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Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations
Note: EPA no longer updates this information, but it may be useful as a reference or resource.
[Federal Register: August 6, 2002 (Volume 67, Number 151)]
[Notices]
[Page 50947-50965]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr06au02-106]
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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, July 12, 2002, through July 25, 2002. The
last biweekly notice was published on July 23, 2002 (67 FR 48213).
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
[[Page 50948]]
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 September 5, 2002, 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,\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 subparagraphs (d)(1) and (2),
regarding petitions to intervene and contentions. Those provisions
are extant and still applicable to petitions to intervene. Those
provisions are as follows: ``In all other circumstances, such ruling
body or officer shall, in ruling on--
(1) A petition for leave to intervene or a request for hearing,
consider the following factors, among other things:
(i) The nature of the petitioner's right under the Act to be
made a party to the proceeding.
(ii) The nature and extent of the petitioner's property,
financial, or other interest in the proceeding.
(iii) The possible effect of any order that may be entered in
the proceeding on the petitioner's interest .
(2) The admissibility of a contention, refuse to admit a
contention if:
(i) The contention and supporting material fail to satisfy the
requirements of paragraph (b)(2) of this section; or
(ii) The contention, if proven, would be of no consequence in
the proceeding because it would not entitle petitioner to relief.''
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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
hearingdocket@nrc.gov. 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 OGCMailCenter@nrc.gov. 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).
[[Page 50949]]
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 pdr@nrc.gov.
AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New York
Date of amendment request: June 26, 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 next 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 1.09 to 1.12 for three-recirculation-loop
operation, and to 1.11 for four-or five-recirculation-loop operation.
The proposed amendment would also editorially revise references to
topical reports which document the approved methodology, and make
editorial corrections to the TSs.
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?
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 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 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. The balance of the changes is purely
administrative. 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 balance of the changes is purely
administrative. 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.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant (BSEP), Units 1 and 2, Brunswick County,
North Carolina
Date of amendments request: June 26, 2002.
Description of amendments request: The proposed amendment would
revise the Technical Specifications (TS) to revise the reactor coolant
system pressure-temperature limit curves for operation to 32 effective
full-power years (EFPY).
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 license amendments do not involve a significant
increase in the probability or consequences of an accident
previously evaluated. Development of the revised BSEP, Unit 1 and 2
pressure-temperature limits was performed using the approved
fracture toughness methodologies of 10 CFR 50, Appendix G; the
American Society of Mechanical Engineers (ASME) Boiler and Pressure
Vessel Code, Section XI, Appendix G; and ASME Code Case N-640,
``Alternative Reference Fracture Toughness for Development of P-T
Limit Curves for ASME Section XI, Division 1.'' The revised
pressure-temperature limits were also developed using NRC Regulatory
Guide 1.190, ``Calculational and Dosimetry Methods for Determining
Pressure Vessel Neutron Fluence,'' March 2001, for evaluating
neutron fluence and NRC Regulatory Guide 1.99, Revision 2,
``Radiation Embrittlement of Reactor Vessel Materials,'' for
evaluating predicted irradiation effects on vessel beltline
materials. Use of these methods provides compliance with the intent
of 10 CFR 50, Appendix G, and provides adequate protection against
nonductile-type fractures of the reactor pressure vessel. Therefore,
the probability of occurrence of a previously analyzed event is not
significantly increased.
The consequences of a previously evaluated accident are
dependent on the initial conditions assumed for the analysis, the
behavior of the fuel during the accident, the availability and
successful functioning of the equipment assumed to operate in
response to the accident, and the setpoints at which these actions
are initiated. The proposed revisions do not impact the source term
or pathways assumed in accidents previously evaluated. No analysis
assumptions are violated, and there are no adverse effects on the
factors contributing to offsite and onsite dose. The proposed
changes to the pressure-temperature limits curves do not affect the
performance of any equipment used to mitigate the consequences of a
previously evaluated accident. Also, the proposed changes do not
affect setpoints that initiate protective or mitigative actions.
Based on the above, the proposed changes to the pressure-temperature
limits curves do not significantly increase the consequences of a
previously evaluated accident.
2. The proposed license amendments will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The changes extend the pressure-temperature limits for use up to
32 EFPY of
[[Page 50950]]
operation while providing adequate protection against a nonductile-
type fracture of the reactor pressure vessel. Creation of the
possibility of a new or different kind of accident would require the
creation of one or more new precursors of that accident. New
accident precursors may be created by modifications of the plant
configuration, including changes in allowable modes of operation.
This proposed license amendment does not involve any facility
modifications, and plant equipment will not be operated in a
different manner. Also, no new initiating events or transients
result from the pressure-temperature limits curves changes. As a
result, no new failure modes are being introduced. Therefore, the
proposed changes to the pressure-temperature limits curves will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed license amendments do not involve a significant
reduction in a margin of safety.
The margin of safety is established through the design of the
plant structures, systems, and components; through the parameters
within which the plant is operated; through the establishment of
setpoints for actuation of equipment relied upon to respond to an
event; and through margins contained within the safety analyses. The
proposed changes to the pressure-temperature limit curves do not
adversely impact the performance of plant structures, systems,
components, and setpoints relied upon to respond to mitigate an
accident. The revised pressure-temperature limits were developed
using the approved fracture toughness methodologies of 10 CFR 50,
Appendix G; the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code, Section XI, Appendix G; and ASME
Code Case N-640, ``Alternative Reference Fracture Toughness for
Development of P-T Limit Curves for ASME Section XI, Division 1.''
The proposed changes are acceptable because the ASME guidance
maintains the relative margin of safety commensurate with that which
existed at the time that the ASME Boiler and Pressure Vessel Code,
Section XI, Appendix G, was approved in 1974. In addition, the
revised pressure-temperature limits were also developed using NRC
Regulatory Guide 1.190, ``Calculational and Dosimetry Methods for
Determining Pressure Vessel Neutron Fluence,'' March 2001, for
evaluating neutron fluence and NRC Regulatory Guide 1.99, Revision
2, ``Radiation Embrittlement of Reactor Vessel Materials'' for
evaluating predicted irradiation effects on vessel beltline
materials. Use of these methods has provided revised pressure-
temperature limit curves that will ensure that the reactor pressure
vessel materials continue to behave in a non-brittle manner, thereby
preserving the original safety design bases[.]
No plant safety
limits, setpoints, or design parameters are adversely affected by
the proposed changes to the pressure-temperature limit curves.
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: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Kahtan Jabbour, Acting.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant (BSEP), Units 1 and 2, Brunswick County,
North Carolina
Date of amendments request: July 2, 2002.
Description of amendments request: The proposed amendments would
revise the Technical Specifications (TS) to change the administrative
controls of TS 5.7, ``High Radiation Area.'' The proposed changes would
be consistent with the guidance of Regulatory Guide 8.38, ``Control of
Access to High and Very High Radiation Areas in Nuclear Power Plants,''
Section C, Regulatory Position 2.4, Alternative Methods for Access
Control, with the exception that ``should'' would be changed to
``shall.''
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 license amendments do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The changes are administrative and affect personnel access
control requirements for high radiation areas. The changes do not
affect the operation, physical configuration, or function of plant
equipment or systems. The changes do not impact the initiators or
assumptions of analyzed events; nor do they impact the mitigation of
accidents or transient events. Therefore, these changes do not
increase the probability or consequences of an accident previously
evaluated.
2. The proposed license amendments will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The changes are administrative and affect personnel access
control requirements for high radiation areas. The changes do not
alter plant configuration, require installation of new equipment,
alter assumptions about previously analyzed accidents, or impact the
operation or function of plant equipment or systems. Therefore,
these changes will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. The proposed license amendments do not involve a significant
reduction in a margin of safety.
The changes are administrative and affect personnel access
control requirements for high radiation areas. The changes do not
impact any safety assumptions; nor do the changes have the potential
to reduce any margin of safety as described in the BSEP TS Bases.
The proposed changes maintain an equivalent level of protection for
radiation workers and, thereby, provide reasonable assurance that
individuals will not exceed regulatory dose limits. The proposed
changes are consistent with: (1) the guidance of Regulatory Guide
(RG) 8.38, ``Control of Access to High and Very High Radiation Areas
in Nuclear Power Plants,'' Section C, Regulatory Position 2.4,
Alternative Methods for Access Control, with the exception that
``should'' has been changed to ``shall''; (2) the BSEP TSs prior to
conversion to Improved Standard Technical Specifications; and (3)
other nuclear plants' existing TSs, including the Crystal River, H.
B. Robinson, and Shearon Harris nuclear plants.
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: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Kahtan Jabbour, Acting.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: July 8, 2002.
Description of amendment request: The amendment would revise
Technical Specification (TS) 3/4.8.1.1, ``Electrical Power Systems--
A.C. Sources--Operating'' and TS 3/4.8.1.2, ``Electrical Power Systems-
-A.C. Sources--Shutdown'' by revising the minimum level to a volume-
based indication versus a level-based indication.
