[Federal Register Volume 68, Number 218 (Wednesday, November 12, 2003)]
[Notices]
[Pages 64133-64145]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-28065]
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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 October 17, 2003, through October 30, 2003.
The last biweekly notice was published on October 28, 2003 (68 FR
59212).
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
[[Page 64134]]
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By December 12, 2003, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714, which is
available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records will be accessible from the
Agencywide Documents Access and Management System's (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing
or petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of continuing disruptions in delivery of mail to United States
Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the request for
hearing and petition for leave to intervene should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and because of continuing disruptions in
delivery of mail to United States Government offices, it is requested
that copies be transmitted
[[Page 64135]]
either by means of facsimile transmission to 301-415-3725 or by e-mail
to [email protected]. A copy of the request for hearing and
petition for leave to intervene should also be sent to the attorney for
the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to
[email protected].
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: July 1, 2003.
Description of amendment request: The proposed amendments would
revise Appendix A, Technical Specifications (TS), of Facility Operating
License Nos. NPF-11 and NPF-18. Specifically, the proposed changes will
delete one and add two references to the list of analytical methods in
TS 5.6.5, ``Core Operating Limits Report (COLR),'' that can be used to
determine core operating limits. The deleted reference is to an
analytical method that is no longer applicable to LaSalle County
Station (LSCS). The new references will allow LSCS to use General
Electric Company (GE) methods for the determination of fuel assembly
critical power of Framatome Advanced Nuclear Fuel, Inc. (Framatome)
Atrium-9B and Atrium-10 fuel. The proposed changes are the result of a
LSCS decision to insert GE14 fuel during the upcoming refueling outage
at LSCS Unit 1 in January 2004. GE's safety analysis methodologies have
been previously used at LSCS and GE14 fuel is currently in use at other
Exelon Generation Company, LLC (Exelon), stations.
The first added reference, ``GEXL96 Correlation for Atrium-9B
Fuel,'' will list a method that was previously approved by the NRC for
use by licensees. The second added reference, ``GEXL97 Correlation for
Atrium-10 Fuel,'' will list a GE method for determining the critical
power for Atrium-10 fuel. This correlation has not been previously
reviewed and approved by the NRC for use by licensees. Additionally,
editorial changes will be made to existing references.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes will delete one and add two additional
references to the list of administratively controlled analytical
methods in TS 5.6.5, ``Core Operating Limits Report (COLR),'' that
can be used to determine core operating limits and make minor
editorial changes to the existing references. TS 5.6.5 lists NRC
approved analytical methods used at LaSalle County Station (LSCS) to
determine core operating limits. [LSCS Unit 1 is scheduled to load
GE fuel during its upcoming outage in January 2004.]
The proposed changes to TS Section 5.6.5 will add the fuel
analytical methods that support the initial insertion of GE14 fuel
to the list of methods used to determine the core operating limits.
The deletion or addition of approved methods to TS Section 5.6.5 and
minor editorial changes to the existing references has no effect on
any accident initiator or precursor previously evaluated and does
not change the manner in which the core is operated. The methods
have been reviewed to ensure that the output accurately models
predicted core behavior, have no effect on the type or amount of
radiation released, and have no effect on predicted offsite doses in
the event of an accident. Thus, the proposed changes do not have any
effect on the probability of an accident previously evaluated.
The proposed changes in the administratively controlled
analytical methods does [do] not affect the ability of LSCS to
successfully respond to previously evaluated accidents and does [do]
not affect radiological assumptions used in the evaluations. Thus,
the radiological consequences of any accident previously evaluated
are not increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes to TS Section 5.6.5 do not affect the
performance of any LSCS structure, system, or component credited
with mitigating any accident previously evaluated. The insertion of
a new generation of fuel which has been analyzed with NRC approved
methodologies will not affect the control parameters governing unit
operation or the response of plant equipment to transient
conditions. The proposed changes do not introduce any new modes of
system operation or failure mechanisms.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed changes will delete one and add two additional
references to the list of administratively controlled analytical
methods in TS 5.6.5 that can be used to determine core operating
limits and make minor editorial changes to the titles of existing
references. The proposed changes do not modify the safety limits or
setpoints at which protective actions are initiated, and do not
change the requirements governing operation or availability of
safety equipment assumed to operate to preserve the margin of
safety. Therefore, LSCS has determined that the proposed changes
provide an equivalent level of protection as that currently
provided.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of amendment request: September 11, 2003.
Description of amendment request: Revise the dose model for the
containment activated concrete, rebar (hereafter referred to as
activated concrete) and liner, by incorporating more realistic
radionuclide release rates and to change the associated derived
concentration guideline limit (DCGL) for activated concrete.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the
[[Page 64136]]
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 requested license amendment does not authorize any plant
activities beyond those allowed by 10 CFR Chapter I or beyond those
considered in the DSAR. The bounding accident described in the
Defueled Safety Analysis Report (DSAR) for potential airborne
activity is the postulated resin cask drop accident in the Low Level
Radioactive Waste Storage Building. This accident is expected to
contain more potential airborne activity than can be released from
other decommissioning events. The radionuclide distribution assumed
for the spent resin cask has a greater inventory of transuranic
radionuclides (the major dose contributor) than the distribution of
plant derived radionuclides in the components involved in other
decommissioning accidents. The other accidents considered in the
DSAR include: (1) Explosion of liquid petroleum gas (LPG) leaked
from a front end loader or forklift; (2) Explosion of oxyacetylene
during segmenting of the reactor vessel shell; (3) Release of
radioactivity from the RCS decontamination ion exchange resins; (4)
Gross leak during in-situ decontamination; (5) Segmentation of RCS
piping with unremoved contamination; (6) Fire involving contaminated
clothing or combustible waste; (7) Loss of local airborne
contamination control during blasting or jackhammer operations; (8)
Temporary Loss of Services; (9) Dropping of Contaminated Concrete
Rubble; (10) Natural phenomena; and (11) Transportation accidents.
