nucleate boiling for overpower and
overtemperature events. Therefore, this change does not involve an
increase in the consequences of an accident previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
This proposed change does not involve a physical alteration of
the plant; that is, no new or different type of equipment will be
installed. This proposed change does not introduce any new mode of
plant operation or change the methods governing normal plant
operation. No new failure mode has been created and no new equipment
performance burdens are imposed. Therefore the possibility of a new
or different kind of accident from those previously analyzed has not
been created.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
This license amendment request proposes to increase the
Allowable Value defined in Table 3.3.1-1 for the Pressurizer
Pressure-Low reactor trip. The Allowable Value is determined in
accordance with an NRC accepted setpoint methodology with input from
NRC approved analytical methods. These determinations are performed
to conservatively bound the operating conditions defined by the
Technical Specifications and to demonstrate meeting the regulatory
acceptance limits.
Performance of analyses and evaluations for the cycle specific
reload evaluation process will confirm that the operating envelope
defined by the Technical Specifications continues to be bounded by
the analytical basis and in no case exceeds the acceptance limits.
The proposed Pressurizer Pressure-Low Allowable Value along with the
overtemperature delta-T and overpower delta-T trips will protect
against fuel centerline melting and departure from nucleate boiling
during Condition II events. The proposed Allowable Value conforms to
plant design bases and limits actual plant
[[Page 22752]]
operation within analyzed and licensed boundaries. The margin of
safety provided by the proposed Pressurizer Pressure-Low Allowable
Value is maintained and not reduced. Thus, the increase in the
Pressurizer Pressure-Low reactor trip Allowable Value 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: March 3, 2003.
Description of amendment request: The proposed amendments would
delete requirements from the technical specifications (TS) and other
elements of the licensing bases to maintain a Post-Accident Sampling
System (PASS). Licensees were generally required to implement PASS
upgrades as described in NUREG-0737, ``Clarification of TMI [Three Mile
Island] Action Plan Requirements,'' and Regulatory Guide 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI Unit 2. Requirements related to
PASS were imposed by Order for many facilities and were added to or
included in the TSs for nuclear power reactors currently licensed to
operate. Lessons learned and improvements implemented over the last 20
years have shown that the information obtained from PASS can be readily
obtained through other means or is of little use in the assessment and
mitigation of accident conditions.
The proposed changes are based on NRC-approved Technical
Specification Task Force (TSTF) Standard Technical Specification Change
Traveler, TSTF-413, ``Elimination of Requirements for a Post-Accident
Sampling System (PASS).'' The NRC staff issued a notice of opportunity
for comment in the Federal Register on December 27, 2001 (66 FR 66949),
on possible amendments concerning TSTF-413, including a model safety
evaluation and model no significant hazards consideration (NSHC)
determination, using the consolidated line item improvement process.
The NRC staff subsequently issued a notice of availability of the
models for referencing in license amendment applications in the Federal
Register on March 20, 2002 (67 FR 13027). The licensee affirmed the
applicability of the following NSHC determination in its application
dated March 3, 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 PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
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. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance 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 PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in 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 PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radioisotopes
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 PASS, 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. Methodologies that are
not reliant on PASS are designed 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 use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: 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: December 23, 2002.
Description of amendment request: The amendment would change the
Hope
[[Page 22753]]
Creek Generating Station (HCGS) reactor vessel material surveillance
program required by Appendix H to Title 10 of the Code of Federal
Regulations (10 CFR) part 50. This change would incorporate the Boiling
Water Reactor Vessel and Internals Project Integrated Surveillance
Program (ISP) into the HCGS 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. Does the change involve a significant increase in the
probability or consequences of an accident previously analyzed?
Response: No.
The proposed change implements an integrated surveillance
program that has been evaluated by the NRC staff as meeting the
requirements of paragraph Ill.C of Appendix H to 10 CFR 50.
Consequently, the proposed change does not significantly increase
the probability of any accident previously evaluated. The proposed
change provides the same assurance of RPV [reactor pressure vessel]
integrity. As a result, the consequences of any accident previously
evaluated are not significantly increased.
Therefore, this proposed amendment does not involve a
significant increase in the probability of occurrence or
consequences of an accident previously analyzed.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously analyzed?
Response: No.
The proposed change revises the HCGS licensing basis to reflect
participation in the ISP. The proposed change does not involve a
modification of the design of plant structures, systems or
components (SSC). Also, the proposed change will not degrade the
reliability of SSCs important to safety since protective features
will not be deleted or modified. The proposed change will not impact
the manner in which the plant is normally operated. The proposed
change maintains an equivalent level of RPV material surveillance
and does not introduce any new accident initiators. Therefore, this
proposed amendment does not create the possibility of a new or
different kind of accident from any previously analyzed.
