[Federal Register Volume 68, Number 13 (Tuesday, January 21, 2003)]
[Notices]
[Pages 2796-2810]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-1161]
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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, December 27, 2002 through January 9, 2003.
The last biweekly notice was published on January 7, 2003 (68 FR 798).
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, 11555 Rockville
Pike (first floor), Rockville, Maryland. The filing of requests for a
hearing and petitions for leave to intervene is discussed below.
By February 20, 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
[[Page 2797]]
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,\1\ which is available at the
Commission's PDR, located at One White Flint North, 11555 Rockville
Pike (first floor), Rockville, Maryland. Publicly available records
will be accessible from the Agencywide Documents Access and Management
System's (ADAMS) Public Electronic Reading Room on the Internet at the
NRC web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a
request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
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\1\ The most recent version of Title 10 of the Code of Federal
Regulations, published January 1, 2002, inadvertently omitted the
last sentence of 10 CFR 2.714 (d) and paragraphs (d)(1) and (d)(2)
regarding petitions to intervene and contentions. For the complete,
corrected text of 10 CFR 2.714(d), please see 67 FR 20884; April 29,
2002.
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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, 11555 Rockville Pike (first floor),
Rockville, Maryland, by the above date. Because of continuing
disruptions in delivery of mail to United States Government offices, it
is requested that petitions for leave to intervene and requests for
hearing be transmitted to the Secretary of the Commission either by
means of facsimile transmission to 301-415-1101 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and because of continuing disruptions in delivery of mail to United
States Government offices, it is requested that copies be transmitted
either by means of facsimile transmission to 301-415-3725 or by e-mail
to [email protected]. A copy of the request for hearing and
petition for leave to intervene should also be sent to the attorney for
the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, 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, 304-415-4737 or by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: November 27, 2002.
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
[[Page 2798]]
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 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 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 November 27, 2002.
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: Edward J. Cullen, Deputy General Counsel
Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: November 27, 2002.
Description of amendment request: The proposed amendment would
delete requirements from the Technical Specifications (TSs) 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. However, 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 Nuclear Regulatory Commission (NRC)-
approved Technical Specification Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-413, ``Elimination of Requirements
for a
[[Page 2799]]
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 November 27, 2002.
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.
Based upon the reasoning presented above and the previous
discussion of the proposed amendment, the NRC staff proposes to
determine that the requested amendment does not involve a significant
hazards consideration.
Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius,
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Richard J. Laufer.
AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New York
Date of amendment request: December 16, 2002.
Description of amendment request: The licensee proposed to amend
the Oyster Creek Nuclear Generating Station (OCNGS) Technical
Specifications (TSs) regarding the safety limit minimum critical power
ratio (SLMCPR) to reflect the results of cycle-specific calculations
performed for the current fuel cycle (i.e., Cycle 19), using Nuclear
Regulatory Commission (NRC)-approved methodology for determining SLMCPR
values. Specifically, the licensee proposed to revise TS 2.1.A,
changing the SLMCPR from 1.12 to 1.10 for three-recirculation-loop
operation, and from 1.11 to 1.09 for four-or five-recirculation-loop
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. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is
presented below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Before commencement of Cycle 19, the licensee used NRC-approved
methods and procedures in Topical Report NEDE-24011-P-A-14, ``General
Electric Standard Application for Reactor Fuel'' (GESTAR II) and U.S.
Supplement, NEDE-24011-P-A-14-US, dated June 2000, to derive the SLMCPR
values for OCNGS, Cycle 19. The revised values were approved by the NRC
staff via Amendment No. 233, dated September 26, 2002. Subsequently,
the licensee recalculated these SLMCPR values using the methodology in
Topical Report NEDC-32694-P-A, ``Power Distribution Uncertainties for
Safety Limit MCPR Evaluation,'' and requested to revise these values
further by the December 16, 2002, application.
The analysis methodology incorporates cycle-specific parameters.
These calculations do not change the operating procedures of OCNGS and
have no effect on the probability of an accident initiating event or
transient. The basis of the SLMCPR is to ensure no mechanistic fuel
damage is calculated to occur if the limit is not violated. The new
SLMCPR values preserve the existing margin to transition boiling and
[[Page 2800]]
the probability of fuel damage is not increased (i.e., in the event of
an accident or transient, the amount of fuel damaged would not be
increased as a result of the new SLMCPR values). Furthermore, the
proposed new SLMCPR values do not lead to, nor do they arise as a
result of, plant design or procedural changes. 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?
The new SLMCPR values for OCNGS Cycle 19 core have been calculated
in accordance with the methods and procedures described in NRC-approved
topical reports. The proposed new SLMCPR values do not lead to, nor do
they arise as a result of, plant design or procedural changes. The
changes do not involve any new method for operating the facility and do
not involve any facility modifications. As a result, no new initiating
events or transients could develop from the proposed changes.
Therefore, the proposed TS 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?
The margin of safety as defined in OCNGS's licensing basis will
remain the same. The new, cycle-specific SLMCPR values are calculated
using NRC-approved methods and procedures that are in accordance with
the current fuel design and licensing criteria. The SLMCPR values will
remain high enough to ensure that greater than 99.9% of all fuel rods
in the core are expected to avoid transition boiling if the limits are
not violated, thereby preserving the fuel cladding integrity.
Therefore, the proposed TS changes do not involve a significant
reduction in a margin of safety.