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 amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The Harris Nuclear Plant (HNP) Technical Specification (TS)
Bases for Electrical Power Systems--A. C. Systems states that; ``A
[[Page 50951]]
separate day tank containing a minimum of 1457 gallons of fuel,
which is equivalent to a minimum indicated level of 40% * * *'' and,
the asterisked note states; * * * Minimum indicated level with a
fuel oil specific gravity of 0.83 and the level instrumentation
calibrated to a reference specific gravity of 0.876.'' These changes
do not modify the design or operation of Structures, Systems, and
Components (SSCs) that could initiate an accident. The minimum
volume of fuel in the day tank is unchanged by this amendment and
consequently would not impact the probability or consequences of any
accident scenario.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve new plant components or
procedures, but only revise existing Technical Specification
Limiting Condition for Operation Requirements. No significant impact
on any postulated accident is made due to this change since the
required fuel oil volume is not changed and the level indication for
the operations personnel is not changed. These changes do not modify
the design or operation of Structures, Systems, and Components
(SSCs) that could initiate an accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed changes do not affect the design or operation of
safety related components relied upon to automatically mitigate the
consequences of a design basis event. The day tank level specified
in TS is not accurate for all fuel oil specific gravities so these
changes provide better monitoring capability by reducing the
possibility of confusion. Indicated day tank level is used to
determine volume by comparing the indicated level to the day tank
curve using the actual specific gravity of the fuel. The Diesel
Generator day tank minimum volume is not altered by these changes
and therefore there * * * is no significant impact on any safety
system and these changes do not reduce the margin of safety.
Based on these considerations, 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: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Kahtan N. Jabbour, Acting.
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: July 11, 2002.
Description of amendment request: The proposed amendments would
revise the Technical Specifications to make several administrative
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:
First Standard
Would implementation of this amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
No. This license amendment request makes editorial corrections
to several Oconee Technical Specifications. These corrections are
solely administrative in nature. The deletion of the Reactor
Building Engineered Safeguards Channels, as proposed in the change
to the Technical Specification 3.3.6, Engineered Safeguards
Protective System Manual Initiation, was investigated through Duke's
corrective action program and also confirmed to be administrative in
nature. Therefore, all the changes contained in this license
amendment request are administrative in nature and have no impact on
any accident probabilities or consequences.
Second Standard
Would implementation of this amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
No. There are no new accident causal mechanisms created as a
result of the implementation of this license amendment request. No
changes are being made to the plant which will introduce any new
accident causal mechanisms. This amendment request only makes
administrative changes and does not impact any plant systems that
are accident initiators; therefore, no new accident types are being
created.
Third Standard
Would implementation of this statement involve a significant
reduction in a margin of safety?
No. Margin of safety is related to the confidence in the ability
of the fission product barriers to perform their design functions
during and following an accident situation. The changes proposed in
this license amendment request are administrative in nature and do
not affect the performance of the barriers. Consequently, no safety
margins will be impacted.
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.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: January 10, 2002.
Description of amendment request: Energy Northwest is requesting
changes to the technical specifications (TS) to reflect the application
of a 24-month surveillance test interval (STI) to coincide with its
intention to implement a 24-month fuel cycle.
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 extension of the intervals to 24 months for the subject SRs
[surveillance requirements]
does not impact the ability of any of
the equipment to function as assumed in the Columbia Generating
Station accident analysis. None of the equipment within the scope of
analysis for this TS amendment request performs a function in any of
the systems required for safe shutdown as described in section 7.4
of the Columbia Generating Station FSAR [Final Safety Analysis
Report]. Historical maintenance and surveillance data as well as
projected instrument drift indicate the proposed amendment will not
affect performance or reliability of the equipment tested to meet
the requirements of these SRs. Therefore, the extension of the
surveillance intervals 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.
An event related to surveillance testing Frequency or
instruments drifting beyond Allowable Values is not postulated in
the Columbia Generating Station accident analysis. None of the
analyses performed for this amendment request indicate an increase
in the probability of equipment failure resulting from the
surveillance interval extension. Because all of the equipment
related to the proposed SR interval extensions is expected to
function normally during the longer intervals, extending the subject
SRs does not introduce any new accident initiators.
Therefore, the operation of Columbia Generating Station in
accordance with the
[[Page 50952]]
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed amendment to the Technical Specifications will
extend the intervals at which testing is performed to meet the
requirements of the selected SRs. The overall effect of the
extensions on safety is small due to other more frequent testing
that is performed on the same equipment, projected instrument drift
that is bounded by the current setpoint analysis, or the existence
of redundant mechanical or electrical components. Reviews of
historical surveillance and maintenance records indicate there is no
evidence of time-related failures. The proposed amendment does not
impact the performance of any system, structure, or component relied
upon for accident mitigation. The proposed surveillance interval
extensions do not impact any safety analysis assumptions or results.
Therefore, operation of Columbia Generating Station in
accordance with the proposed amendment will 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: Thomas C. Poindexter, Esq., Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: June 24, 2002.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) surveillance requirements (SR)
3.7.7.1 and SR 3.7.7.2. Specifically, SR 3.7.7.1 would be changed to
require the verification of the city water tank volume rather than city
water header pressure and increase the SR frequency from 12 hours to 24
hours. SR 3.7.7.2 would be revised to require all city water header
isolation valves are open rather than only the one header supply
isolation valve.
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 license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response: No.
The current TS surveillance to verify City Water (CW) header
pressure did not provide assurance that adequate volume of water was
available in the City Water Tank (CWT) as an alternate source of
cooling if Condensate Storage Tank (CST) was not available. The CST
is not designed to withstand the effect of a tornado-generated
missile. However, the Auxiliary Feedwater System (AFS) is provided
sufficient redundancy of water supplies such that an alternate
source of water from the CWT is available in the event the CST is
damaged by a tornado-generated missile. The proposed amendment to
verify CWT volume is [ge]360,000 gallons would ensure that adequate
volume of CW is available in the CWT to cool the RCS [reactor
coolant system]
from 102% rated thermal power to RHR [residual heat
removal]
entry conditions in 10 hours, if the CST is unavailable or
depleted for any reason. The surveillance frequency for the CWT
volume is 24 hours. The proposed amendment to change SR 3.7.7.2 to
include additional isolation valves that are in the flow path from
CWT to AFS suction would ensure that all applicable isolation valves
in the flow path are properly positioned. Thus, the proposed
amendment involves changes to the Technical Specifications that
would properly reflect the Surveillance Requirements for CWT. The
CWT is not an initiator of any accident addressed in the FSAR [Final
Safety Analysis Report]
and the proposed amendment does not have any
change to the accident analysis addressed in the FSAR.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment involves changes to the Technical
Specifications to properly reflect the surveillance requirements of
City Water Tank. The proposed change provides assurance of
availability of adequate volume of water in the CWT to cool the RCS
from 102% rated thermal power to RHR entry conditions in 10 hours,
if the CST is unavailable or depleted for any reason, and verifies
the correct position of isolation valves in the flow path between
the CWT and the AFS pump suction. These changes do not affect any
accident initiators.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed license amendment involve a significant
reduction in a margin of safety?
Response: No.
The proposed amendment involves changes to the Technical
Specifications to properly reflect the surveillance requirements of
City Water Tank. The proposed change to verify the CWT volume would
ensure that an adequate volume of CW is available in the tank to
cool the RCS from 102% rated thermal power to RHR entry conditions
in 10 hours, if the CST is unavailable or depleted for any reason.
The proposed change to verify the valve position for isolation
valves in the flow path between the CWT and the AFS pump suction
would ensure that isolation valves in the flow path are properly
positioned. The proposed amendment does not involve any changes to
plant equipment, or the way in which the plant is operated.
Therefore, the proposed amendment 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: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: June 26, 2002.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 5.6.5.b, ``Core Operating Limits
Report (COLR),'' to incorporate the reference to Westinghouse topical
report WCAP-12945-P-A, ``Code Qualification Document for Best Estimate
Loss-of-Coolant Analysis [LOCA],'' dated March 1998. The proposed
amendment would also allow the use of the analytical methodology to
determine the core operating limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
No physical changes are being made by this change. The proposed
changes involve use of the Best Estimate Large Break LOCA [loss-of-
coolant accident]
analysis methodology and associated TS [technical
specification]
changes. The plant conditions assumed in the analysis
are bounded by the design conditions for all equipment in the plant.
Therefore, there will be no increase in
[[Page 50953]]
the probability of a loss of coolant accident. The consequences of a
LOCA are not being increased. That is, it is shown that the
emergency core cooling system is designed so that its calculated
cooling performance conforms to the criteria contained in 10 CFR
50.46 paragraph b, that is it meets the five criteria listed in
Section II of this evaluation. No other accident is potentially
affected by this change.
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 previously analyzed?
Response: No.
There are no physical changes being made to the plant. No new
modes of plant operation are being introduced. The parameters
assumed in the analysis are within the design limits of existing
plant equipment. All plant systems will perform equally during the
response to a potential accident.
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
the margin of safety?
Response: No.