The probabilities and consequences for these accidents are estimated
in the basis documentation for DSAR Section 7. No systems,
structures, or components that could initiate or be required to
mitigate the consequences of an accident are affected by the
proposed change in any way not previously evaluated in the DSAR.
Since Maine Yankee does not exceed the salient parameters associated
with the plant referenced in the basis documentation in any material
respects, it is concluded that these probabilities and consequences
are not increased. Therefore, the proposed change to the Maine
Yankee license does not involve any increase in the probability or
consequences of any 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 requested license amendment does not authorize any plant
activities that could precipitate or result in any accidents beyond
those considered in the DSAR. The accidents previously evaluated in
the DSAR are described above. These accidents are described in the
basis documentation for DSAR Section 7. The proposed change does not
affect plant systems, structures, or components in any way not
previously evaluated in the DSAR. Since Maine Yankee does not exceed
the salient parameters associated with the plant referenced in the
basis documentation in any material respects, it is concluded that
these accidents appropriately bound the kinds of accidents possible
during decommissioning. Therefore, the proposed change to the Maine
Yankee license would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The margin of safety defined in Maine Yankee's license basis for
the consequences of decommissioning accidents has been established
as the margin between the bounding decommissioning accident and the
dose limits associated with the need for emergency plan offsite
protection, namely the Environmental Protection Agency Protective
Action Guidelines EPA-PAGs. As described above, the bounding
decommissioning accident is the postulated resin cask drop accident
in the Low Level Radioactive Waste Storage Building. Since the
bounding decommissioning accident is expected to contain more
potential airborne activity than can be released from other
decommissioning events and since the radionuclide distribution
assumed for the spent resin cask has more transuranics (the major
dose contributor) than the distribution in the components involved
in other decommissioning accidents, the margin of safety associated
with the consequences of decommissioning accidents cannot be
reduced. The margin of safety defined in the statements of
consideration for the final rule on the Radiological Criteria for
License Termination is described as the margin between the 100 mrem/
yr public dose limit established in 10 CFR 20.1301 for licensed
operation and the 25 mrem/yr dose limit to the average member of the
critical group at a site considered acceptable for unrestricted use.
This margin of safety accounts for the potential effect of multiple
sources of radiation exposure to the critical group. Since the
license termination plan (LTP) was designed to comply with the
radiological criteria for license termination for unrestricted use,
the margin of safety cannot be reduced. Therefore, the proposed
changes to the Maine Yankee license would not involve a significant
reduction in any margin of safety.
Conclusion
Based on the above, Maine Yankee concludes that the proposed
amendment presents no significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendment involves no significant hazards consideration.
Attorney for licensee: Joe Fay, Esquire, Maine Yankee Atomic Power
Company, 321 Old Ferry Road, Wiscasset, Maine 04578.
NRC Section Chief: Claudia M. Craig.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: April 11, 2003.
Description of amendment request: The proposed amendment would make
various administrative, editorial, and typographical changes to
Technical Specification (TS) Section 5.0, ``Administrative Controls.''
Specifically, the proposed changes would:
(1) Correct TS 5.4.1.a by adding ``Appendix A'' after the reference
to ``Regulatory Guide 1.33, Revision 2,'' and deleting ``of'' before
this reference.
(2) Change TS 5.5.2.e by deleting the phrase ``(approximately 44
psig)'' which is an invalid reference to the normal hydrostatic head
from the safety injection refueling water tank for the test conditions
required for maximum allowable leakage from recirculation heat removal
systems' components.
(3) Make several editorial changes to TS 5.6.1 to be consistent
with the wording of NUREG-1432, ``Standard Technical Specifications-
Combustion Engineering Plants,'' Revision 2 (STS), and the changes to
the STS in Technical Specification Task Force (TSTF) Traveler TSTF-152.
The editorial changes include (a) adding the word ``collective'' to
describe the associated collective deep dose equivalent, (b) adding
``thermoluminescence dosimeter'' to define its acronym ``(TLD),'' (c)
changing ``stations'' to ``station,'' (d) adding the words ``received
from'' when describing the 80 percent of total deep dose equivalent
received from external sources, and (e) making punctuation 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
consideration, which is presented below:
The following evaluation supports the finding that operation of the
facility in accordance with the proposed change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed license amendment provides changes to Technical
Specification (TS) Administrative Controls sections 5.4.1.a,
5.5.2.e, and 5.6.1. The proposed corrections to TS 5.4.1.a are
editorial in nature. The proposed correction to TS 5.5.2.e, which
[[Page 64137]]
deletes an erroneous approximate value from the description of test
conditions for maximum allowable leakage from recirculation heat
removal system components, is consistent with the existing plant
design as described in the Palisades Final Safety Analysis Report.