Based on the above review, it appears that the three standards of
10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the requested amendment involves no significant hazards
consideration.
Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius,
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Richard J. Laufer.
Consumers Energy Company, Docket No. 50-155, Big Rock Point Nuclear
Plant, Charlevoix County, Michigan
Date of amendment request: November 20, 2002.
Description of amendment request: The amendment request reflects
organizational changes due to the transfer of the Palisades Plant from
Consumers Energy to Nuclear Management Company. The revision reduces
redundancy between the Defueled Technical Specifications (DTS) and the
Big Rock Point Quality Program Description for Nuclear Power Plants.
Other changes are being proposed to correct minor typographical,
grammatical, and spelling errors.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Because this proposed change involves only a change in reporting
relationships, and no accidents previously evaluated consider
administrative controls, the change does not involve a significant
increase in the probability or consequences of an accident
previously considered.
2. Will the proposed change create the possibility of a new or
different kind of accident [from any other accident] previously
evaluated?
The proposed change would result in moving requirements for
certain reporting relationships from the Defueled Technical
Specifications to the Consumers Energy Quality Program Description
for Nuclear Power Plants, Part 1--Big Rock Point Plant (CPC-2A).
Because the Topical Report, CPC-2A, requires prior NRC [U.S. Nuclear
Regulatory Commission] approval for any changes which would reduce
the level of commitment in that document, an equivalent level of NRC
oversight is applied to changes to CPC-2A as are applied to changes
to Chapter 6 (Administrative Controls) of the Defueled Technical
Specifications. Therefore, no changes in administrative controls
defined in CPC-2A that might create the possibility of a new or
different kind of accident previously evaluated would be permitted
by the proposed change.
3. Will the proposed change involve a significant reduction in a
margin of safety?
The proposed change stipulates that individuals who perform
audits, surveillances and independent safety reviews will report as
indicated in CPC-2A, which states that independent safety reviews
are performed by the Restoration Safety Review Committee (RSRC). The
proposed change involves no significant change in a margin of safety
because margins of safety (in the Defueled Technical Specifications)
are directly controlled by system design and operation in accordance
with Limiting Conditions of Operation, Surveillances and Design
Features specified in the Defueled Technical Specifications are
affected by this proposed change.
To the extent that design and operation of systems having safety
margins might be affected by independent oversight, the following is
offered as evidence that no significant reduction in margin of
safety will result from the proposed change:
[sbull] The Manager, Nuclear Performance Assessment Department
(NPAD) and the RSRC both report their findings directly to the
Senior Nuclear Officer; therefore there will be no change in the
ultimate reporting relationship.
[sbull] The membership of NPAD and RSRC consists of individuals
who are independent of the plant organization.
[sbull] Changes to the Topical Report, CPC-2A that would result
in a reduction in level of commitment in the Quality Program
Description require a review and approval process equivalent to
proposed changes to the administrative controls specified in the
Defueled Technical Specifications.
[sbull] The requirements for performing onsite and offsite
reviews and audits are specified in CPC-2A; the proposed change to
the DTS to place the reporting relationship for individuals
performing these audits and reviews eliminates redundancy between
the Defueled Technical Specifications and CPC-2A.
The NRC staff has reviewed the licensee's significant hazards
analysis and, based on this review, it appears that the three
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: David A. Mikelonis, Esquire, Consumers
Energy Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: Robert A. Gramm.
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit No. 3, New London County, Connecticut
Date of amendment request: December 11, 2002.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) related to N-1 loop operation.
Specifically, the proposed changes would eliminate N-1 loop operation
from particular sections of the TS and would make other changes that
are clarifying and/or administrative in nature. In addition, the TS
Bases would be revised to address the proposed changes.
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 2801]]
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes do not alter the way any structure, system,
or component functions and would not alter the way [in] which the
plant is operated. The proposed changes do not involve any physical
plant modifications. The proposed changes incorporate existing plant
operational restrictions into the facility Technical Specifications
and provide for the removal of information which is not applicable
to plant operation.
The proposed allowed outage times (i.e. the required action
times for Specification 3.4.1.5) are reasonable and consistent with
the existing technical specification outage times and consistent
with industry guidelines, thereby ensuring affected components are
restored in a timely manner. The proposed changes to surveillance
requirements are also consistent with existing surveillance
frequencies and focus the Technical Specifications on verifying
normal plant configurations are maintained. The design basis
accidents, including the uncontrolled rod withdrawal from
subcritical and boron dilution events, will remain the same
postulated events described in the Millstone Unit No. 3 Final Safety
Analysis Report (FSAR), and the consequences of these events will
not be affected.
Therefore, the proposed changes will not increase the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes do not alter the plant configuration (no
new or different type of equipment will be installed) or require any
new or unusual operator actions. The proposed changes do not alter
the way any structure, system, or component functions and do not
alter the manner in which the plant is operated. The proposed
changes do not introduce any new 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. Involve a significant reduction in a margin of safety.
The proposed changes will not reduce the margin of safety since
they have no impact on any accident analysis assumption. The
proposed changes do not decrease the scope of equipment currently
required to be OPERABLE or subject to surveillance testing, nor do
the proposed changes affect any instrument setpoints or equipment
safety functions. The effectiveness of Technical Specifications will
be maintained since the changes will not alter the operation of any
component or system, nor will the proposed changes affect any safety
limits or safety system settings which are credited in a facility
accident analysis. Therefore, there is no reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Section Chief: James W. Clifford.