It has been shown that the analytic technique used in the
analysis more realistically describes the expected behavior of the
Indian Point 3 reactor system during a postulated loss of coolant
accident. Uncertainties have been accounted for as required by 10
CFR 50.46. A sufficient number of loss of coolant accidents with
different break sizes, different locations and other variations in
properties have been analyzed to provide assurance that the most
severe postulated loss of coolant accidents were calculated. It has
been shown by the analysis that there is a high level of probability
that all criteria contained in 10 CFR 50.46 paragraph b) are met.
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. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: June 26, 2002.
Description of amendment request: Extend the use of the pressure-
temperature (P-T) limits in Technical Specification (TS) Figure
3.4.6.1-1 to 32 effective full power years by deleting a note on each
unit's TS Figure limiting the validity of the Figure.
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?
No. The proposed change to the technical specifications to
extend the use of the existing pressure-temperature (P-T) limits
does not affect the operation or configuration of any plant
equipment. Thus, no new accident initiators are created by this
change. The existing P-T limits are based on the projected reactor
vessel neutron fluence at 32 effective full power years (EFPY) of
operation specified in the current licensing basis for LGS [Limerick
Generating Station], Units 1 and 2. A plant-specific calculation of
reactor vessel 32 EFPY fast neutron fluence has been completed for
LGS, Units 1 and 2, using the methodology described in a General
Electric (GE) Company Licensing Topical Report (LTR), which adheres
to the guidance in Regulatory Guide 1.190, ``Calculational and
Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.''
The three-dimensional spatial distribution of neutron flux was
modeled by combining the results of two separate two-dimensional
neutron transport calculations. The latest available cross section
libraries for the important components of Boiling Water Reactor
(BWR) neutron flux calculations, i.e., oxygen, hydrogen and
individual iron isotopes, were included. The resulting reactor
vessel fast neutron fluence value is lower than the value in the
current licensing basis for LGS, Units 1 and 2. Therefore, the
existing 32 EFPY P-T limits bound the fast neutron fluence value
calculated using the GE methodology. This provides sufficient
assurance that the LGS, Unit 1 and Unit 2, reactor vessels will be
operated in a manner that will protect them from brittle fracture
under all operating conditions. 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?
No. The proposed change to the technical specifications to
extend the use of the existing P-T limits does not affect the
operation or configuration of any plant equipment. The current P-T
limits will remain valid and conservative during the proposed
extension. 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?
No. The proposed change extends the use of the existing P-T
limits. The existing P-T limits are based on the projected reactor
vessel neutron fluence at 32 EFPY of operation specified in the
current licensing basis for LGS, Units 1 and 2. A plant-specific
calculation of reactor vessel 32 EFPY fast neutron fluence has been
completed for LGS, Units1 and 2, using the NRC [Nuclear Regulatory
Commission]
approved methodology in a GE LTR, which adheres to the
guidance in Regulatory Guide 1.190. The three-dimensional spatial
distribution of neutron flux was modeled by combining the results of
two separate two-dimensional neutron transport calculations. The
latest available cross section libraries for the important
components of BWR neutron flux calculations, i.e., oxygen, hydrogen
and individual iron isotopes, were included. The resulting reactor
vessel fast neutron fluence value is lower than the value in the
current licensing basis for LGS, Units 1 and 2. Therefore, the
existing 32 EFPY P-T limits bound the fast neutron fluence value
calculated using the GE methodology. This provides sufficient margin
such that the LGS, Unit 1 and Unit 2, reactor vessels will be
operated in a manner that will protect them from brittle fracture
under all operating conditions. 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. Edward Cullen, Vice President & General
Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett
Square, PA 19348.
NRC Acting Section Chief: Jacob I. Zimmerman.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket No. 50-
277, Peach Bottom Atomic Power Station, Unit 2, York County,
Pennsylvania
Date of application for amendment: June 10, 2002
Description of amendment request: Exelon Generation Company, LLC,
the licensee, is proposing a change to the Peach Bottom Atomic Power
Station (PBAPS), Unit 2, Technical Specifications (TSs) contained in
Appendix A to the Operating License. This proposed change will revise
the TS section on safety limits to incorporate revised safety limit
minimum critical power ratios (SLMCPRs) due to the cycle-specific
analysis performed by Global Nuclear Fuel for PBAPS, Unit 2, Cycle 15.
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 50954]]
issue of no significant hazards consideration, which is presented
below:
1. The proposed TS change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The derivation of the cycle specific safety limit minimum
critical power ratios (SLMCPRs) for incorporation into the (TS[s]),
and their use to determine cycle specific thermal limits, has been
performed using the methodology discussed in ``General Electric
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-
II), and U.S. Supplement, NEDE-24011-P-A-14-US, June, 2000, which
incorporates Amendment 25. Amendment 25 was approved by the NRC in a
March 11, 1999 safety evaluation report.
The basis of the SLMCPR calculation is to ensure that greater
than 99.9% of all fuel rods in the core avoid transition boiling if
the limit is not violated. The new SLMCPRs preserve the existing
margin to transition boiling. The GE-14 fuel is in compliance with
Amendment 22 to ``General Electric Standard Application for Reactor
Fuel,'' NEDE-24011-P-A-14 (GESTAR-II), and U.S. Supplement, NEDE-
24011-P-A-14-US, June, 2000, which provides the fuel licensing
acceptance criteria. The probability of fuel damage will not be
increased as a result of this change. Therefore, the proposed TS
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The SLMCPR is a TS numerical value, calculated to ensure that
transition boiling does not occur in 99.9% of all fuel rods in the
core if the limit is not violated. The new SLMCPRs are calculated
using NRC approved methodology discussed in ``General Electric
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-
II), and U.S. Supplement, NEDE-24011-P-A-14-US, June, 2000, which
incorporates Amendment 25. Additionally, the GE-14 fuel is in
compliance with Amendment 22 to ``General Electric Standard
Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-II), and
U. S. Supplement, NEDE-24011-P-A-14-US, June, 2000, which provides
the fuel licensing acceptance criteria. The SLMCPR is not an
accident initiator, and its revision will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
There is no significant reduction in the margin of safety
previously approved by the NRC as a result of the proposed change to
the SLMCPRs, which includes the use of GE-14 fuel. The new SLMCPRs
are calculated using methodology discussed in ``General Electric
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-
II), and U.S. Supplement, NEDE-24011-P-A-14-US, June, 2000, which
incorporates Amendment 25. The SLMCPRs ensure that greater than
99.9% of all fuel rods in the core will avoid transition boiling if
the limit is not violated when all uncertainties are considered,
thereby preserving the fuel cladding integrity. Therefore, the
proposed TS change will not involve a significant reduction in the
margin of safety previously approved by the NRC.
Based on the above, Exelon Generation Company, LLC, concludes
that the proposed amendment presents no significant hazards
consideration under the standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of ``no significant hazards consideration''
is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for Licensee: Mr. Edward Cullen, Vice President and
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way,
Kennett Square, PA 19348.
NRC Section Chief: Jacob I. Zimmerman, Acting.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: July 18, 2002
Description of amendment request: The proposed amendments would
implement an administrative change to relocate the Technical
Specifications (TS) requirements for the spent fuel crane to the
respective unit's Updated Final Safety Analysis Report (UFSAR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would operation of the facility in accordance with the
proposed amendment involve a significant increase in the probability
or consequences of an accident previously evaluated?
No. The proposed changes to the Technical Specifications are
administrative in nature in that the Technical Specifications for
operation and surveillance of the spent fuel cask crane and the fuel
handling crane will be relocated from Appendix A of the facility
operating license to the UFSAR for each unit. The crane operation
and surveillance requirements are not altered by this relocation.
Once relocated, any future changes will be controlled by 10 CFR
50.59, and the UFSARs will be updated pursuant to 10 CFR 50.71(e).
Because no operating requirements are changed by the proposed
amendment, crane operation following the proposed amendment would
not differ from current crane operation. The proposed Technical
Specification changes do not involve any change to the configuration
or method of operation of any plant equipment that is used to
mitigate the consequences of an accident, nor do the changes alter
any assumptions or conditions in any of the plant accident analyses.
Therefore, facility operation in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated in
the UFSAR.
2. Would operation of the facility in accordance with the
proposed amendment create the possibility of a new or different kind
of accident from any accident previously evaluated?
No. The proposed amendment will not affect the design function
of any system, structure, or component. Relocating the existing
Technical Specification requirements for the spent fuel cask crane
and the fuel handling crane to the UFSAR is an administrative change
and will not modify the physical plant or the modes of plant
operation defined in the Facility Operating License. The operating
restrictions imposed on the spent fuel-related cranes by the
existing Technical Specifications will be retained in the UFSAR
under this change. The change does not involve the addition or
modification of equipment, nor does it alter the design or operation
of plant systems. Therefore, operation of the facility in accordance
with the proposed amendment would not create the possibility of a
new or different accident from any accident previously evaluated.
3. Would operation of the facility in accordance with the
proposed amendment involve a significant reduction in a margin of
safety?
No. The proposed changes to the Technical Specifications are
administrative in nature in that the Technical Specifications for
operation and surveillance of the spent fuel cask crane and the fuel
handling crane will be relocated from Appendix A of the facility
operating license to the UFSAR for each unit. The crane operating
restrictions that are being relocated to the UFSAR by this change
are not being relaxed or eliminated. The proposed changes do not
alter the basis for any technical specification that is related to
the establishment of or the maintenance of a nuclear safety margin.