3. Does the change involve a significant reduction in the margin
of safety?
Response: No.
The proposed change has been evaluated as providing an
acceptable alternative to the plant-specific RPV material
surveillance program that meets the requirements of the regulations
for RPV material surveillance. Therefore, these changes do not
involve a significant reduction in 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: James W. Clifford.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: September 20, 2002, as revised on
February 14, 2003. This notice supercedes a previous notice (67 FR
75884) published on December 10, 2002, which was based on the
licensee's application dated September 20, 2002.
Description of amendment request: The proposed amendment will: (1)
Add a new limiting condition for operation (LCO) for spent fuel pool
(SFP) boron concentration; (2) relocate requirements from Technical
Specification (TS) Section 5.0, ``Design Features,'' to a new LCO in TS
Section 3/4.7; and (3) revise existing TS 3/4.9.1 for refueling
operations by relocating requirements for boron concentration to the
Core Operating Limits Report (COLR) described in TS 6.9.1.9. The
licensee also proposed related changes to the TS Bases. By letter dated
February 14, 2003, PSEG revised its request, including lowering the
minimum SFP boron concentration from 2300 parts per million (ppm) to
800 ppm.
Therefore, this notice supercedes a previous notice published on
December 10, 2002, to reflect this change.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR), Section 50.91(a), the licensee has provided a
revised analysis of the issue of no significant hazards consideration.
The NRC staff has reviewed the licensee's analysis against the
standards of 10 CFR 50.92(c). The NRC staff's review is presented
below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The licensee proposed to change the Salem Nuclear Generating
Station (Salem) TSs by: (1) adding a new LCO for SFP boron
concentration; (2) relocating requirements from TS Section 5.0,
``Design Features,'' to a new LCO in TS Section 3/4.7; and (3) revising
existing TS 3/4.9.1 for refueling operations by relocating requirements
for boron concentration to the COLR. These changes are consistent with
applicable LCOs in NUREG-1431, Revision 2, ``Improved Standard
Technical Specifications, Westinghouse Plants,'' and will continue to
provide administrative controls to ensure that a proper boron
concentration is maintained in accordance with Salem's accident
analyses. Because there are no changes to any of the input assumptions
associated with postulated accidents involving refueling operations and
the SFP, the proposed amendment does not involve a significant increase
in the probability of occurrence or consequences of an accident
previously analyzed.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously analyzed?
Adding new LCOs for boron concentration in the SFP and relocating
boron concentration requirements to the COLR will not change the
conduct of operations in the SFP, refueling cavity and fuel transfer
tube at Salem. Therefore, because plant operations will not change, the
proposed amendment does not create the possibility of a new or
different kind of accident from any previously analyzed.
3. Does the proposed change involve a significant reduction in the
margin of safety?
Refueling operations and SFP boron concentration limits will be
based on approved methodologies and accident analyses that are
unchanged as a result of the proposed TS amendments. Therefore, because
existing margins of safety will be maintained, the proposed change does
not involve a significant reduction in the margin of safety.
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: James W. Clifford.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: February 28, 2003.
Description of amendment requests: The proposed license amendments
would revise Action A of Technical Specification (TS) 3.5.2, ``ECCS--
Operating,'' to change the completion time for restoring centrifugal
charging pump (CCP) 1-1 to operable status during Diablo Canyon Power
Plant (DCPP) Unit 1 Cycle 12, from 72 hours to 7 days. The 72-hour
allowed completion time is not sufficient to
[[Page 22754]]
accomplish such emergent repairs on an inoperable CCP. This license
amendment request also removes a similar one-time change for DCPP Unit
2 CCP 2-1 which has expired.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The emergency core cooling system (ECCS) and the centrifugal
charging pumps (CCPs) are designed to respond to mitigate the
consequences of an accident. They are not an accident initiator, and
as such cannot increase the probability of an accident.
The loss of both CCPs, due to an inoperable CCP 1-1 and a single
failure of CCP 1-2, could increase the consequences of an accident.
A probabilistic risk assessment was performed to evaluate the
increased consequences. The worst case risk increment due to the
increased completion time for CCP 1-1 and the maximum allowed
results in only a small quantitative impact on plant risk.