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: December 4, 2002.
Description of amendment request: The amendments revise Technical
Specification 3.7.6 by changing the minimum combined inventory for
Emergency Feedwater from 72,000 gallons to 155,000 gallons and
eliminating the condensate storage tank as a source of this inventory.
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:
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:
No. This revision to Technical Specification (TS) 3.7.6 changes
the inventory requirements for the Upper Surge Tank (UST) and
hotwell. These components provide a suction source to the Emergency
Feedwater System (EFW). This increase in inventory from 72,000
gallons to 155,000 gallons increases the required available
inventory. This increase in inventory does not affect the
probability or consequences of any previously evaluated accident.
(2) Create the possibility of a new or different kind of
accident from any kind of accident previously evaluated:
No. This revision to the combined UST and hotwell inventory
increases the required amount of water available to the EFW system.
No new or different kind of accident is created by this change as
only the required inventory is revised.
(3) Involve a significant reduction in a margin of safety:
No. The increase in required UST and hotwell inventory does not
reduce the margin of safety. The increase provides the required
inventory to ensure that the EFW can provide a Reactor Coolant
System cooldown at a rate of 50[deg] F/hour to decay heat removal
entry conditions following a reactor trip.
Duke has concluded, based on the above, that there is no
significant hazards considerations involved in this amendment
request.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC 20005.
NRC Section Chief: John A. Nakoski.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: May 1, 2002, as supplemented December 4,
2002.
Description of amendment request: The proposed amendment would
extend the applicability of the current Pilgrim Nuclear Power Station
(Pilgrim) reactor pressure vessel pressure-temperature (P-T) curves
through the end of Operating Cycle (OC) 16. The current P-T curves were
approved for use in License Amendment 190, dated April 13, 2001, and
are limited to use through the end of OC 14. The proposed change would
delete the 20 and 32 Effective Full Power Year (EFPY) curves and
replace the wording of the title blocks to allow use through the end of
OC 16. The proposed amendment would change Pilgrim Technical
Specification 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 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 proposed change involves a request to extend the use of the
current reactor pressure vessel P-T curves for two additional OCs. The
P-T curves were generated in accordance with the fracture toughness
requirements of 10 CFR Part 50, Appendix G, and American Society of
Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),
Section XI, Appendix G and Regulatory Guide 1.99, Revision 2, Radiation
Embrittlement of Reactor Vessel Materials, and were established in
[[Page 2802]]
compliance with the methodology used to calculate and predict effects
of radiation on embrittlement of reactor pressure vessel beltline
materials. There are no physical changes to the plant or new modes of
operation being introduced by the proposed change. Further, the
proposed change does not involve a change to any activities or
equipment and is not assumed in the safety analysis to initiate any
accident sequence. The proposed change does not adversely affect the
integrity of the reactor coolant pressure boundary such that its
function in the containment of radioactive materials is affected.
Additionally, the proposed change will not create any failure mode not
bounded by previously evaluated accidents. Therefore, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The current P-T curves were generated in accordance with the
fracture toughness requirements of 10 CFR Part 50, Appendix G, and ASME
Code, Section XI, Appendix G, and were approved by the U.S. Nuclear
Regulatory Commission for use through OC 14. The proposed change would
extend use of the P-T curves for two additional OCs. No new modes of
operation are introduced by the proposed change. Plant operation in
compliance with the current P-T curves ensures conditions in which
brittle fracture of primary coolant pressure boundary materials is
avoided. Accidents involving a breach of the primary coolant pressure
boundary have previously been evaluated and no other types of accidents
associated with the proposed change have been identified. Therefore,
the proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The proposed curves were established in compliance with the
methodology used to calculate and predict effects of radiation on
embrittlement of reactor pressure vessel beltline materials and are
estimated for 48 effective full-power years. The current curves are
approved for use through the end of OC 14 ([sim]19 EFPYs) which
provides a conservatism factor of 1.7 between the actual EFPYs at the
end of OC 14 and the end-of-life curve (32 EFPY). The change would
extend the use of the proposed curves to the end of OC 16 ([sim]23
EFPYs) which provides a conservatism factor of approximately 2.0. The
actual EFPYs at the end of OC 16 is bounded by the 48 EFPYs estimated
for the current curves. Therefore, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts 02360-5599.
NRC Section Chief: James W. Clifford.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: December 10, 2002
Description of amendment request: The proposed Technical
Specification (TS) amendment request changes the diesel fuel
specification to a more current revision in TS 4.10.C. The change would
also makes administrative revisions to reflect generic position titles
in TS 6.0, correct page numbers and titles in the Table on Contents,
and delete the General Table of Contents. Bases pages were also revised
to reflect the fuel specification revision as well as to make
administrative changes to provide clarity and correct a misspelling.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The operation of the Vermont Yankee Nuclear Power Station
[VY] in accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
VY has determined that the probability of occurrence of a
previously evaluated accident is not increased because the proposed
changes do not impact any accident initiating conditions. The
proposed changes will have no significant impact on any safety
related structures, systems or components. Additionally, the
administrative changes do not affect any system operation or
function.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
VY has determined that the proposed changes do not involve any
physical alteration of plant equipment and do not change the method
by which any safety-related system performs its function. No new or
different types of equipment will be installed. The proposed changes
do not create any new accident initiators or involve an activity
that could be an initiator of an accident of a different type.