The proposed correction to TS 5.6.1 is editorial in nature and is
consistent with the Nuclear Regulatory Commission approved standard
technical specifications. The proposed amendment does not involve
operation of the required structures, systems or components (SSCs)
in a manner or configuration different from those previously
recognized or evaluated.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed amendment does not involve a physical alteration of
any SSC or a change in the way any SSC is operated. The proposed
amendment does not involve operation of any required SSCs in a
manner or configuration different from those previously recognized
or evaluated. No new failure mechanisms will be introduced by the
changes being requested.
Thus, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed amendment does not affect any margin of safety. The
proposed amendment does not involve any physical changes to the
plant or manner in which the plant is operated.
Therefore, the proposed amendment would 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: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Section Chief: L. Raghavan.
Southern Nuclear Operating Company, Inc, Docket Nos. 50-348, Joseph M.
Farley Nuclear Plant, Unit 1, Houston County, Alabama
Date of amendment request: September 19, 2003.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TS) Limiting Conditions for Operation
(LCO) 3.8.4, ``DC Sources--Operating,'' for the remainder of operating
cycle 19. Specifically, the proposed TS change would increase the
Completion Time for the 1B Auxiliary Building DC electrical power
system inoperability due to an inoperable battery to allow for on-line
replacement of individual cells. Cycle 19 is presently scheduled to end
on October 2, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change to LCO 3.8.4 creates an extended Completion
Time for an inoperable 1B Auxiliary Building DC electrical power
subsystem due to an inoperable battery on Unit 1 only for the
remainder of operating cycle 19. The Auxiliary Building battery is
not a direct initiator of any analyzed accident sequence. The
radiological consequences of any associated accidents are not
impacted by the proposed amendment. Therefore, the proposed
amendment does not involve a significant increase in the probability
or consequences of an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change involves no change to the physical plant. It
allows additional time for corrective maintenance on the 1B
Auxiliary Building battery on Unit 1. The proposed amendment
involves an extension of a previously determined acceptable mode of
operation. The proposed amendment does not introduce any new
equipment, create new failure modes for existing equipment, or
create any new limiting single failures. Therefore, the proposed
amendment does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The physical plant is unaffected by these changes. The proposed
changes do not impact accident offsite dose, containment pressure or
temperature, emergency core cooling system (ECCS) or reactor
protection system (RPS) settings or any other parameter that could
affect a margin of safety. Under the proposed amendment, the unit
will continue to be operated in a condition that will ensure that
emergency power will be available as needed. The extended Completion
Time for an inoperable battery has been shown to have a very small
impact on plant risk using the criteria of Regulatory Guides 1.174,
An Approach for Using Probabilistic Risk Assessments in Risk-
Informed Decision-making and 1.177, An Approach for Plant-Specific.
Risk-Informed Decisionmaking: Technical Specifications and is
acceptable. Therefore, the proposed amendment does not involve a
significant reduction in a margin to safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Section Chief: John A. Nakoski.
Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364,
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama
Date of amendment request: August 29, 2003.
Description of amendment request: The proposed amendments would
revise Technical Specifications Limiting Condition of Operation 3.9.3,
``Containment Penetrations.'' The proposed changes would allow the
equipment hatch to be open during core alterations and/or during
movement of irradiated fuel assemblies within containment.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed changes will allow the equipment hatch to be
open during core alterations and movement of irradiated fuel
assemblies inside containment. The proposed changes will not alter
the manner in which fuel is handled or core alterations are
performed. The equipment hatch is not an initiator of any accident.
The status of the equipment hatch during refueling operations has no
effect on the probability of the occurrence of any accident
previously evaluated. The radiological consequences of a fuel
handling accident inside containment have been determined to be well
within the limits of 10 CFR 100 and they meet the acceptance
criteria of General Design Criterion (GDC) 19. Therefore the
proposed changes do not involve a significant increase in the
probability or consequences of [any] accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
No. The proposed changes do not create any new failure modes for
any system or component, nor do they adversely affect plant
operation. No new equipment will be added and no new limiting single
failures
[[Page 64138]]
will be created. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident [from any
accident] previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
No. The dose consequences were determined to be well within the
limits of 10 CFR 100 and they meet the acceptance criteria of GDC
19. Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Section Chief: John A. Nakoski.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: October 3, 2003.
Description of amendment request: The proposed amendments would add
a Limiting Condition for Operation (LCO) for the Linear Heat Generation
Rate. The new LCO will be included in Section 3.2, Power Distribution
Limits. The proposed amendments would also change the recirculation
loop LCO, Section 5.6.5, and the appropriate Bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed addition of LCO 3.2.3 and supporting Bases are
being made to support new modeling improvements in core monitoring.
This change is administrative in nature in that it does not involve,
require, or result from any physical change to the plant, including
the reactor core or its fuel. The addition of LCO 3.2.3 and Bases B
3.2.3 is consistent with Revision 2 of Volumes 1 and 2 of NUREG-
1433. Changes being proposed for Bases section B 3.2.1 and TS
Section 5.6.5 are simply supportive in nature to the relocation of
LHGR [linear heat generation rate] from the APLHGR [averageplant
linear heat generation rate] Section Bases B 3.2.1 to the new
section LHGR B 3.2.3.