Allowing 7 days to complete the seal replacement and post-
maintenance testing of CCP 1-1 is acceptable since the ECCS system
remains capable of performing its intended function of providing at
least the minimum flow assumed in the accident analyses. During the
extended maintenance and test period, appropriate compensatory
measures will be implemented to restrict high risk activity. The
consequences of accidents, which rely on the ECCS system, will not
be significantly affected.
Therefore, the proposed changes will not result in 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.
There are no new failure modes or mechanisms created due to
plant operation for an extended period to perform repairs and post-
maintenance testing of CCP 1-1. Extended operation with an
inoperable CCP does not involve any modification in the operational
limits or physical design of the systems. There are no new accident
precursors generated due to the extended allowed completion time.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Plant operation for seven days with an inoperable CCP 1-1 does
not adversely affect the margin of safety. During the extended
allowable completion time the ECCS system maintains the ability to
perform its safety function of providing at least the minimum flow
assumed in the accident analyses. During the extended maintenance
and test period, appropriate compensatory measures will be
implemented to restrict high-risk activity.
Therefore, the change does not involve a significant reduction
in a margin of safety as defined in the basis for any Technical
Specification.
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: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, PO Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed no Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Florida Power & Light Company, et al. (FPL's), Docket Nos. 50-335, and
50-389, St. Lucie Plant, Unit No. 1, and Unit No. 2, St. Lucie County,
Florida
Date of amendment request: October 23, 2002.
Description of amendment request: The proposed license amendments
would revise the Technical Specifications to include the design of a
new cask pit spent fuel storage rack for each unit to increase the
allowable spent fuel wet storage capacity at both units and include the
description of Boral TM as the neutron absorbing material
used in the new cask pit storage racks. The proposal would also revise
the spent fuel pool thermal-hydraulic analyses for core offload times
and include a change in FPL's commitments regarding the Unit 2 spent
fuel cooling system design basis described in the Updated Final Safety
Analysis Report.
Date of publication of individual notice in the Federal Register:
January 28, 2003 (68 FR 4244), as corrected March 31, 2003 (64 FR
15487).
Expiration date of individual notice: February 27, 2003.
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant,
Unit 1, Hamilton County, Tennessee
Date of application for amendments: February 14, 2003.
Description of amendments request: Revise the Updated Final
Analysis Report to change the methodology using a through-bolted
connection frame that is different than the original design and
construction of the steam generator roof compartment.
Date of publication of individual notice in the Federal Register:
March 14, 2003 (68 FR 12382).
Expiration date of individual notice: April 14, 2003.
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant,
Unit 1, Hamilton County, Tennessee
Date of application for amendments: March 18, 2003.
Description of amendments request: Revise the Updated Final
Analysis Report to provide an alternative methodology using a Bar-Lock
mechanical splice in lieu of the Cadweld splice used in the original
design and construction of the concrete shield building dome.
Date of publication of individual notice in the Federal Register:
March 17, 2003 (68 FR 12718).
Expiration date of individual notice: April 16, 2003.
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
[[Page 22755]]
published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) The
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by e-mail to [email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: November 27, 2002.
Brief description of amendment: The amendment deleted Section 6.17,
``Post Accident Sampling,'' and thereby eliminating the requirements to
have and maintain the subject system. The subject requirements were
imposed by a July 7, 1981, Nuclear Regulatory Commission Confirmatory
Order.
Date of Issuance: April 4, 2003.
Effective date: As of the date of issuance and shall be implemented
within 180 days.
Amendment No.: 237.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 21, 2003 (68 FR
2798).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated April 4, 2003.
No significant hazards consideration comments received: No.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: June 11, 2002, as supplemented
January 22, 2003.
Brief description of amendments: The amendment changes Technical
Specifications 3.7.11 related to the operation of the spent fuel pool
exhaust ventilation system during the movement of irradiated fuel
assemblies.
Date of issuance: April 7, 2003.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 234, 257.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 15, 2002 (67 FR
63689).
The January 22, 2003, supplemental letter provided clarifying
information that did not enlarge the scope of the amendments as noticed
in the original Federal Register notice or change the initial proposed
no significant hazards consideration determination. The Commission's
related evaluation of these amendments is contained in a Safety
Evaluation dated April 7, 2003.
No significant hazards consideration comments received: No.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: July 17 and August 6, 2002.
Brief description of amendments: These amendments permit operation
of Calvert Cliffs Unit 2 with a core containing up to eight lead fuel
assemblies with fuel rods clad with an advanced zirconium-based alloy.
Date of issuance: April 14, 2003.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 258 and 235.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 17, 2002 (67
FR 58637).