Therefore, operation of the facility in accordance with the proposed
amendment will not involve a significant reduction in a margin of
safety as defined in the basis for any Technical Specification or in
any licensing document.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
[[Page 50955]]
NRC Section Chief: Kahtan N. Jabbour, Acting.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: July 3, 2002.
Description of amendment request: The amendment would revise the
Improved Technical Specifications (ITS) 3.8.1 and associated bases,
``AC Sources--Operating,'' by extending the allowed outage time for the
emergency diesel generators (EDGs) from 72 hours to 14 days and to
modify a note for two EDG ITS Surveillance Requirements (SRs).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does not involve a significant increase in the probability
or consequences of an accident previously analyzed.
The proposed license amendment extends the Completion Time for
restoring an inoperable EDG to OPERABLE status and permits
performance of certain SRs at power under specified conditions. The
EDGs are designed to supply backup AC power to equipment in
essential safety systems in the event of a loss of offsite power,
and as such, the EDGs are not initiators of any design basis
accident.
The design functions, operational characteristics, and
interfaces between the EDGs and other plant systems will not be
affected by the change. In addition, the initial conditions and
assumptions for accidents that require the EDGs will remain
unchanged. Defense in depth will be maintained by the redundant
OPERABLE EDG, diverse 1E offsite power sources, and the availability
of multiple emergency feedwater (EFW) and auxiliary feedwater (AFW)
equipment capable of operating independently of both offsite power
and the EDGs.
A Probabilistic Safety Assessment (PSA) has been performed to
quantitatively assess the risk impact of an increase in Completion
Times. Although the proposed changes result in slight increases in
core damage frequency (CDF) and incremental conditional core damage
probability (ICCDP), and large early release frequency (LERF) and
incremental conditional large early release probability (ICLERP),
these increases are well below values that are considered risk
significant in accordance with current regulatory guidance.
Based on the above, the proposed changes will not significantly
increase the probability or consequences of an accident previously
evaluated.
(2) Does not create the possibility of a new or different kind
of accident from any accident previously analyzed.
The proposed amendment extends the Completion Time for restoring
an inoperable EDG to OPERABLE status and permits performance of
certain SRs at power under specified conditions. The proposed
amendment will not result in changes to the design, physical
configuration or operation of the plant or the assumptions made in
the safety analysis for accidents that require the EDGs. In
addition, the proposed amendment will not result in changes to
corrective or preventive maintenance activities associated with the
EDGs, plant operating procedures, or the procedures used to respond
to abnormal or emergency conditions. Assumptions made in the safety
analysis related to EDG availability will also remain unchanged.
Performance of certain SRs at power requires an evaluation to assure
plant safety is maintained or enhanced, which would include
evaluation for new or different plant conditions. As such, no new
failure modes are being introduced. Therefore, the proposed change
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
(3) Does not involve a significant reduction in the margin of
safety.
The proposed license amendment increases the Completion Times
for restoring an inoperable EDG to OPERABLE status and permits
performance of certain SRs at power under specified conditions. The
proposed changes will improve EDG reliability by providing
flexibility in scheduling and performing EDG preventive and
corrective maintenance activities. This flexibility will reduce the
probability (and associated risk) of a plant shutdown to repair an
inoperable EDG that cannot be restored within the current ITS 3.8.1
Completion Times. Performance of the proposed SRs at power requires
an evaluation to assure plant safety is maintained or enhanced. The
proposed change will also increase the availability of the EDGs
during MODE 5 and 6 outages, thus reducing shutdown risk.
The proposed amendment will not change the plant design, safety
analysis, or the design, configuration or operation of the EDGs. The
EDGs are designed to supply backup AC power to equipment in
essential safety systems in the event of a loss of offsite power.
Either EDG is capable of performing this function; therefore, as
long as one train is available, the margin of safety is maintained.
Defense in depth will be provided by the redundant OPERABLE EDG, the
availability of diverse offsite circuits capable of supplying power
to plant emergency loads, and EFW and AFW equipment that can perform
their design function independently of both offsite power and the
EDGs.
To ensure these defense in depth capabilities are maintained
during required EDG maintenance, maintenance and surveillance
activities that have the ability to impact the availability of the
redundant EDG, required support systems and/or backup systems, the
EFW and AFW systems and the 1E offsite power circuits will be
controlled in accordance with the normal work controls process. As
part of this process, weekly qualitative and quantitative risk
assessments of scheduled on-line maintenance activities, and
additional risk assessments of emergent work activities, will be
performed in accordance with the guidance provided in CR-3
Compliance Procedure CP-253, ``Power Operation Risk Assessment and
Management.'' If the results of these assessments indicate an
increase in risk, appropriate actions to control temporary and
aggregate risk increases and minimize risk increases above the
overall plant baseline will be implemented in accordance with CP-
253.
Additional measures to minimize risk will include increased
administrative controls related to switchyard access, and increased
inspection of identified risk significant fire areas within the
plant. A Tier 2 analysis has also been performed to identify the
dominant risk significant plant configurations during the time that
an EDG is inoperable due to required corrective or preventive
maintenance, and appropriate configuration controls/restrictions
will be established prior to extended EDG maintenance.
As discussed in question (1) above and in the submittal, the
slight increases in CDF, ICCDP, LERF and ICLERP resulting from the
proposed amendment are all below values that are considered risk
significant in accordance with the guidance provided in Regulatory
Guide 1.174, ``An Approach for Using Probabilistic Risk Assessment
in Risk-Informed Decisions on Plant-Specific Changes to the
Licensing Basis,'' for changes to the plant, and Regulatory Guide
1.177, ``An Approach for Plant-Specific, Risk-Informed
Decisionmaking: Technical Specifications,'' for proposed increases
in ITS Completion Times.
Based on the above, this 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: R. Alexander Glenn, Associate General
Counsel (MAC-BT15A), Florida Power Corporation, P.O. Box 14042, St.
Petersburg, Florida 33733-4042.
NRC Acting Section Chief: Kahtan N. Jabbour.
GPU Nuclear Inc., Docket No. 50-320, Three Mile Island Nuclear
Generating Station, Unit 2, (TMI-2) Dauphin County, Pennsylvania
Date of amendment request: June 13, 2002.
Description of amendment request: The proposed technical
specifications change request (TSCR) No. 79, Revision 1, is to revise
Three Mile Island Nuclear Generating Station, Unit 2 (TMI-2) Technical
Specification (TS) Administrative Controls section that will provide
consistency with Three Mile Island Nuclear Generating Station, Unit 1,
(TMI-1) TS changes submitted
[[Page 50956]]
by AmerGen Energy Company, LLC (AmerGen) and Exelon Generation Company,
LLC (EGC), which are currently under review by the U.S. Nuclear
Regulatory Commission (NRC). GPU Nuclear utilizes EGC/AmerGen
administrative controls under contract to TMI-2. The proposed request
would delete TS Sections 6.4, ``Training,'' and 6.5.4, ``Independent
Onsite Safety Review Group'' (IOSRG) from the administrative
requirements in Section 6 of the TMI-2 Post Defueled Monitored Storage
(PDMS) TS. Additionally, the IOSRG has been removed from the list of
recipients of audit reports in Section 6.5.3.2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
TMI-2 is a defueled facility holding a Possession Only License
is being maintained in Post Defueling Monitored Storage (PDMS). The
introduction of the PDMS Quality Assurance Plan states in part in
the second paragraph, ``Since the plant will be in a non-operating
and defueled status, there will no longer be any structures,
systems, or components that perform a safety function.''
Deletion of the technical specifications requirements for
training and the IOSRG will have no adverse effect on any plant
system; will not alter the source term, containment isolation, or
allowable radiological consequences. These administrative changes
will have no effect on any plant systems, structures or components
and do not affect the physical plant, operating procedures,
maintenance procedures, or emergency procedures at TMI-2.
The elimination of the IOSRG oversight function removes a
function that is redundant to other oversight programs, not required
by NRC regulation, and is not needed for the safe monitoring of TMI-
2. Programmatic assessments of the TMI-2 programs will continue to
be assessed by Nuclear Oversight personnel in accordance with the
PDMS Quality Assurance Plan. Training will continue to be conducted
in accordance with regulatory requirements.
The training programs for appropriate unit staff personnel other
than licensed operators is now addressed by 10 CFR 50.120. With the
10 CFR 50.120 rule, the NRC is emphasizing the need to ensure that
industry personnel training programs are based upon job performance
requirements. This will be accomplished using the systems approach
to training implemented by INPO [Institute of Nuclear Power
Operations]
accredited training programs for selected nuclear
personnel. Included within the rule is the requirement that the
training program must reflect industry experience. Deletion of the
training requirements in the technical specifications will conform
the license to the current requirements of 10 CFR 50.120.
Therefore, these changes will 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?
These changes are administrative in nature and do not affect any
system functional requirements, plant maintenance, or operability
requirements. The proposed changes involve the elimination of a
redundant oversight function and the replacement of training
requirements by the more vigorous requirements of 10 CFR 50.120,
which are applicable to operating plants.