Also, no changes are being proposed to any plant system,
structure, or component designed to prevent or mitigate the
consequences of a previously evaluated event.
Therefore, because the physical characteristics and performance
requirements of the plant systems, structures, and components
(including the reactor core and fuel) will not be altered, the
proposed license amendment does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
No plant systems, structures, or components (including the
reactor core and fuel) will be altered by the proposed change to the
LCO or supporting Bases.
Additionally, this TS [technical specification] change request
does not propose changes in the operation of any plant system.
Consequently, new and unanalyzed modes of operation are not
introduced.
As a result, the possibility of a new or different kind of
accident from any previously evaluated is not introduced.
3. The proposed change does not involve a significant reduction
in the margin of safety.
Previously, the LHGR was included in the monitoring of the
APLHGR. Now, SNC [Southern Nuclear Company] proposes to monitor LHGR
on its own while continuing to monitor APLHGR. This proposed TS
change adds an LCO for LHGR and a corresponding requirement for the
COLR [core operating limits report].
The margin of safety is not reduced since the LHGR and APLHGR
will continue to be monitored.
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.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: May 13, 2003.
Description of amendment request: The proposed license amendment
would allow use of a revised methodology, for performance of certain
accident analyses, described in Westinghouse Electric Corp. (W) report
WCAP-14882-S1-P, Revision 0 (Proprietary), ``RETRAN-02, Modeling and
Qualification for Westinghouse Pressurized Water Reactors Non-LOCA
Safety Analyses, Supplement 1--Thick Metal Mass Heat Transfer Model and
NOTRUMP-Based Steam Generator Mass Calculation Method,'' dated December
2002.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No. The proposed methodology uses more realistic
computer models with unnecessary conservatism removed. The
methodology used to analyze the consequences of a postulated
accident is not an initiator that can affect the probability or
consequences of that accident. The change does not alter assumptions
previously made in the radiological consequences of the accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No. The proposed methodology uses more realistic
computer models with unnecessary conservatism removed. The
methodology used to analyze the consequences of a postulated
accident is not an initiator that can cause an accident to occur.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The proposed methodology uses more realistic
computer models with unnecessary conservatism removed. Using the
methodology of WCAP-14882-S1-P results in additional margin to
pressurizer overfill for a postulated loss of normal feedwater/ loss
of offsite power at STP. Therefore, the proposed changes do not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: Robert A. Gramm.
[[Page 64139]]
STP Nuclear Operating Company (STPNOC), Docket Nos. 50-498 and 50-499,
South Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: September 22, 2003.
Description of amendment request: The proposed amendments would
change the requirements for the Engineered Safety Feature sequencer,
and the Surveillance Requirements that are applicable in Mode 5 and 6
to provide needed clarification. In addition, the proposed amendment
would correct a typographical error in that requirement ``c.'' in
Technical Specification 3.2.4 should actually be requirement ``b.''.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not change the plant design basis,
system configuration or operation, and do not add or affect any
accident initiator.
Therefore, STPNOC concludes that there is no 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 do not change the plant design basis,
system configuration or operation, and do not add or affect any
accident initiator.
Therefore, STPNOC concludes the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
No actual plant equipment or accident analyses will be affected
by the proposed change. Additionally, the proposed changes will not
relax any criteria used to establish safety limits, will not relax
any safety systems settings, or will not relax the bases for any
limiting conditions of operation. Therefore, STPNOC concludes the
proposed changes do not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. With regard to the licensee's proposed correction of a
typographical error in TS 3.2.4, the NRC staff notes the following:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Correction of a typographical error does not change the plant
design basis, system configuration or operation, and does not add or
affect any accident initiator. Therefore, there is no 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.
Correction of a typographical error does not change the plant
design basis, system configuration or operation, and does not add or
affect any accident initiator. Therefore, the proposed change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
No actual plant equipment or accident analyses will be affected
by the proposed change. Additionally, the proposed changes will not
relax any criteria used to establish safety limits, will not relax
any safety systems settings, or will not relax the bases for any
limiting conditions of operation. Therefore, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the above, the NRC staff concludes that the standards of
10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the request for amendments involves no significant
hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: Robert A. Gramm.
STP Nuclear Operating Company (STPNOC), Docket Nos. 50-498 and 50-499,
South Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: September 22, 2003.
Description of amendment request: The proposed amendments would
change the Technical Specification 3.3.2 requirements for Loss of Power
Instrumentation (Functional Unit 8).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not change the plant design basis,
system configuration or operation, and do not add or affect any
accident initiator.
Therefore, STPNOC concludes that there is no 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 do not change the plant design basis,
system configuration or operation, and do not add or affect any
accident initiator.
Therefore, STPNOC concludes the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
No actual plant equipment or accident analyses will be affected
by the proposed change. Additionally, the proposed changes will not
relax any criteria used to establish safety limits, will not relax
any safety systems settings, or will not relax the bases for any
limiting conditions of operation. Therefore, STPNOC concludes the
proposed changes do not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: Robert A. Gramm.
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: October 16, 2003.