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated April 14, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: October 10, 2002, as
supplemented on November 22, 2002, and January 28, 2003. The October
10, 2002, application replaced the original application dated December
12, 2001.
Brief description of amendment: This amendment changes Technical
Specification (TS) Tables 3.2.A, 3.2.B, 4.2.A, and 4.2.B. The proposed
changes affect various instrument trip level settings and decrease
calibration frequencies for a variety of instruments. The proposed
changes identify that the Reactor Water Cleanup (RWCU) system requires
one channel in each of the two trip systems for each location. The
proposed changes also clarify the titles of certain trip systems, move
note numbers to their proper location, and correct a mis-referenced
figure in a table note. Appropriate Bases pages were also changed to
reflect the TS changes.
Date of issuance: April 17, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 90 days.
Amendment No.: 198.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 18, 2003, (68
FR 7815).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 17, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of application for amendments: September 27, 2002.
Brief description of amendments: The amendments revise Appendix B,
``Environmental Protection Plan (Non-Radiological),'' of the licenses
to remove a parenthetical reference to a superseded section of 10 CFR
part 51.
Date of issuance: April 4, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 132/132.
Facility Operating License Nos. NPF-37 and NPF-66: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 29, 2002 (67 FR
66009).
[[Page 22756]]
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 4, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: June 4, 2002, as supplemented by
letter dated February 19, 2003.
Brief description of amendment: This amendment revises Technical
Specification (TS) Surveillance Requirement (SR) 4.0.3 to extend the
delay period, before entering a Limiting Condition for Operation,
following a missed surveillance to ``* * * up to 24 hours or up to the
limit of the specified frequency, whichever is greater.'' In addition,
the amendment adds requirements to SR 4.0.3 to perform a risk
evaluation for any Surveillance delayed greater than 24 hours and
manage the risk impact, and specifies actions to be taken when a
delayed surveillance is not performed or not met. The amendment is
consistent with TS Task Force traveler TSTF-358, which has been
approved by the Nuclear Regulatory Commission for incorporation into
standard technical specifications in NUREG-1430. The TS Bases will be
revised under the licensee's existing TS Bases control program to be
consistent with the bases for TSTF-358.
Date of issuance: April 11, 2003.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 254.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 7, 2003 (68 FR
804).
The supplemental information contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 11, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: June 10, 2002.
Brief description of amendment: This amendment revises Surveillance
Requirement (SR) 3.0.3 to extend the delay period, before entering a
Limiting Condition for Operation, following a missed surveillance. The
delay period is extended from the current limit of ``* * * up to 24
hours or up to the limit of the specified Frequency, whichever is
less'' to ``* * * up to 24 hours or up to the limit of the specified
Frequency, whichever is greater.'' In addition, the following
requirement is added to SR 3.0.3: ``A risk evaluation shall be
performed for any Surveillance delayed greater than 24 hours and the
risk impact shall be managed.''
Date of issuance: April 17, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 125.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 18, 2003 (68 FR
12954).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 17, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: March 14, 2002, as supplemented
by letter dated January 20, 2003.
Brief description of amendment: This amendment revised technical
specification (TS) 5.5.12, ``Primary Containment Leakage Rate Testing
Program,'' to allow a one-time exception to Nuclear Energy Institute
94-01, ``Industry Guidance for Implementing Performance-Based Option of
10 CFR part 50 Appendix J,'' that extends the test interval of the
containment integrated leak rate test from 10 to 15 years.
Date of issuance: April 8, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 126.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5676).
The supplemental information contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 8, 2003.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: August 15, 2002, as
supplemented December 13, 2002.
Brief description of amendments: The amendments revise Technical
Specifications Section 6.8.4.h, Containment Leakage Rate Testing
Program, to allow a one-time 5-year extension to the current 10-year
test interval for the containment integrated leak rate test (ILRT). The
changes were submitted on a risk-informed basis as described in
Regulatory Guide 1.174, An Approach for Using Probabilistic Risk
Assessment in Risk-Informed Decisions on Plant-Specific Changes to the
Licensing Basis. The risk-informed analysis supporting the changes
indicates that the increase in risk from extending the ILRT test
interval from 10 to 15 years is insignificant.
Date of Issuance: April 10, 2003.
Effective Date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 187 & 130.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 17, 2002 (67
FR 58647).
The supplement dated December 13, 2002, provided clarifying
information that did not change the scope of the August 15, 2002,
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 April 10, 2003.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: June 28, 2002, as supplemented
December 18, 2002, January 18, 2003, and February 25, 2003.