Therefore, the proposed changes will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
VY has determined that the proposed changes do not alter the
basic operation of process variables, systems, or components as
described in the safety analysis. No new equipment is introduced.
The proposed changes do not impact design margins of any system
to perform its intended safety functions. There is no physical or
operational change being made which would alter the sequence of
events, plant response, or margins in existing safety analyses. The
proposed changes result in no impact on analyzed accident event
precursors or effects. These proposed changes do not alter the
physical design of the plant. There is no change in methods of
operation.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment 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.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois; Docket
Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle
County, Illinois; Docket Nos. 50-352 and 50-353, Limerick Generating
Station, Units 1 and 2 Docket Nos. 50-277 and 50-278, Peach Bottom
Atomic Power Station, Units 2 and 3, York County, Pennsylvania; Docket
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and
2, Rock Island County, Illinois
Date of amendment request: November 27, 2002.
[[Page 2803]]
Description of amendment request: The proposed amendments 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 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 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 November 27, 2002.
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 requests
involve no significant hazards consideration.
Attorneys for licensees: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chiefs: Anthony J. Mendiola, James W. Clifford.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: October 30, 2002.
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).
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 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 Three Mile Island (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
[[Page 2804]]
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 23 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, the Emergency Operating Procedures (at PNPP, these
procedures are titled the Plant Emergency Instructions), and site
survey monitoring that support modification of Emergency Plan
Protective Action Recommendations (PARs). Therefore, the elimination
of PASS requirements from Technical Specifications does not involve
a significant increase in the consequences of any accident
previously evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any previously evaluated.
The 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.
3. The proposed change will 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 do
not rely on PASS are designed to provide rapid assessment of current
reactor core conditions and the trending 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.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: March 22, 2002, as supplemented May 13,
June 24, July 29, and December 20, 2002.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) Surveillance Requirement (SR)
4.0.3 to extend the delay period, before entering a Limiting Condition
for Operation (LCO), following a missed surveillance. The delay period
would be extended from the current limit of ``* * * up to 24 hours to
permit the completion of the surveillance when the allowable outage
time limits of the ACTION requirements are less than 24 hours'' to ``*
* * up to 24 hours or up to the limit of the specified Frequency,
whichever is greater.'' In addition, the following requirement would be
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 NRC staff issued a notice of opportunity for comment in the
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments
concerning missed surveillances, including a model safety evaluation
and model no significant hazards consideration (NSHC) determination,
using the consolidated line item improvement process (CLIIP). The NRC
staff subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on September 28, 2001 (66 FR 49714). The licensee affirmed the
applicability of the following NSHC determination in its application
dated March 22, 2002, as supplemented May 13, June 24, July 29, and
December 20, 2002.
The proposed amendment would also add a requirement for a TS Bases
Control Program to the administrative controls section of TSs. This
change is necessary to be consistent with the CLIIP and is also
consistent with the TS Bases Control Program presented in Section 5.5
of NUREG-1431, Revision 2, ``Standard Technical Specifications
Westinghouse Plants.'' The licensee provided its analysis of the issue
of NSHC for this proposed change in its application.
The proposed amendment would also modify SR 4.0.1, and its
associated Bases, to link it with SR 4.0.3. The modification to SR
4.0.1 is consistent with NUREG-1431, Revision 2, ``Standard Technical
Specifications Westinghouse Plants.'' The licensee provided its
analysis of the issue of NSHC for this proposed change in its
application.
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.
[CLIIP Change]
The proposed change relaxes the time allowed to perform a missed
surveillance. The time between surveillances is not an initiator of
any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The
equipment being tested is still required to be operable and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected. Any reduction
in confidence that a standby system might fail to perform its safety
function due to a missed surveillance is small and would not, in the
absence of other unrelated failures, lead to an increase in
consequences beyond those estimated by existing analyses. The
addition of a requirement to assess and manage the risk introduced
by the missed surveillance will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or
[[Page 2805]]
consequences of an accident previously evaluated.
[Addition of TS Bases Control Program and Changes to SR 4.0.1]
The proposed changes to adopt the ITS [Improved Standard Technical
Specifications] wording for Specification 4.0.1 and formally adopt a
[TS] Bases Control Program are administrative in nature and do not
adversely affect accident initiators or precursors nor alter the
design assumptions, conditions, configuration of the facility or the
manner in which it is operated. The proposed changes do not alter or
prevent the ability or structures, systems, or components to perform
their intended function to mitigate the consequences of an
initiating event within the acceptance limits assumed in the
Seabrook Station Updated Final Safety Analysis Report (UFSAR).
Future changes to the TS Bases will continue to be
administratively controlled pursuant to the provisions of 10 CFR
50.59. The TS Bases is a licensee-controlled document that contains
bases information for the [TS]. Future changes to the information
contained in the TS Bases will be reviewed and approved in
accordance with the FPLE Seabrook Regulatory Compliance Manual and
TS Section 6.7.6j (TS Bases Control Program) of the Seabrook Station
[TS]. Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident From Any Previously
Evaluated.