Description of amendment request: The proposed amendments request a
one-time change to Technical Specification (TS) 4.4.5.3a to extend the
40-month steam generator inspection interval to 44 months for Unit 1
only.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not alter the plant design. The scope
of inspections
[[Page 64140]]
performed during 1RE10 [Refueling Outage 10 for Unit 1], the first
refueling outage following SG [steam generator] replacement,
exceeded the TS requirements for the first two refueling outages
after replacement combined. That is, more tubes were inspected than
were required by TS. Currently, South Texas Project Unit 1 does not
have an active SG damage mechanism and will meet the current
industry examination guidelines without performing inspections
during the next refueling outage. The results of the Condition
Monitoring Assessment after 1RE10 demonstrated that all performance
criteria were met during 1RE10. The results of the 1RE10 Operational
Assessment show that all performance criteria will be met over the
proposed operating period.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not alter any plant design basis or
postulated accident resulting from potential SG tube degradation.
The scope of inspections performed during 1RE10, the first refueling
outage following SG replacement, significantly exceeded the TS
requirements for the scope of the first two refueling outages after
SG replacement combined.
The proposed change does not affect the design of the SGs, the
method of operation, or reactor coolant chemistry controls. No new
equipment is being introduced and installed equipment is not being
operated in a new or different manner. The proposed change involves
a one-time extension to the SG tube inservice inspection interval,
and therefore will not give rise to new failure modes. In addition,
the proposed change does not impact any other plant system or
components.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Steam generator tube integrity is a function of design,
environment, and current physical condition. Extending the SG tube
inservice inspection frequency [interval] by four months does not
alter the function or design of the SGs. Inspections conducted prior
to placing the SGs into service (preservice inspections) and
inspection during the first refueling outage following SG
replacement demonstrate that the SGs do not have fabrication damage
or an active damage mechanism. The scope of those inspections
significantly exceeded those required by the TS. These inspection
results were comparable to similar inspection results for the same
model of RSGs [replacement steam generators] installed at other
plants, and subsequent inspections at those plants yielded results
that support this extension request. The improved design of the
replacement SGs also provides reasonable assurance that significant
tube degradation is not likely to occur over the proposed operating
period.
Therefore, the proposed 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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: Robert A. Gramm.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: September 23, 2003.
Brief description of amendments: The proposed change would revise
Technical Specification (TS) 3.6.3 entitled, ``Containment Isolation
Valves,'' to extend the frequency of Surveillance Requirement 3.6.3.7
for containment and hydrogen purge valves and containment pressure
relief valves with resilient seats.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR), Section 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration, which is
presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
Operability and leakage control effectiveness of the containment
purge, hydrogen purge and containment pressure relief system
isolation valves have no effect on whether or not an accident
occurs. Consequently, increasing the interval between surveillances
of isolation valve leakrate does not involve a significant increase
in the probability of an accident previously evaluated. The
consequences of a non-isolated reactor containment building at the
time of a fuel-handling accident or LOCA [loss-of-coolant accident]
is release of radionuclides to the environment. Analyses have
conservatively assumed that a containment pressure relief system
line is open at the time of an accident, and release to the
environment continues until the isolation valves are closed. In
addition, LOCA analyses assume containment leakage of 0.1% of the
containment volume per day for the first 24 hours and 0.05% per day
for the duration of the accident. Consequently, increasing the
interval between surveillances of isolation valve leakrate does not
involve a significant increase in the consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not involve a modification to the
physical configuration of the plant (i.e., no new equipment will be
installed) or change in the methods governing normal plant
operation. The proposed change will not impose any new or different
requirements or introduce a new accident initiator, accident
precursor, or malfunction mechanism. The functions of the
containment purge, hydrogen purge and containment pressure relief
systems are not altered by this change. Therefore, this proposed
change does not create the possibility of an accident of a different
kind than previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
This proposed change only increases the interval between
surveillance tests of the containment purge, hydrogen purge and
containment pressure relief system valves. Analyses have
conservatively assumed that the containment purge valves are open at
the time of a fuel handling accident, and that the containment
pressure relief valve is open at the time of a loss-of-coolant
accident. In addition, LOCA analyses assume containment leakage of
0.1% of the containment volume per day for the first 24 hours and
0.05% per day for the duration of the accident. The radiological
consequences of both an fuel handling accident and a LOCA are
unchanged and remain within the 10 CFR 100 limits. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: October 17, 2003.
Description of amendment request: The licensee is proposing to
revise Technical Specification (TS) Section 5.5.6, ``Containment Tendon
Surveillance Program,'' for consistency with the requirements of 10 CFR
50.55a(g)(4) for components classified as Code Class CC. The proposed
revision to
[[Page 64141]]
TS 5.5.6 is to indicate that the Containment Tendon Surveillance
Program, inspection frequencies, and acceptance criteria shall be in
accordance with Section XI, Subsection IWL of the ASME Boiler and
Pressure Vessel Code and the applicable addenda as required by 10 CFR
50.55a, except where an exemption or relief has been authorized by the
NRC. The licensee has also proposed to delete the provisions of
Surveillance Requirement (SR) 3.0.2 from this specification. In
addition, the licensee is proposing to revise TS 5.5.16, ``Containment
Leakage Rate Testing Program,'' to add exceptions to Regulatory Guide
1.163, ``Performance-Based Containment Leak-Testing Program.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change revises the TS administrative controls
programs for consistency with the requirements of 10 CFR
50.55a(g)(4) for components classified as Code Class CC. The revised
requirements do not affect the function of the containment post-
tensioning system components. The post-tensioning systems are
passive components whose failure modes could not act as accident
initiators or precursors.