The proposed changes have no direct effect on any plant systems
or components. The programs for the monitoring, surveillance, or
maintenance of TMI-2 are unaffected. Oversight of TMI-2 will
continue to be provided by Nuclear Oversight personnel and the TMI-2
Safety Oversight Committee in accordance with the requirements of
the PDMS Quality Assurance Plan.
Therefore, the proposed changed will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The training and IOSRG requirements contained in TMI-2 Technical
Specifications Section 6.0 ``Administrative Controls'' are
administrative in nature. The proposed changes have no direct effect
on any plant systems. There are currently no safety limits that
apply to TMI-2 during PDMS. 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 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 & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Robert A. Gramm.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: June 28, 2002.
Description of amendment request: The licensee proposed changes to
surveillance requirements in Table 4.6.2b, ``Instrumentation that
Initiates Primary Coolant System or Containment Isolation,'' of the
Nine Mile Point Nuclear Station, Unit No. 1 (NMP1) Technical
Specifications (TS) regarding the isolation capability of the shutdown
cooling system (SDCS). Specifically, the changes will remove the
restriction to perform channel functional testing and channel
calibration associated with SDCS high area temperature only during
refueling outages. The changes will allow these surveillance activities
to be performed during other operating conditions on a once-per-
operating-cycle basis, thereby maintaining SDCS availability to support
reactor shutdown operations during refueling.
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 Nine Mile Point Unit 1 in accordance with
the proposed amendment will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The only safety-related functions of the SDCS are (i) to
maintain the integrity of the reactor coolant pressure boundary, and
(ii) to provide primary containment isolation of the shutdown
cooling lines. The proposed amendment removes an unnecessary
restriction to perform channel functional testing and calibration
associated with SDCS isolation capability only during refueling
outages. It provides the flexibility to perform these surveillances
during other operating conditions on a ``once per operating cycle''
basis. The change does not modify the surveillance frequency,
surveillance acceptance criteria, high area temperature setpoint
limit for initiating SDCS isolation, plant equipment configurations
during SDCS surveillances, or the existing requirements for
maintaining SDCS isolation and reactor coolant pressure boundary
integrity.
Based on the above, the operation of NMP1 in accordance with the
proposed amendment will not involve a significant increase in the
probability or the consequences of an accident previously evaluated.
2. The operation of Nine Mile Point Unit 1 in accordance with
the proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change does not involve any physical modifications
to the plant and does not alter equipment configuration, setpoints,
safety parameters, surveillance interval durations, or surveillance
acceptance criteria. It does not affect the operation of any safety-
related structure, system, or component in a manner that could
introduce a new accident precursor or a new failure mechanism. The
SDCS isolation valves will continue to perform their isolation
function by remaining closed with power removed during power
operation of the reactor.
Based on the above, the operation of NMP1 in accordance with the
proposed amendment cannot create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The operation of Nine Mile Point Unit 1 in accordance with
the proposed
[[Page 50957]]
amendment will not involve a significant reduction in a margin of
safety.
The proposed change does not affect any of the plant's fission
product barriers or safety/operational limits. The high area
temperature setpoint for SDCS isolation will remain within the
existing TS limit.
The SDCS isolation valves will continue to remain closed with
power removed during power operation of the reactor. The proposed
``[o]nce per operating cycle'' surveillances will be adequate to
ensure acceptable SDCS equipment operability and reliability.
Based on the above, the operation of NMP1 in accordance with the
proposed amendment will not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Richard J. Laufer.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: July 12, 2002.
Description of amendment request: The licensee proposed to change
the Technical Specifications (TSs), Sections 3.1.1 and 4.1.1, ``Control
Rod System,'' by reducing the power level below which the rod worth
minimizer (RWM) or a second independent verification of rod positions
must be used from 20% rated thermal power (RTP) to 10% RTP. The
licensee stated that analysis has shown that no significant control rod
drop accident (CRDA) can occur above 10% RTP. The low power setpoint
change will reduce the time necessary for both reactor startup and
shutdown.
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 reproduced below:
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The TS revision lowers the power level at which the analyzed rod
position sequence must be followed by use of the RWM or a second
independent verification of rod positions. The RWM enforces the
analyzed rod position sequence to ensure that the initial conditions
of the CRDA analysis are not violated. Compliance with the analyzed
rod position sequence and operability of the RWM is required in the
startup and run modes when thermal power is less than 10% RTP. When
thermal power is 10% RTP or greater, there is no possible control
rod configuration that results in a control rod worth that could
exceed the 280 cal/gram fuel design limit during a CRDA. None of the
accidents previously evaluated assume the RWM is an initiator of the
accident and therefore, the probability of an accident is not
significantly increased by the change. Because the fuel design limit
is not exceeded, the change to the low power setpoint will not
significantly increase the consequences of an accident previously
evaluated.
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The TS revision lowers the power level below which the analyzed
rod position sequence must be followed. The change does not
introduce a new mode of plant operation and does not involve a
physical modification to the plant. Therefore, a new or different
type of accident from any accident previously evaluated is not
created.
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
The RWM enforces the analyzed rod position sequence to ensure
that the initial conditions of the CRDA analysis are not violated.
Compliance with the analyzed rod position sequence and operability
of the RWM are required in the startup and run modes when thermal
power is less than 10% RTP. When thermal power is 10% RTP and
greater, there is no possible control rod configuration that results
in a control rod worth that could exceed the 280 cal/gram fuel
design limit during a CRDA. Because the fuel design limit is not
exceeded at 10% RTP and greater, the change to the RWM low power
setpoint does not significantly reduce 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: 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: July 12, 2002.
Description of amendment request: The proposed amendment would
revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications
(TS) Section 3.1.a.3, ``Pressurizer Safety Valves.'' Also, the proposed
amendment would reformat TS 3.1.a.3 to more closely resemble the format
of Improved Standard Technical Specification (ISTS) to improve clarity.
The proposed amendment would allow both pressurizer safety valves to be
inoperable or removed while the reactor vessel head is on. This would
only be applicable when the temperature and pressure are low enough
such that the Low Temperature Overpressure Protection (LTOP) System can
safely protect the Reactor Coolant System (RCS). The TSs currently
requires the LTOP System to protect the RCS when the RCS temperature is
less than LTOP enabling temperature.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The format changes are administrative in nature and therefore
have no effect on the probability or consequences of an accident.
The situation where the plant has two inoperable or removed
pressurizer safeties while the LTOP System is enabled is not
considered an accident initiator. Therefore, any change to the
system would not affect the probability of an accident previously
evaluated. The risk of core damage/release of radioactivity would
not increase with all of the other plant safety features still in
place.
The proposed changes adds clarity to the TSs by describing a
specific situation when the RCS is at low temperature & pressure
while overpressure protection is provided by the LTOP System. Since
this TS change is not an accident initiator and existing TS will
ensure the LTOP System will continue to protect the RCS pressure
boundary, this proposed amendment does not involve an 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 situation where the plant has two inoperable pressurizer
safeties while the LTOP System is enabled is not considered an
accident initiator. A failure of this system will not result in an
accident. The format changes are administrative in nature and
therefore have no effect on the probability or consequences of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve a change to the physical
plant or operations. As the RCS temperature is lowered to less than
200 deg.F, the LTOP System provides the RCS overpressure protection
required. Since the LTOP System is currently approved for use by TS
3.1.b.4, it would not create the
[[Page 50958]]
possibility of a new or different kind of accident from any accident
previously evaluated.
Therefore, any change to the system would not affect the
probability of an accident previously evaluated.
3. Involve a significant reduction in the margin of safety.
The format changes are administrative in nature and therefore
are not involved in a significant reduction in the margin of safety.
Margin of safety relates to overpressure protection when the RCS is
less than 200 deg.F. This margin is controlled by the LTOP System
completely and does not rely on the pressurizer safeties. This
proposed amendment allows KNPP to have both pressurizer safeties to
be inoperable as long as the RCS is below the LTOP System enabling
temperature. Therefore, NMC concludes that there is not 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: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
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 Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: June 24, 2002.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 5.5.3, ``Post Accident Sampling
System (PASS),'' to eliminate the requirements to have and maintain the
PASS at Plant Hatch. 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 application 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 June 24,
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 [Three Mile Island, Unit 2]
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 amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes 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.
[[Page 50959]]
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: July 11, 2002.
Description of amendment request: The proposed amendments would
delete Technical Specification 3.3.1.1.I.2, which requires returning
the Oscillating Power Range Monitor to operable status within 120 days
of discovering its operability.
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 Oscillating Power Range Monitor (OPRM) is not designed for
the prevention of an instability event or any other previously
evaluated event. Accordingly, it cannot increase the probability of
an instability event or any other previously evaluated event.
The consequences of the instability event are not significantly
increased, because the alternate method of detection and suppression
of thermal-hydraulic instability oscillations is well established at
Plant Hatch. Furthermore, operators are adequately trained on
instabilities.
This proposed change to delete the 120-day Completion Time
restriction on an inoperable OPRM does not affect any other system
designed for the mitigation of previously analyzed events.