[CLIIP Change]
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. A
missed surveillance will not, in and of itself, introduce new
failure modes or effects and any increased chance that a standby
system might fail to perform its safety function due to a missed
surveillance would not, in the absence of other unrelated failures,
lead to an accident beyond those previously evaluated. The addition
of a requirement to assess and manage the risk introduced by the
missed surveillance will further minimize possible concerns. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
[Addition of TS Bases Control Program and Changes to SR 4.0.1]
The proposed changes do not alter the design assumptions,
conditions, or configuration of the facility or the manner in which
the plant is operated. There are no changes to the source term or
radiological release assumptions used in evaluating the radiological
consequences in the Seabrook Station UFSAR. The proposed changes
have no adverse impact on component or system interactions. The
proposed changes will not adversely degrade the ability of systems,
structures and components important to safety to perform their
safety function nor change the response of any system, structure or
component important to safety as described in the UFSAR. The
proposed changes are administrative in nature and do not change the
level of programmatic and procedural details of assuring operation
of the facility in a safe manner. Since there are no changes to the
design assumptions, conditions, configuration of the facility, or
the manner in which the plant is operated and surveilled, the
proposed changes do not create the possibility of a new or different
kind of accident from any previously analyzed.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
[CLIIP Change]
The extended time allowed to perform a missed surveillance does
not result in a significant reduction in the margin of safety. As
supported by the historical data, the likely outcome of any
surveillance is verification that the LCO is met. Failure to perform
a surveillance within the prescribed frequency does not cause
equipment to become inoperable. The only effect of the additional
time allowed to perform a missed surveillance on the margin of
safety is the extension of the time until inoperable equipment is
discovered to be inoperable by the missed surveillance. However,
given the rare occurrence of inoperable equipment, and the rare
occurrence of a missed surveillance, a missed surveillance on
inoperable equipment would be very unlikely. This must be balanced
against the real risk of manipulating the plant equipment or
condition to perform the missed surveillance. In addition, parallel
trains and alternate equipment are typically available to perform
the safety function of the equipment not tested. Thus, there is
confidence that the equipment can perform its assumed safety
function.
Therefore, this change does not involve a significant reduction
in a margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
[Addition of TS Bases Control Program and Changes to SR 4.0.1]
There is no adverse impact on equipment design or operation and
there are no changes being made to the [TS] required safety limits
or safety system settings that would adversely affect plant safety.
The proposed changes are administrative in nature and do not reduce
the level of programmatic or procedural controls associated with the
activities presently performed via the aforementioned surveillance
requirements.
Future changes to the TS Bases information will be reviewed and
approved in accordance with Seabrook Station [TS], Section 6.7, and
as outlined in [FPLE Seabrook's] Regulatory Compliance programs.
Specifically, changes to the Seabrook Station [TS] Bases require an
evaluation pursuant to the provisions of 10 CFR 50.59 and review and
approval by the Station Operation Review Committee (SORC) prior to
implementation.
Therefore, formal adoption of a TS-required TS Bases Control
Program and adoption of ITS wording for Specification 4.0.1 do not
involve a significant reduction in the margin of safety provided in
the existing specifications.
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 NSHC.
Attorney for licensee: M. S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Section Chief: James W. Clifford.
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit 2, Berrien County, Michigan
Date of amendment request: November 15, 2002.
Description of amendment request: The proposed amendment would
revise the Donald C. Cook Nuclear Plant, Unit 2, operating license and
Technical Specifications to increase the licensed power level to 3468
Mega Watts Thermal (MWt), or 1.66 percent greater than the current
level of 3411 MWt.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
Response: No.
Probability of Occurrence of an Accident Previously Evaluated--
In support of this Measurement Uncertainty Recapture (MUR) power
uprate, a comprehensive evaluation was performed for nuclear steam
supply system (NSSS) and balance of plant systems and components and
analyses that could be affected by this change. A power calorimetric
uncertainty calculation was performed, and the effect of increasing
plant power by 1.66 percent on the plant's design and licensing
basis was evaluated. The result of these evaluations is that all
plant components will continue to be capable of performing their
design function at an uprated core power of 3468 MWt. In addition,
an evaluation of the accident analyses demonstrates that applicable
analysis acceptance criteria continue to be met. No accident
initiators are affected by this uprate and no challenges to any
plant safety barriers are created by this change.
Consequences of an Accident Previously Evaluated--This change
does not affect the release paths, the frequency of release, or the
source term for release for any accidents previously evaluated in
the Updated Final Safety Analysis Report. Structures, systems, and
components (SSC) required to mitigate transients remain capable of
performing their design functions, and thus were found acceptable.
The reduced uncertainty in the feedwater flow input to the power
[[Page 2806]]
calorimetric measurement ensures that applicable accident analyses
acceptance criteria continue to be met, to support operation at a
core power of 3468 MWt. Analyses performed to assess the effects of
mass and energy remain valid. The source terms used to assess
radiological consequences have been reviewed and determined to bound
operation at the uprated condition.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or single
failures are introduced as a result of the proposed changes. The
installation of the Caldon Leading Edge Flow Meter
CheckPlusTM system has been analyzed, and failures of
this system will have no adverse effect on any safety-related system
or any SSCs required for transient mitigation. SSCs previously
required for the mitigation of a transient remain capable of
fulfilling their intended design functions. The proposed changes
have no adverse effects on any safety-related system or component
and do not challenge the performance or integrity of any safety-
related system.