The proposed change affects the frequency of visual examinations
that will be performed for the concrete surfaces of the containment
for the purpose of the Containment Leakage Rate Testing Program. In
addition, the proposed change allows those examinations to be
performed during power operation[,] as opposed to during a refueling
outage. The frequency of visual examinations of the concrete
surfaces of the containment and the mode of operation during which
those examinations are performed has no relationship to or adverse
impact on the probability of any of the initiating events assumed in
the accident analyses. The proposed change would allow visual
examinations[,] that are performed pursuant to NRC approved ASME
Section XI Code requirements (except where relief has been granted
by the NRC)[,] to meet the intent of visual examinations [as]
required by Regulatory Guide 1.163, without requiring additional
visual examinations pursuant to the Regulatory Guide. The intent of
early detection of deterioration will continue to be met by the more
rigorous requirements of the Code[-]required visual examinations. As
such, the safety function of the containment as a fission product
barrier is maintained.
The proposed change does not impact any accident initiators or
analyzed events or assumed mitigation of accident or transient
events. They do not involve the addition or removal of any
equipment, or any design changes to the facility.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change revises the TS administrative controls
programs for consistency with the requirements of 10 CFR
50.55a(g)(4) for components classified as Code Class CC. The
function of the containment post-tensioning system components are
not altered by this change. The change affects the frequency of
visual examinations that will be performed for the concrete surfaces
containments. In addition, the proposed change allows those
examinations to be performed during power operation[,] as opposed to
during a refueling outage. The proposed change does not involve a
modification to the physical configuration of the plant (i.e., no
new equipment will be installed) or change in the methods governing
normal plant operation. The proposed change will not impose any new
or different requirements or introduce a new accident initiator,
accident precursor, or malfunction mechanism. Additionally, there is
no change in the types or increases in the amounts of any
effluent[s] that may be released off-site and there is no increase
in individual or cumulative occupational exposure.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change revises the TS administrative controls
programs for consistency with the requirements of 10 CFR
50.55a(g)(4) for components classified as Code Class CC. The
function of the containment post-tensioning system components are
not altered by this change. The change affects the frequency of
visual examinations that will be performed for the concrete surfaces
containments. In addition, the proposed change allows those
examinations to be performed during power operation[,] as opposed to
during a refueling outage. The safety function of the containment as
a fission product barrier will be maintained.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear
Generating Unit No. 2, (IP2) Westchester County, New York
Date of application for amendment: March 27, 2002, as supplemented
May 30, 2002, July 10, 2002, October 10, 2002, October 28, 2002,
November 26, 2002, December 18, 2002, January 6, 2003, January 27,
2003, February 26, 2003, April 8, 2003, May 19, 2003, June 23, 2003,
June 26, 2003, July 15, 2003, August 6, 2003, September 11, 2003,
October 8, 2003, and October 14, 2003.
Brief description of amendment: The licensee proposed to convert
the current Technical Specifications (TSs) for IP2, to a set of
improved TSs based on NUREG-1431, ``Standard Technical Specifications
for Westinghouse Plants,'' Revision 2, dated April 2001.
Date of publication of individual notice in Federal Register:
September 26, 2003 (68 FR 55660).
Expiration date of individual notice: October 27, 2003.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
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
[[Page 64142]]
Commission's rules and regulations. The Commission has made appropriate
findings as required by the Act and the Commission's rules and
regulations in 10 CFR Chapter I, which are set forth in the license
amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by email to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendment: January 16, 2003, as
supplemented June 11, 2003.
Brief description of amendment: The amendment revised the Technical
Specifications to incorporate changes associated with Cycle 15 core
reload design analysis. The Cycle 15 core reload design implements the
Framatome ANP Statistical Core Design methodology . This amendment
permits the licensee to determine the minimum departure from nucleate
boiling ratio using an NRC-approved methodology based on statistical
analysis of operational and design uncertainties.
Date of issuance: October 20, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 247.
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 18, 2003 (68 FR
12948). The supplement dated June 11, 2003, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated October 20, 2003.
No significant hazards consideration comments received: No.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: November 7, 2002, as
supplemented by letters dated April 25, July 10, July 30, August 13,
September 18, and October 1, 2003.
Brief description of amendments: The amendments revise Technical
Specification (TS) 3.2.4, ``Departure From Nucleate Boiling Ratio
(DNBR),'' TS 3.3.1, ``Reactor Protective System (RPS) Instrumentation--
Operating,'' TS 3.3.3, ``Control Element Assembly Calculators
(CEACs),'' and TS 5.4.1, ``Administrative Controls--Procedures.'' The
revisions are to Limiting Conditions for Operations (LCOs), LCO
Actions, LCO Surveillance Requirements, and the procedures used to
modify the core protection calculator addressable constants.
Date of issuance: October 24, 2003.
Effective date: October 24, 2003, and shall be implemented for Unit
1 no later than prior to entry of Unit 1 into Mode 4 during the restart
from the Unit 1 spring 2004 refueling outage; for Unit 2 within 90 days
of the date of issuance, but no later than prior to entry of Unit 2
into Mode 4 during the restart from the Unit 2 fall 2003 refueling
outage; and for Unit 3 no later than prior to entry of Unit 3 into Mode
4 during the restart from the Unit 3 fall 2004 refueling outage.