For the above reasons, the probability and consequences of a
previously analyzed event are not increased.
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 only deletes a Technical Specification
requirement. It does not physically alter the design, operation,
testing, or maintenance of any plant system or piece of equipment.
The proposed change introduces no new modes of operation.
Consequently, the change does not create the possibility of a new or
different kind [of]
event.
3. The change does not involve a significant reduction in the
margin of safety.
The proposed change deletes the requirement to restore the OPRM
system to operable status within 120 days of discovering its
inoperability. A manual alternate method to detect and suppress
thermal-hydraulic instability oscillations has been included in
Plant Hatch procedures for many years. Also, operators are trained
on instability events.
Accordingly, the manual alternate method is adequate and thus,
the margin of safety for the instability event 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.
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: August 2, 2001.
Description of amendment request: The proposed amendment revises
Technical Specifications to extend, on a one-time basis, the current
interval for Type A testing from 10 years to 15 years.
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 proposed Technical Specification revision extends the
current interval for Type A testing. The current test interval of
ten years would be extended on a one-time basis to 15 years from the
preceding Type A test. Pursuant to 10 CFR 50.91, this analysis
provides a determination that the proposed change to the Technical
Specifications for a one-time extension of the interval for
Integrated Leakage Rate Testing does not involve any significant
hazards consideration as defined in 10 CFR 50.92.
Criterion 1: The proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
The proposed extension to the Type A testing interval will not
increase the probability of an accident previously evaluated. The
containment Type A testing interval extension is not a modification
and the testing interval extension is not of a type that could lead
to equipment failure or accident initiation.
The proposed extension to the Type A testing interval does not
involve a significant increase in the consequences of an accident.
Research documented in NUREG-1493 has determined that Type B and C
tests can identify the vast majority (more than 95%) of all
potential leakage paths.
NUREG-1493 concluded that reducing the Type A test frequency to
one per twenty years leads to an imperceptible increase in risk.
Testing and inspection provide a high degree of assurance that the
containment will not degrade in a manner detectable only by Type A
testing. Previous Type A tests show leakage does not exceed
acceptance criteria, indicating a very leak-tight containment.
Inspections required by the Maintenance Rule and ASME code are
performed in order to identify indications of containment
degradation that could affect leak tightness.
Experience at the South Texas Project demonstrates that
excessive containment leakage paths are detected by Type B and C
Local Leakage Rate Tests. Type B and C testing will identify any
containment opening, such as a valve, that would otherwise be
detected by the Type A tests. These factors show that a Type A test
interval extension will not involve a significant increase in the
consequences of an accident.
Criterion 2: The proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
The proposed extension of the Type A testing interval will not
create the possibility of a new or different type of accident from
any previously evaluated. There are no physical changes being made
to the plant and there are no changes in operation of the plant that
could introduce a new failure mode creating an accident or affecting
the mitigation of an accident.
Criterion 3: The proposed change does not involve a significant
reduction in the margin of safety.
The proposed extension of the Type A testing interval will not
significantly reduce the margin of safety. The NUREG-1493 generic
study of the effects of extending containment leakage testing found
that a 20-year interval in Type A leakage testing results in an
imperceptible increase in risk to the public. NUREG-1493 found that,
generically, the design containment leakage rate contributes about
0.1 percent to the individual risk and that the decrease in Type A
testing frequency would have a minimal effect on this risk because
95% of the potential leakage paths are detected by Type B and C
testing.
Deferral of Type A testing for the South Texas Project does not
increase the level of public risk due to loss of capability to
detect and measure containment leakage or loss of containment
structural capability. Other containment testing methods and
inspections will assure all limiting conditions of operation will
continue to be met. The margin of safety inherent in existing
accident analyses is maintained.
Based on the evaluation provided above, the South Texas Project
concludes that the proposed change does not involve a significant
hazards consideration and will not have a significant effect on safe
operation of the plant. Therefore, there is reasonable
[[Page 50960]]
assurance that operation of the South Texas Project in accordance
with the proposed revised Technical Specifications will not endanger
the public health and 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: Alvin H. Gutterman, Esqr., Morgan, Lewis, &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: Robert A. Gramm.
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant,
Unit 1, Hamilton County, Tennessee
Date of amendment request: July 10, 2002.
Description of amendment request: The proposed one-time technical
specification (TS) change revises the Sequoyah Unit 2 Limiting
Condition for Operation for Section TS 3.7.4, ``Essential Raw Cooling
Water System,'' to include provisions for maintaining operability of
this system during performance of heavy load lifts associated with the
Unit 1 steam generator replacement (SGR) project. The provisions should
ensure safe operation of Unit 2 during heavy load lift activities. In
addition, compensatory measures proposed should ensure safe shutdown
capability of Unit 2 in the unlikely event a heavy load drop occurs
over Essential Raw Cooling Water system piping.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the Tennessee Valley
Authority (TVA) has provided its analysis of the issue of no
significant hazards consideration, which is presented below:
TVA has concluded that operation of Sequoyah (SQN) Unit 2, in
accordance with the proposed change to Technical Specification (TS)
3/4.7.4, does not involve a significant hazards consideration. TVA's
conclusion is based on its evaluation, in accordance with 10 CFR
50.91(a)(1), of the three standards set forth in 10 CFR 50.92(c).
TVA's proposed license amendment is a one-time change to the SQN
Unit 2 TSs. The proposed change revises SQN Limiting Condition for
Operation 3.7.4, ``Essential Raw Cooling Water System,'' to include
provisions for maintaining operability of this system during
performance of heavy load lifts associated with the Unit 1 steam
generator replacement (SGR) project. The provisions ensure safe
operation of Unit 2 during heavy load lift activities. In addition,
compensatory measures ensure safe shutdown capability of Unit 2 in
the unlikely event a heavy load drop occurs over ERCW system piping.
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
No changes in event classification as discussed in SQN Updated
Final Safety Analysis Chapter 15 will occur due to the proposed TS
amendment. The one-time TS provision ensures that the SQN essential
raw cooling water (ERCW) system remains operable for continued safe
operation of Unit 2 during heavy load lifts performed on Unit 1
during SGR replacement activities.
Accordingly, the proposed modification to SQN Unit 2 TSs and the
implementation of compensatory measures for a postulated load drop
will not significantly increase the probability or consequences of
an accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The possibility of a new or different accident scenario
occurring as a result of activities conducted during the SQN Unit 1
SGR project are not created. Three postulated scenarios related to
heavy load handling during the SGR project were examined for their
potential to represent a new or different kind of accident from
those previously evaluated: (1) a breach of the old steam generator
(OSG), resulting in the release of contained radioactive material,
(2) flooding in the Auxiliary Building caused by the failure of
piping in the ERCW tunnel, and (3) loss of ERCW to support safe
shutdown of the operating unit.
Failure of an OSG that results in a breach of the primary side
of the steam generator (SG) could potentially result in a release of
a contained source outside containment. The consequences of this
event, both offsite and in the control room, were examined and found
to be within the consequences of the failure of other contained
sources outside containment at the SQN site (i.e., within the SQN
design basis).
With regard to flooding of the Auxiliary Building from a heavy
load drop, the protective measure taken prior to the lifting of
heavy loads include installation of a wall in the ERCW tunnel near
the Auxiliary Building interface. The wall provides protection
against a postulated flood of the ERCW tunnel and protects against
flooding of the Auxiliary Building beyond those events previously
evaluated.
With regard to the potential for a heavy load drop causing the
loss of ERCW cooling water to the operating unit (i.e., Unit 2), TVA
is implementing provisions to preclude a load drop. A heavy load
drop is considered an unlikely accident for the following reasons:
The lifting equipment was specifically designed and
chosen for the subject heavy lifts,
Crane operators will be specially trained in the
operation of the lift equipment and in the SQN site conditions,
Qualifying analyses and administrative controls will be
used to protect the lifts from the effects of external events,
The areas over which a load drop could cause loss of
ERCW are a small part of the total travel path of the loads.
In addition, protection against the potential for a loss of ERCW
is established prior to any heavy load lifts. Compensatory measures
ensure the ERCW system is isolated should a pipe break occur, and
that ERCW flow is redirected to equipment essential for safe
shutdown capability of Unit 2.
Accordingly, the possibility of a new or different kind of
accident from any accident previously evaluated is not created.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change to the Unit 2 TSs support safe operation and
safe shutdown capability of Unit 2 during replacement of the Unit 1
SGs. These measures do not result in changes in the design basis for
plant structures, systems, and components (SSCs). Consequently, the
proposed change will not affect any margins of safety for plant
SSCs.
Accordingly, a significant reduction in the margin of safety is
not created by the proposed change.
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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A Knoxville, Tennessee 37902.
NRC Section Chief: Kahtan N. Jabbour, Acting.
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant,
Unit 1, Hamilton County, Tennessee
Date of application for amendments: March 29, 2002 (TS 02-02).