This change does not adversely affect any current system
interfaces or create any new interfaces that could result in an
accident or malfunction of a different kind than previously
evaluated. Operating at a core power level of 3468 MWt does not
create any new accident initiators or precursors. The reduced
uncertainty in the feedwater flow input to the power calorimetric
measurement ensures that applicable accident analyses acceptance
criteria continue to be met, to support operation at a core power of
3468 MWt. Credible malfunctions continue to be bounded by the
current accident analysis of record or evaluations that demonstrate
that applicable acceptance criteria continue to be met.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margins of safety associated with this MUR Uprate Program
are those pertaining to core power. This includes those associated
with the fuel cladding, Reactor Coolant System pressure boundary,
and containment barriers. A comprehensive engineering review was
performed to evaluate the 1.66 percent increase in the licensed core
power from 3411 MWt to 3468 MWt. The 1.66 percent increase required
that revised NSSS design thermal and hydraulic parameters be
established, which then served as the basis for all of the NSSS
analyses and evaluations. This engineering review concluded that no
design transient modifications are required to accommodate the
revised NSSS design conditions. NSSS systems and components were
evaluated and it was concluded that the NSSS equipment has
sufficient margin to accommodate the 1.66 percent power uprate. NSSS
accident analyses were evaluated for the 1.66 percent power uprate.
In all cases, the evaluations demonstrate that the applicable
analyses acceptance criteria continue to be met. As such, the
margins of safety continue to be bounded by the current analyses of
record for this change.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
In summary, based upon the above evaluation, [Indiana Michigan
Power Company] has concluded that the proposed amendment involves no
significant hazards consideration under the standards set forth in
10 CFR 50.92(c), and, accordingly, a finding of ``no significant
hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: L. Raghavan.
Nine Mile Point Nuclear Station, LLC (NMPNS), Docket No. 50-220, Nine
Mile Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: December 19, 2002.
Description of amendment request: The proposed amendment would
update and clarify the Technical Specifications (TSs) requirements for
demonstrating shutdown margin (SDM). The proposed changes incorporate
new, more restrictive, SDM limits; add the required limiting condition
for operation (LCO) actions if the SDM is not met; and also add the
surveillance requirements for verifying the SDM. These LCO actions and
surveillance requirements are not currently specified in the TSs. The
revised SDM limits account for the uncertainty in the demonstration of
adequate SDM analytically or by measurement. The proposed changes also
eliminate the unnecessary restriction requiring SDM demonstration in
the cold shutdown condition. The option for SDM demonstration in the
cold shutdown condition is retained consistent with the existing
special test exception.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Adequate SDM provides assurance that inadvertent criticalit[y]
and potential control rod drop accidents (CRDAs) involving high
worth control rods will not cause significant fuel damage. The SDM
is not an accident initiator and, as such, will have no effect on
the probability of an accident. The proposed changes incorporate
more restrictive SDM limits and provide the necessary actions and
verifications to assure that there will be no adverse effect on the
initial conditions and assumptions of the accidents previously
evaluated in the Updated Final Safety Analysis Report (UFSAR). 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.
The proposed changes to the SDM limits and requirements 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. Moreover, the proposed changes
will have no impact on conformance to 10 CFR [Code of Federal
Regulations] 50, Appendix A, General Design Criterion 26 (GDC 26),
in that the control rods will continue to satisfy the SDM
requirements and provide assurance that the reactor can be made
subcritical from all applicable operating conditions, transients,
and design basis events. 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 provide separate SDM limits for testing
consistent with the Improved Standard Technical Specifications
(NUREG-1433 and NUREG-1434) where the highest worth control rod is
determined analytically (0.38% Dk/k) or by measurement (0.28% Dk/k).
The proposed SDM limits are more restrictive than the current limit
(0.25% Dk/k) and account for the uncertainty in the demonstration of
SDM by testing. The SDM will continue to account for changes in core
reactivity during the fuel cycle. Therefore, the margin of safety is
increased relative to the SDM assumptions for the control rod
withdrawal error transient and CRDA analyses.
[[Page 2807]]
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
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Richard J. Laufer.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of amendment request: December 19, 2002.
Description of amendment request: The proposed amendment would
revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications
(TS) reporting requirements for the discovery of defective or degraded
steam generator tubes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not have any effect on structures,
systems, and components (SSCs) of the Kewaunee Nuclear Power Plant.
The changes do not affect plant operations, any design function or
an analysis that verifies the capability of an SSC to perform a
design function. The changes do not change any previously evaluated
accidents in the updated safety analysis report (UFSAR). As these
changes are administrative, there is no increase in the probability
and consequences of analyzed accidents.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes are administrative and do not change the
design function or operation of any plant SSCs. The proposed changes
do not create the possibility of a new or different kind of accident
due to credible new failure mechanisms, malfunctions, or accident
initiators not considered in the design and licensing bases.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes modify NRC reporting requirements only. The
changes do not exceed or alter a design basis or safety limit or
significantly reduce the margin of safety.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
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., Esq., Shaw Pittman,
Potts & Trowbridge, 2300 N. Street, NW, Washington, DC 20037-1128.
NRC Section Chief: L. Raghavan.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch Nuclear
Plant, Unit 2, Appling County, Georgia
Date of amendment request: December 4, 2002.
Description of amendment request: The proposed amendment changes
the Hatch Unit 2 turbine building high temperature primary containment
isolation value specified in Technical Specification Table 3.3.6.1-1,
Item 1f.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. [Does] the [* * *] proposed [* * *] change involve a
significant increase in the probability or consequences of an
accident previously evaluated[?]