Amendment Nos.: Unit 1-150, Unit 2-150, Unit 3-150.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revise the Technical Specifications.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75868) with a later notice on August 18, 2003 (68 FR 49527).
The August 13, September 18 and October 1, 2003, supplemental
letters provided clarifying information that was within the scope of
the Federal Register Notice (68 FR 49257) and did not change the no
significant hazards consideration determination.
The Commission's related evaluation of the amendments are contained
in a Safety Evaluation dated October 24, 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: December 16, 2002, as supplemented by
letter dated September 11, 2003.
Brief description of amendment: The amendment revises the current
main steam isolation valve (MSIV) Technical Specification (TS) 3/4
7.1.5 to more closely reflect TS 3.7.2 contained in NUREG-1432,
Revision 2. In addition, this change removes the MSIVs from the scope
of containment isolation valve TS 3/4 6.3 such that only TS 3/4.7.1.5
will apply to the MSIVs.
Date of issuance: October 21, 2003.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 190.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5671).
The licensee attached a revised no significant hazards
consideration (NSHC) determination with the supplement dated September
11, 2003. This revised NSHC determination contained minor wording
changes as compared with the NSHC determination sent with the original
application dated December 16, 2002, changes made to reflect the new TS
changes, and provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the conclusions of the original NSHC
determination.
The Commission's related evaluation of the amendment is contained
in a
[[Page 64143]]
Safety Evaluation dated October 21, 2003.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: May 12, 2003, as supplemented by
letter dated August 7, 2003.
Brief description of amendment: The amendment would revise the
Technical Specifications to remove the MODE restrictions for
performance of Surveillance Requirements 3.8.4.7 and 3.8.4.8 for the
Division 3 direct current electrical power subsystem.
Date of issuance: October 27, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No: 159.
Facility Operating License No. NPF-29: The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: June 10, 2003 (68 FR
34665). The August 7, 2003, supplemental letter provided clarifying
information that did not change the scope of the original Federal
Register notice or the original no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 27, 2003.
No significant hazards consideration comments received: No.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of application for amendments: December 20, 2002, as
supplemented August 15, 2003.
Brief description of amendments: The amendments provide editorial
and administrative changes to the Technical Specifications. The changes
correct typographical, spelling, numbering syntax, page break, and font
consistency errors as well as removing blank pages and associated
references. There are no substantive changes made in the proposed
amendment.
Date of issuance: October 21, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos: 224 and 219.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5677). The supplemental letter provided clarifying information that was
within the scope of the initial notice and did not change the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in an Environmental Assessment dated October 17, 2003, and a Safety
Evaluation dated October 21, 2003.
No significant hazards consideration comments received: No.
GPU Nuclear Inc., Docket No. 50-320, Three Mile Island Nuclear Station,
Unit 2, Dauphin County, Pennsylvania
Date of amendment request: July 21, 2003.
Brief description of amendment request: The amendment revises the
technical specification (TS) administrative controls to make the Three
Mile Island (TMI) Unit 2 radioactive effluent control program
consistent with the program for the TMI Unit 1 operating reactor TS.
The proposed change adopts the TMI Unit 1 liquid discharge limits since
both Units 1 and 2 use the same liquid discharge monitor and have a
common discharge pathway. The gaseous discharge limits will also be
updated to reflect the current 10 CFR 20 nomenclature along with some
minor editorial changes. Additionally, the definition of a member of
the public will be made consistent with the definition in 10 CFR 20.
Date of issuance: October 20, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 60.
Facility Operating License No. DPR-73: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 18, 2003 (68
FR 54750).
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated October 20, 2003.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station, Unit No. 1, Oswego County, New York
Date of application for amendment: November 15, 2002, as
supplemented by letters dated January 15, July 31, and September 15,
2003.
Brief description of amendment: The amendment revised the reactor
coolant system pressure-temperature limit curves and tables in Section
3/4.2.2, ``Minimum Reactor Vessel Temperature for Pressurization,'' of
the Technical Specifications. The revised curves and tables are
effective up to 28 effective full-power years.
Date of issuance: October 27, 2003.
Effective date: October 27, 2003, to be implemented within 60 days.
Amendment No.: 183.
Facility Operating License No. DPR-63: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75882).
The supplemental letters of January 15, July 31, and September 15,
2003, provided clarifying information that did not change the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 27, 2003.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of application for amendment: April 30, 2003.
Brief description of amendment: The amendment revises Technical
Specification Section 5.3, ``Plant Staff Qualifications,'' to update
requirements that have been outdated based on licensed operator
training programs being accredited by the National Academy for Nuclear
Training and promulgation of the revised 10 CFR Part 55, ``Operators''
Licenses.''
Date of issuance: October 24, 2003.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 212.
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 10, 2003 (68 FR
34670).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 24, 2003.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: October 17, 2002.
Brief description of amendments: The amendment revises Technical
[[Page 64144]]
Specification 3.7.9, ``Control Room Emergency Filtration System
(CREFS),'' by deleting the one-time extension to the allowed outage
time (AOT) for CREFS and the exception requirements of Limiting
Condition for Operation 3.04 and Surveillance Requirement 3.04 that
were allowed during the AOT.