Brief description of amendments: The proposed amendment would
change the Sequoyah (SQN) Unit 1 Technical Specifications (TSs) by
revising Specification 3/4.4.5 to eliminate surveillance requirements
associated with two alternate repair criteria. The associated License
Condition 2.C.9.d is also deleted. In addition, the proposed change
revises SR 3/4.4.5.3.a to allow a one-time, 40-month steam generator
(SG) inspection interval after the first (post-Unit 1 SG replacement)
inservice inspection resulting in a C-1 category. The proposed change
is in lieu of the current TS criteria that requires two consecutive
category C-1 inspections for application of the 40-month SG inspection
interval.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), Tennessee Valley
Authority (TVA), the licensee, has provided its analysis of the issue
of no significant hazards consideration, which is presented below:
[[Page 50961]]
TVA has concluded that operation of Sequoyah Nuclear Plant (SQN)
Unit 1, in accordance with the proposed change to the technical
specifications and License Condition, does not involve a significant
hazards consideration. TVA's conclusion is based on its evaluation,
in accordance with 10 CFR 50.91(a)(1), of the three standards set
forth in 10 CFR 50.92(c).
TVA is proposing to modify SQN Unit 1 TS 3/4.4.5, ``Steam
Generators'' to delete surveillance requirements (SRs) that describe
steam generator (SG) tube plugging limits for two alternate repair
criteria (ARC). The first ARC is for axial outside diameter stress
corrosion cracking (ODSCC) at non-dented tube support plates and the
second ARC is for axial primary water stress corrosion cracking
(PWSCC) at dented tube support plates. TVA's proposed amendment
removes both ARCs through the deletion of the following SRs: SR
4.4.5.2.b.4, 4.4.5.2.d, 4.4.5.2.e, a portion of 4.4.5.4.a.6,
4.4.5.4.a.10, 4.4.5.4.a.11, 4.4.5.5.d, and 4.4.5.5.e. TVA's proposed
removal of these SRs for ARC reestablishes standard tube plugging
criteria within the TS for SQN Unit 1. Returning to the standard TS
40 percent through-wall tube plugging limit is inherently more
conservative.
Included with the above change is deletion of License Condition
2.C.9.d that references prior TVA commitment letters for SG
inspection. The TVA letters and their commitments will no longer
apply following replacement of the Unit 1 SGs.
In addition, TVA is proposing a revision to TS 3/4.4.5.3.a to
allow application of the 40-month inspection interval after one SG
inspection resulting in a C-1 category. The proposed change replaces
the current TS requirement that invokes the extended 40-month
inspection interval after two consecutive inspections resulting in a
category of C-1. TVA's proposed change provides a relaxation of the
SG inspection requirements and schedule. The relaxation in the
inspection schedule is intended to coincide with replacement of SQN
Unit 1 SGs during the Cycle 12 refueling outage (Spring 2003). The
replacement of the SQN Unit 1 SGs incorporate significant design
improvements that include thermally treated Alloy 690 SG tubing. The
improvements in SG design and tube material properties increase the
resistance to SG tube degradation mechanisms and allow optimization
of SG inspection schedules. The proposed optimization of SG
inspections reduce the cumulative number of SG inspections over the
life of the plant and result in significant dose, schedule, and cost
savings to TVA.
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
TVA's proposed TS amendment does not compromise limits
associated with SG tube integrity. TVA's proposed change removes
existing SG tube plugging criteria (i.e., ARC) from the TS and
reestablishes the standard TS criteria (40 percent through-wall
criteria). This change is inherently more conservative. The proposed
allowance for an extended inspection interval is a conservative
inspection strategy that is based on improved SG design features and
SG tube materials that have been shown to resist degradation and
preserve SG tube integrity.
The proposed revision does not alter plant equipment, test
methods or operating practices. The proposed change continues to
provide controls for safe operation of SQN SGs within the required
limits. The proposed change does not contribute to events or
assumptions associated with postulated design basis accidents (i.e.,
SG tube rupture). The proposed change does not affect operator
indicators or actions required to diagnose or mitigate a SG tube
rupture accident. The proposed revisions continue to maintain the
required safety functions. Accordingly, the probability of an
accident or the consequences of an accident previously evaluated is
not increased.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
TVA's proposed amendment removes existing repair criteria and
incorporates the more conservative TS limit for SG tube plugging
(i.e., plug tubes with degradation depths equal to or greater than
40 percent through-wall). This change will not give rise to new
failure modes. The failure of a SG tube to maintain leakage
integrity during operation is an analyzed event in the SQN Updated
Final Safety Analysis Report. TVA's proposed change to the SG
inspection interval will not introduce a new or different kind of
accident scenario. Accordingly, the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
TVA's proposed TS amendment is conservative with respect to the
margin of safety. The margin of safety is preserved through ensuring
structural integrity and leakage integrity of the SG tubes.
TVA's proposed change that to remove ARC from the TS does not
compromise structural integrity or leakage integrity of SG tubes.
The proposed change invokes the standard TS tube plugging criteria
limit (40 percent through-wall criteria) which is inherently
conservative.
TVA's proposed change to include a one-time extension to the SQN
Unit 1 SG inspection interval retains conservative inspection
strategy that maintains the structural and leakage integrity of the
SGs. TVA intends to replace SQN Unit 1 SGs during the Cycle 12
refueling outage and perform a 100 percent full length inspection of
SG tubes during the Cycle 13 refueling outage to verify that damage
mechanisms do not exist. Twelve years of SG operation history
indicate that corrosion damage mechanisms do not appear in
replacement SGs that contain thermally treated Alloy 690 tubing. The
replacement SG design also contains design improvements that provide
reasonable assurance that tube degradation is not likely to occur
over the proposed 40-month operating period (Cycle 13 refueling
outage to Cycle 15 refueling outage). The corrosion resistant
properties of the thermally treated Alloy 690 tubing and the
improved design will limit the initiation of damage mechanisms and
limit growth rate such that tube structural and leakage integrity
will be maintained over two operating cycles.
TVA's proposed change to extend the SG inspection interval does
not result in a change to system design features. The proposed
change does not affect the plant conditions, setpoints, or safety
limits that could result in precursors to accidents or degrade
accident mitigation systems. Accordingly, plant system safety
functions are not altered by the proposed change.
The effect of this change is to extend allowable SG inspection
intervals while retaining conservative margins to maintain the
structural and leakage integrity of the SGs. Consequently, the
proposed TS revisions does not reduce the margin of safety.
The NRC 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Kahtan N. Jabbour, Acting.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: July 10, 2002 (TS 01-09).
Brief description of amendments: The proposed amendment would
change the Sequoyah (SQN) Unit 1 and 2 Technical Specifications (TSs)
by removing the requirement to not make positive reactivity changes
during certain conditions and replace it with requirements to maintain
shutdown margin or boron concentration. The changes will permit limited
positive reactivity changes that are necessitated by plant operations.
These changes will limit the amount of reactivity changes to those that
will continue to assure appropriate reactivity limits are met. The
proposed changes are consistent with TS Task Force 286 and Revision 2
to NUREG-1431.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), Tennessee Valley
Authority (TVA), the licensee, has provided its analysis of the issue
of no significant hazards consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the
[[Page 50962]]
probability or consequences of an accident previously evaluated.
The proposed change does not involve an increase in the
probability or consequences of an accident previously evaluated. The
proposed activities to be allowed during certain operating
conditions are permitted at other times during routine operating
conditions. The changes do not affect the limits on reactivity that
are specified in other specifications. The proposed changes continue
to ensure restrictions on additions and flowpaths of unborated water
that are in the existing specifications. The proposed change does
not affect the limits on reactivity that are credited in the safety
analysis. Therefore, no increase in the probability or consequences
of any accident previously evaluated will occur.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes permit the conduct of normal operating
evolutions during limited periods when additional controls over
reactivity margin are imposed by the TSs. The proposed change does
not introduce any new equipment into the plant or significantly
alter the manner in which existing equipment will be operated. The
changes to operating allowances are minor and are only applicable
during certain conditions. The operating allowances are consistent
with those acceptable at other times. Since the proposed changes
only allow activities that are presently approved and routinely
conducted, no possibility exists for a new or different kind of
accident from those previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed changes do not involve a significant reduction in a
margin of safety because the ability to make the reactor subcritical
and maintain it subcritical during all operating conditions and
modes of operation will be maintained. The margin of safety is
defined by the shutdown margin limits and the refueling boron
concentration limit. The proposed changes do not affect these
operating restrictions and the margin of safety which assures the
ability to make and maintain the reactor subcritical is not
affected.
The NRC 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Kahtan N. Jabbour, Acting.
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.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
Date of amendment request: July 18, 2002.
Brief description of amendment request: These amendments would
revise the Facility Operating Licenses (FOLs) to change the
implementation date for the Improved Technical Specifications (ITS),
including the relocation of certain existing TS requirements to
licensee-controlled documents, from no later than September 2, 2002, to
no later than December 20, 2002.
Date of publication of individual notice in Federal Register: July
25, 2002 (67 FR 48679).
Expiration date of individual notice: August 26, 2002.
Notice of Issuance of Amendments to Facility Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management Systems (ADAMS) Public Electronic
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/
reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the NRC Public Document Room (PDR) Reference staff at
1-800-397-4209, 301-415-4737 or by e--mail to pdr@nrc.gov.