Brief description of amendment: The amendment relaxes certain
Technical Specifications (TSs) requirements for containment isolation
and removes references to the Filtration Recirculation and Ventilation
System charcoal filters.
Date of issuance: April 15, 2003.
[[Page 22757]]
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 146.
Facility Operating License No. NPF-57: This amendment revised the
TSs.
Date of initial notice in Federal Register: February 18, 2003 (68
FR 7818).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 15, 2003.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: October 9, 2002, as supplemented
November 22, 2002, and December 6, 2002.
Brief description of amendment: The amendment grants, on a one-time
basis, an extension of the Type A Integrated Leak Rate Test interval
from 10 years to 15 years.
Date of issuance: April 16, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 147.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 18, 2003 (68
FR 7819).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 16, 2003.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: October 23, 2002.
Brief description of amendments: The amendments revise the Salem
Nuclear Generating Station, Unit Nos. 1 and 2, Technical Specification
(TS) 6.12, ``High Radiation Area'' to be consistent with the Standard
TSs for Westinghouse Plants (NUREG-1431, Revision 2) by updating the
current reference to Title 10 of the Code of Federal Regulations (10
CFR), Section 20.203 with the corresponding reference to 10 CFR
20.1601.
Date of issuance: April 10, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment Nos.: 255 and 236.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 4, 2002 (68 FR
5681).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 10, 2003.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: July 25, 2002, as supplemented
October 21, 2002.
Brief description of amendments: The amendments revise Technical
Specifications (TSs) Surveillance Requirement (SR) 4.0.3 to extend the
delay period, before entering a Limiting Condition for Operation,
following a missed surveillance. The delay period is extended from the
current limit of up to 24 hours to ``* * * up to 24 hours or up to the
limit of the specified frequency, whichever is greater.'' In addition,
the following requirement is added to SR 4.0.3: ``A risk evaluation
shall be performed for any surveillance delayed greater than 24 hours
and the risk impact shall be managed.'' The amendments also add a
requirement for a TS Bases Control Program to the administrative
controls section of TSs and makes administrative changes to SRs 4.0.1
and 4.0.3 to be consistent with NUREG-1431, Revision 2, ``Standard
Technical Specifications Westinghouse Plants.''
Date of issuance: April 16, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment Nos.: 256 and 237.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the TSs.
Date of initial notice in Federal Register: February 18, 2003 (68
FR 7820).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 16, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant , Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: August 1, 2002.
Description of amendments request: The amendments revised the
Updated Safety Analysis Report (UFSAR) to eliminate consideration of a
pressure regulator downscale failure as an abnormal operational
transient.
Date of issuance: April 4, 2003.
Effective date: As of the date of issuance, to be incorporated into
the UFSAR at the time of its next update.
Amendment Nos.: 244, 281 and 239.
Facility Operating License Nos. DPR-33, DPR-52, and DPR-68:
Amendments revised the UFSAR.
Date of initial notice in Federal Register: October 15, 2002 (67 FR
63697).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 4, 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
[[Page 22758]]
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
Room, located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland. Publicly available
records will be accessible from the Agencywide Documents Assess and
Management System's (ADAMS) Public Electronic Reading Room on the
Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html.
If you do not have access to ADAMS or if there are problems in
accessing the documents located in ADAMS, contact the NRC Public
Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or
by e-mail to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By May 16, 2003, 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,
[[Page 22759]]
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).
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: April 14, 2003, as supplemented by
letter dated April 15, 2003.
Description of amendment request: The amendment revises Limiting
Condition for Operation (LCO) 3.7.5, ``Control Building Chiller (CBC)
System,'' Required Action A.1 to add a provision that temporarily
removes the restrictions of LCO 3.0.4 until May 16, 2003. This
amendment allows entry into LCO 3.7.5 with an inoperable CBC subsystem.
Date of issuance: April 16, 2003.
Effective date: As of the date of issuance and shall be implemented
immediately.
Amendment No.: 250.
Facility Operating License No. DPR-49: Amendment revises the
technical specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No.
The Commission's related evaluation of the amendment, finding of
emergency circumstances, state consultation, and final NSHC
determination are contained in a safety evaluation dated April 16,
2003.
Attorney for licensee: Mr. Alvin Gutterman, Morgan Lewis, 1111
Pennsylvania Avenue NW., Washington, DC 20004
NRC Section Chief: L. Raghavan.
Dated at Rockville, Maryland, this 21st day of April, 2003.
For the Nuclear Regulatory Commission
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 03-10396 Filed 4-28-03; 8:45 am]
BILLING CODE 7590-01-P