Date of issuance: October 16, 2003.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 210 and 215.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 18, 2003 (68
FR 7818).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 16, 2003.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: April 30, 2003.
Brief description of amendments: The amendments revise Technical
Specification Section 5.3, ``Plant Staff Qualifications,'' to update
requirements that have been outdated based on licensed operator
training programs being accredited by the National Academy for Nuclear
Training and promulgation of the revised 10 CFR Part 55, ``Operators''
Licenses.''
Date of issuance: October 24, 2003.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 211 and 216.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 10, 2003 (68 FR
34670).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 24, 2003.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket No. 50-323, Diablo Canyon
Nuclear Power Plant, Unit 2, San Luis Obispo County, California
Date of application for amendment: June 26, 2003, as supplemented
by letters dated September 3 and September 30, 2003.
Brief description of amendments: The amendment authorizes revisions
to the Diablo Canyon Power Plant (DCPP) Final Safety Analysis Report
(FSAR) Update to incorporate the NRC approval of a revised steam
generator (SG) voltage-based repair criteria probability of detection
(POD) method for DCPP Unit No. 2. The revised POD, based on the
probability of prior cycle detection method, is approved to determine
the beginning of cycle voltage distribution for DCPP Unit 2 Cycle 12
operational assessment.
Date of issuance: October 21, 2003.
Effective date: October 21, 2003, and shall be implemented within
30 days of the date of issuance. The implementation of the amendment
includes the incorporation into the FSAR Update the changes discussed
above, as described in the licensee's application dated June 26, 2003,
and supplements dated September 3 and September 30, 2003, and evaluated
in the staff's Safety Evaluation attached to the amendment.
Amendment No.: 164.
Facility Operating License No. DPR-82: The amendment authorized
revision of the FSAR Update.
Date of initial notice in Federal Register: July 22, 2003 (68 FR
43392). The supplemental letters dated September 3 and September 30,
2003, provided additional clarifying information, did not expand the
scope of the application as originally noticed, and did not change the
NRC staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 21, 2003.
No significant hazards consideration comments received: No.
PSEG Nuclear, LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: April 11, 2003, as supplemented
on August 28 and September 22, 2003.
Brief description of amendments: The amendments modify the Salem
Nuclear Generating Station, Unit Nos. 1 and 2, Technical Specifications
(TS) Surveillance Requirements (SRs) 4.3.1.1.3 and 4.3.2.1.3, and TS
Bases Sections B 3/4.3.1 and B 3/4.3.2 relating to response time
testing of the Engineered Safety Features Actuation System and the
Reactor Trip System. In addition, the amendment for Salem, Unit No. 1,
deletes a footnote associated with SR 4.3.2.1.3, regarding a one-time
extension to the SR, that is no longer required.
Date of issuance: October 28, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment Nos.: 260 and 241.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the TSs.
Date of initial notice in Federal Register: June 10, 2003 (68 FR
34672).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 28, 2003.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of application for amendment: January 14, 2003, as
supplemented by letters dated July 1, 2003, and August 20, 2003.
Brief description of amendment: This amendment revises Technical
Specification 4.4.5.3.a, maximum inspection interval from 40 calendar
months to 58 calendar months after two consecutive inspections which
were classified as C-1.
Date of issuance: October 29, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 165.
Facility Operating License No. NPF-12: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 4, 2003 (68 FR
10280).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 29, 2003.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: March 25, 2003.
Brief description of amendments: The proposed changes would revise
Technical Specification 3.5.2, ``Emergency Core Cooling Systems
(ECCS)--Operating,'' Surveillance Requirement 3.5.2.5. Specifically,
the changes replace the requirement to verify specific surveillance
test values for the ECCS pumps with the requirement to verify the
developed head for each ECCS pump in accordance with the inservice
testing Program.
[[Page 64145]]
These changes are requested to implement recommendations of the
Standard Technical Specifications for Combustion Engineering Plants,
NUREG-1432, Revision 2.
Date of issuance: October 24, 2003.
Effective date: October 24, 2003, to be implemented within 60 days
of issuance.
Amendment Nos.: Unit 2-190; Unit 3-181.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 15, 2003 (68 FR
18285).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 24, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendment: January 14, 2003 (TS 02-08).
Brief description of amendment: The proposed amendments revised
applicability requirements for Technical Specification (TS) 3.3.9.4,
``Containment Building Penetrations.'' This modified the applicability
requirement associated with movement of ``irradiated fuel'' by adding a
new applicability statement for the containment building equipment
door. The requested also modified the current licensing basis to
replace the current accident source term used in the design basis fuel
handling accident radiological analyses with alternate source term.
Date of issuance: October 28, 2003.
Effective date: As of the date of issuance and shall be implemented
within 45 days of issuance.
Amendment Nos.: 288 and 278.
Facility Operating License No. DPR-77: Amendments revised the TSs.
Date of initial notice in Federal Register: February 18, 2003 (68
FR 7822).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 28, 2003.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 3rd day of November 2003.
For the Nuclear Regulatory Commission.
Eric J. Leeds,
Deputy Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 03-28065 Filed 11-10-03; 8:45 am]
BILLING CODE 7590-01-P