This TS [Technical Specification] revision request changes the
allowable value for the turbine building high temperature primary
containment isolation. The setpoint at which the isolation occurs
has nothing to do with preventing a system break; therefore, this
proposed change will not change the probability of occurrence of a
small primary coolant system break.
For the turbine building high temperature primary containment
isolation, the analytical limit has been calculated at 207[deg]F
with the allowable value at 200[deg]F. The calculation supporting
these values accounts for instrument uncertainties thus confirming
that adequate margin exists between the allowable value and the
analytical limit. Accordingly, the consequences of a small primary
system break are not significantly increased.
2. [Does] the [* * *] proposed [* * *] change create the
possibility of a new or different kind of accident from any
previously evaluated[?]
Changing an allowable value does not introduce any new operating
modes for any plant system or piece of equipment. All plant systems
will continue to be operated, tested and maintained as before, and
within their licensing and design basis. As a result, no new failure
modes are introduced and the possibility of a new or different type
of accident is not created.
3. [Does] the [* * *] proposed [* * *] change involve a
significant decrease in the margin of safety[?]
Increasing the allowable value by 6[deg]F does not result in a
significant reduction in a margin of safety. A formal calculation
was performed which justified an analytical limit of 207[deg]F. This
calculation determined the analytical limit based on a primary leak
into the turbine building and confirmed that the allowable value
adequately protects the analytical limit. As a result, the margin of
safety is not significantly reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: John A. Nakoski.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these
[[Page 2808]]
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, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management Systems (ADAMS) Public Electronic
Reading Room on the internet at the NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by email to [email protected].
Dominion Nuclear Connecticut, Inc. Docket Nos. 50-336 and 50-423,
Millstone Power Station, Unit Nos. 2 and 3, New London County,
Connecticut
Date of application for amendment: February 14, 2002, as
supplemented on September 9, 2002.
Brief description of amendment: The amendments revised the
Millstone Power Station, Unit No. 2 (MP2) and 3 (MP3) Technical
Specifications (TSs) by relocating selected MP2 and MP3 TSs related to
the Reactor Coolant System and Plant Systems to the respective unit's
Technical Requirements Manual.
The amendment does not address changes to MP2 TS 3/4.7.10,
``Snubbers,'' and MP3 TSs 3/4.7.10, ``Snubbers,'' and 3/4.7.14, ``Area
Temperature Monitoring,'' as described by the application dated
February 14, 2002, because these proposed TSs changes were withdrawn by
the supplement dated September 9, 2002.
Date of issuance: January 2, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 272 and 214.
Facility Operating License Nos. DPR-65 and NPF-49: This amendment
revised the TSs.
Date of initial notice in Federal Register: April 16, 2002 (67
FR18645). The September 9, 2002, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination or expand the amendment beyond the scope of
the initial notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 2, 2003.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: September 12, 2002, as
supplemented by letter dated December 30, 2002.
Brief description of amendments: The amendments temporarily revised
Technical Specifications (TS) 3.5.2, ``Emergency Core Cooling System;''
TS 3.6.6, ``Containment Spray System;'' TS 3.7.5, ``Auxiliary Feedwater
System;'' TS 3.7.7, ``Component Cooling Water System;'' TS 3.7.8,
``Nuclear Service Water System;'' and TS 3.8.1, ``AC Sources.''
Date of issuance: January 7, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 203 & 196.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 15, 2002 (67 FR
63692). The supplement dated December 30, 2002, provided clarifying
information that did not change the scope of the September 12, 2002,
application, nor the initial no significant hazard consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 7, 2003.
No significant hazards consideration comments received: No
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: August 29, 2002. Brief
description of amendments: The amendments revise Technical
Specification (TS) 3.8.4.7, to modify the note to eliminate the ``once
per 60 months'' restriction on replacing the battery service test by
the battery modified performance discharge test.
Date of issuance: January 9, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 204 & 197.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68733). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 9, 2003.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: October 22, 2002.
Brief description of amendment: The amendment deletes a reference
to Section 2.E in Section 2.F of Facility Operating License No. NPF-21.
Section 2.E requires the licensee to fully implement and maintain in
effect all provisions of the Commission-approved physical security,
guard training and qualification, and safeguards contingency plans.
Section 2.E is redundant because the reporting requirements and
criteria for the physical security programs are specified in 10 CFR
73.71 and Appendix G of 10 CFR Part 73.
Date of issuance: January 9, 2003.
Effective date: January 9, 2003 to be implemented within 60 days
from the date of issuance.
Amendment No.: 183.
Facility Operating License No. NPF-21: The amendment revised the
operating license.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75871). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 9, 2003.
No significant hazards consideration comments received: No.
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: August 21, 2002.
Brief description of amendment: The 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,
[[Page 2809]]
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: January 2, 2003.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 127.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 1, 2002 (67 FR
61679). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 2, 2003.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request October 15, 2001, as supplemented by
letter dated August 27, 2002.
Brief description of amendment: The amendment provides additional
information to support a modification to Technical Specification 3.4.7
and limits Reactor Coolant System activity permitted by the ACTION
statement to 60 microcuries per gram at all power levels. The letdown
line break accident analysis in the Final Safety Analysis Report was
also changed.
Date of issuance: January 8, 2003.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 184.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications and Final Safety Analysis Report.