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of application for amendment: December 28, 2000, as
supplemented May 31, 2002.
Brief description of amendment: The amendment decreases the allowed
outage time for an inoperable channel or channels of the anticipated
transient without scram recirculation pump trip instrumentation.
Date of issuance: July 17, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 153.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 7, 2001 (66 FR
9378). The supplemental letter did not significantly change the
requested amendment or affect the proposed no significant hazards
consideration determination.
[[Page 50963]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 17, 2002.
No significant hazards consideration comments received: No.
Consumers Energy Company, Docket No. 50-155, Big Rock Point Plant,
Charlevoix, County, Michigan
Date of amendment request: July 31, 2001, as supplemented by
letters dated March 6, and April 23, 2002.
Brief description of amendment: The amendment revises License
Condition 2.C.(3) of Operating License DPR-6 to reference revisions of
the Big Rock Point Defueled Security Plan, Defueled Suitability
Training and Qualification Plan, Defueled Safeguards Contingency Plan,
and Independent Spent Fuel Storage Installation Security Plan.
Date of issuance: July 18, 2002.
Effective date: As of the date of issuance and shall be implemented
prior to placing the spent fuel in the Big Rock Point Plant independent
spent fuel storage installation.
Amendment No.: 123.
Facility Operating License No. DPR-6: The amendment revised the
Facility Operating License.
Date of initial notice in Federal Register: August 22, 2001 (66 FR
44166). The March 6 and April 23, 2002, supplemental letters provided
additional clarifying information that 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 July 18, 2002.
No significant hazards considerations comments received: No.
Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: February 5, 2002 as supplemented
on March 6, 2002.
Brief description of amendment: The amendment changes the term in
the technical specifications ``once each REFUELING INTERVAL'' to ``once
per 24 months'' in several surveillance requirements.
Date of issuance: July 24, 2002.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 206.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 28, 2002 (67 FR
36930). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 24, 2002.
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: December 20, 2001.
Brief description of amendments: The amendments revised the
Technical Specifications (TS) 5.5.14 to eliminate the use of the term
``unreviewed safety question,'' and replace the word ``involve'' with
the word ``require'' as it applies to changes made to the updated Final
Safety Analysis Report and the TS Bases.
Date of issuance: July 17, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 200 & 193.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 5, 2002 (67 FR
10010). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 17, 2002.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendment: December 20, 2001.
Brief description of amendments: The amendments revised the
Technical Specifications (TS) 5.5.14 to eliminate the use of the term
``unreviewed safety question,'' and replace the word ``involve'' with
the word ``require'' as it applies to changes made to the updated Final
Safety Analysis Report and the TS Bases.
Date of issuance: July 17, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 204 & 185.
Facility Operating License Nos. NPF-and NPF-17: Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: January 22, 2002 (67 FR
2921). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 17, 2002.
No significant hazards consideration comments received: No.
Energy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Energy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: January 31, 2002, as
supplemented by letter dated June 20, 2002.
Brief description of amendment: The amendment revises Technical
Specification 3.8.1, ``AC Sources-Operating,'' to extend the allowed
outage time for a Division 1 or Division 2 Diesel Generator from the
current 72 hours to 14 days.
Date of issuance: July 16, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No: 151.
Facility Operating License No. NPF-29: The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 2, 2002 (67 FR
15623). The June 20, 2002, 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 July 16, 2002.
No significant hazardous consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-a254 and 50-265, Quad
Cites Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: May 1, 2002.
Brief description of amendments: The amendments revise the start
delay time in the surveillance for the emergency diesel generators from
``[le]10 seconds'' to ``[le]13 seconds.''
Date of issuance: July 17, 2002.
Effective date: For Unit 2, as of the date of issuance and shall be
implemented within 30 days of the completion of Unit 1 refueling outage
17, which is scheduled for November 2002. For Unit 1, as of the date of
issuance and shall be implemented within 30 days following the date
when General Electric (GE)-14 fuel is loaded into the reactor, which is
scheduled during refueling outage 17 in November 2002. The amendment
may not be implemented prior to the date GE-14 fuel is loaded into the
reactor.
Amendment Nos.: 206 and 202.
[[Page 50964]]
Facility Operating License Nos. DPR-29 and DRP-30: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: a May 28, 2002 (67 FR
36931). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 17, 2002.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendments: July 24, 2001, as supplemented
June 5, and July 1.
Brief description of amendments: The amendments revise the Improved
Technical Specifications (ITS) to accommodate future changes in plant
design, including increased levels of Once--Through Steam Generator
(OTSG) tube plugging. The changes are categorized into two sets. The
first set of changes relocate parameters from the ITS to the cycle-
specific Core Operating Limits Report (COLR). These parameters are the
Variable Low Pressure Trip equation specified in ITS Table 3.3.1-1, and
Reactor Coolant System (RCS) pressure limit within Surveillance
Requirement (SR) 3.4.1.1. The second set of changes are applicable to
raising the OTSG tube plugging limit to a maximum of 20% equivalent of
all tubes, and addresses its impact. These changes include the revision
of the hot leg maximum temperature limit, and the revision of the RCS
minimum flow limits for four- and three-reactor coolant pump operation.
The RCS limits associated with 20% tube plugging will be maintained in
its ITS. Cycle-specific values of these limits, however, have been
relocated to the COLR. The hot leg temperature and RCS flow limit
values within SR 3.4.1.2 and 3.4.1.3 ``RCS Pressure, Temperature, and
Flow DNB [departure from nucleate boiling]
Limits,'' were relocated to
reflect their location in the COLR. For both sets of changes, ITS
5.6.2.18(a) was modified to reflect the relocation of cycle-specific
values from the ITS and the COLR.
Date of issuance: July 16, 2002.
Effective date: As of the date of issuance shall be implemented
within 60 days of issuance.
Amendment Nos.: 204.
Facility Operating License Nos. DPR-72: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: a August 22, 2001 (66
FR 44173). The supplemental letters provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 16, 2002.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: July 30, 2001, as supplemented by letter
dated August 23, 2001.
Description of amendment request: The amendment revises the Cooper
Nuclear Station's licensing basis.
Date of issuance: July 19, 2002.
Effective date: The amendment is effective on the date of issuance,
to be implemented within 30 days from the date of issuance.
Amendment No.: 192.
Facility Operating License No. DPR-46: Amendment revises the Cooper
Nuclear Station's licensing basis.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. June 25, 2002 (67 FR 42828). The notice
provided an opportunity to submit comments on the Commission's proposed
NSHC determination. No comments have been received. The notice also
provided an opportunity to request a hearing by July 29, 2002, but
indicated that, if the Commission makes a final NSHC determination, any
such hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated July 19, 2002.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station (CNS), Nemaha County, Nebraska.
Date of amendment request: May 20, 2002, as supplemented by letters
dated June 19, July 3 (two letters), and July 12, 2002. The letters
dated July 3 (two letters), and July 12, 2002, were of a clarifying
nature, did not expand the application beyond the scope of the initial
notice, and did not affect the staff's initial proposed no significant
hazards consideration determination.
Description of amendment request: The amendment revises the Cooper
Nuclear Station's Technical Specifications (TS) 3.7.2 and 3.7.3
reflecting increases in TS temperature limits for ultimate heat sink
and reactor equipment cooling water temperatures.
Date of issuance: July 22, 2002.
Effective date: The amendment is effective on the date of issuance,
to be implemented within 30 days from the date of issuance.
Amendment No.: 193.
Facility Operating License No. DPR-46: Amendment revises the Cooper
Nuclear Station's TS.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. 67 FR 43688 dated June 28, 2002. The notice
provided an opportunity to submit comments on the Commission's proposed
NSHC determination. No comments have been received. The notice also
provided an opportunity to request a hearing by July 12, 2002, but
indicated that if the Commission makes a final NSHC determination, any
such hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated July 22, 2002.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
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 application for amendments: September 20, 2001, as
supplemented by letters dated March 27 and April 12, 2002.
Brief description of amendments: The amendments revised the
Technical Specifications to support extension of the operating cycle
from 18 months to 24 months.
Date of issuance: July 12, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 232/174.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: November 28, 2001 (66
FR 59512). The supplements dated March 27 and April 12, 2002, provided
clarifying information that did not
[[Page 50965]]
change the scope of the September 20, 2001, application nor the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 12, 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: October 22, 2001, as supplemented by
letters dated May 16 and June 25, 2002.
Brief description of amendments: The amendments change TS 3/4.9.4,
``Refueling Operations--Containment Building Penetrations'', to allow
the equipment hatch to be open during core alterations or movement of
irradiated fuel within the containment.
Date of issuance: July 18, 2002.
Effective date: July 18, 2002.
Amendment Nos.: Unit 1--139; Unit 2--128.
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 May 16 and June 25, 2002, supplemental letters provided
clarifying information that was within the scope of the original
Federal Register notice and did not change the staff's initial no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 18, 2002.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 26th day of July 2002.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Acting Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 02-19420 Filed 8-5-02; 8:45 am]
BILLING CODE 7590-01-P