Date of initial notice in Federal Register: October 28, 2002 (67 FR
66009). The August 27, 2002, supplemental letter provided additional
information and revised the no significant hazards consideration
determination. The original Federal Register notice was published on
November 28, 2001 (66 FR 56504), but was superceded by the October 28,
2002 publication.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 8, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: June 26, 2002, as supplemented
September 12, 2002.
Brief description of amendments: Extend the use of the pressure-
temperature limits in Technical Specification Figure 3.4.6.1-1 to 32
effective full power years.
Date of issuance: As of date of issuance and shall be implemented
within 30 days.
Effective date: January 2, 2003.
Amendment Nos. 163 and 125.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 6, 2002 (67 FR
50953). The supplement dated September 12, 2002, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 2, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-171, Peach Bottom Atomic
Power Station, Unit 1, York County, Pennsylvania
Date of application for amendment: May 21, 2002.
Brief description of amendment: This proposed amendment will revise
the Peach Bottom Atomic Power Station, Unit 1, License and Technical
Specifications (TS) to: (1) Delete License Condition C(4) to reflect
satisfaction of the minimum decommissioning trust fund amount at the
time of transfer of the Facility Operating License; (2) revise License
Condition C(5)(d) to reflect 30 days prior written notification to the
Director of Nuclear Material Safety and Safeguards before modification
of the decommissioning trust agreement in any material respect; (3)
delete TS 2.1(B)3 and TS 2.4(b) to eliminate inconsistencies with
reporting requirements in 10 CFR 20.2202, 50.73, and 73.71; (4) revise
TS 2.2 to refer to the Facility Operating License; and (5) revise TS
2.3 to refer to the radiological hazards associated with the facility.
Date of Issuance: December 26, 2002.
Effective Date: On the date of issuance of this amendment and must
be fully implemented no later than 30 days from the date of issuance.
Amendment No.: 11.
Facility Operating License No. DPR-12: Amendment revised the
License and TS with respect to administrative procedures or
requirements.
Date of initial notice in Federal Register: October 1, 2002 (67 FR
61682). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 26, 2002.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendments: June 1, 2001, as supplemented
by letters dated June 13, 2001, May 20, 2002, and June 28, 2002.
Brief description of amendments: These amendments revised TS 3.7.1,
``Residual Heat Removal Service Water (RHRSW) System and Ultimate Heat
Sink (UHS),'' to add operability requirements and surveillance
requirements for the UHS spray bypass and large array valves, and
reduce the allowed Completion Times for the conditions applicable to
the RHRSW system.
Date of issuance: December 30, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 206 and 180.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 5, 2001 (66
FR 46481). The June 13, 2001, May 20, 2002, and June 28, 2002, letters
provided additional information that clarified the application, but did
not expand the scope of the application as originally noticed, and did
not change the staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 30, 2002.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: December 3, 2001, as supplemented by
letter dated August 29, 2002.
Brief description of amendments: The amendments revise TS 3.7.1.2,
[[Page 2810]]
``Auxiliary Feedwater System,'' to better reflect the four train
auxiliary feedwater (AFW) system design at STP. Specifically, the
changes specify the same allowed outage time (AOT) for any one
inoperable motor-driven pump, regardless of train. The amendments also
extend the AOT for one inoperable motor-driven pump from 72 hours to 28
days. A sentence has also been added to Action d. stating that Limiting
Condition for Operation (LCO) 3.0.3 and all other LCO actions requiring
Mode changes are suspended until one of the four inoperable AFW pumps
is restored to operable status. There is also an administrative change
in the wording of the LCO to clarify that there are only four AFW pumps
in each STP unit.
Date of issuance: December 31, 2002.
Effective date: December 31, 2002.
Amendment Nos.: Unit 1--146; Unit 2--134.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 22, 2002 (67 FR
2930). The supplement provided additional information that clarified
the application, did not expand the scope as originally noticed, and
did not change the staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 31, 2002.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: April 8, 2002.
Brief description of amendments: The amendments revise Technical
Specification (TS) 3.4.16, ``RCS [Reactor Coolant System] Specific
Activity,'' to lower the Limiting Condition For Operation and
associated Surveillance Requirements for Dose Equivalent Iodine-131 in
the RCS from a specific activity of 1.0 [mu]Ci/gm to 0.45 [mu]Ci/gm.
Date of issuance: January 6, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 102 and 102.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 11, 2002 (67 FR
40026). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 6, 2003.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket No. 50-338, North Anna
Power Station, Unit 1, Louisa County, Virginia
Date of application for amendment: December 7, 2001, as
supplemented by letters dated June 28 and July 25, 2002.
Brief description of amendment: This amendment permits a one-time
extension of the current 10-year Title 10 of the Code of Federal
Regulations Part 50, Appendix J, Option B, Type A test interval from
April 3, 2003, to April 2, 2008.
Date of issuance: December 31, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 234.
Facility Operating License No. NPF-4: Amendment changes the
Technical Specifications.
Date of initial notice in Federal Register: April 30, 2002 (67 FR
21295). The supplemental letters dated June 28 and July 25, 2002,
contained clarifying information only and did not change the proposed
no significant hazards consideration determination or expand the scope
of the initial application.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 31, 2002.
No significant hazards consideration comments received: No.
For the Nuclear Regulatory Commission.
Dated at Rockville, Maryland, this 13th day of January 2003.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 03-1161 Filed 1-17-03; 8:45 am]
BILLING CODE 7590-01-P