[Federal Register Volume 67, Number 190 (Tuesday, October 1, 2002)]
[Notices]
[Pages 61674-61694]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-24616]
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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, September 6, 2002, through September 19,
2002. The last biweekly notice was published on September 17, 2992 (67
FR 58635).
[[Page 61675]]
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, 11555 Rockville
Pike (first floor), Rockville, Maryland. The filing of requests for a
hearing and petitions for leave to intervene is discussed below.
By October 31, 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\ 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
[[Page 61676]]
participate fully in the conduct of the hearing, including the
opportunity to present evidence and cross-examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, 11555 Rockville Pike (first floor),
Rockville, Maryland, by the above date. Because of continuing
disruptions in delivery of mail to United States Government offices, it
is requested that petitions for leave to intervene and requests for
hearing be transmitted to the Secretary of the Commission either by
means of facsimile transmission to 301-415-1101 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and because of continuing disruptions in delivery of mail to United
States Government offices, it is requested that copies be transmitted
either by means of facsimile transmission to 301-415-3725 or by e-mail
to [email protected]. A copy of the request for hearing and
petition for leave to intervene should also be sent to the attorney for
the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, 11555 Rockville
Pike (first floor), Rockville, Maryland. Publicly available records
will be accessible from the Agencywide Documents Access and Management
System's (ADAMS) Public Electronic Reading Room on the Internet at the
NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not
have access to ADAMS or if there are problems in accessing the
documents located in ADAMS, contact the NRC PDR Reference staff at 1-
800-397-4209, 304-415-4737 or by e-mail to [email protected].
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: August 28, 2002.
Description of amendment request: The amendment would revise
Technical Specification (TS) 3/4.9.9, ``Containment Ventilation
Isolation System'' and associated Bases to allow the use of
administrative controls on open containment penetrations during core
alterations.
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 proposed changes modify TS requirements similar to that
previously reviewed and approved by the NRC in Harris Nuclear Plant
(HNP) License Amendment 104. The administrative controls proposed by
this change are currently being used for the same applicable
penetrations as part of TS 3.9.4. This change would permit opening
up the applicable penetrations under administrative controls if the
containment ventilation isolation system were inoperable. HNP has
demonstrated (in License Amendment 104) that the radiological
consequences were acceptable for a fuel handling accident occurring
simultaneously with an open penetration. For the purpose of the
applicable analysis, no credit was given for isolating the
penetration and dose consequences remained below applicable
regulatory limits. The proposed change does not modify the design or
operation of equipment used to move spent fuel or to perform core
alterations.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Containment penetrations are designed to form part of the
containment pressure boundary. The proposed change provides for
administrative controls and operating restrictions for containment
penetrations consistent with guidance approved by the NRC staff.
Containment penetrations are not an accident initiating system as
described in the Final Safety Analysis Report [FSAR]. The proposed
change does not affect other Structures, Systems, or Components. The
operation and design of containment penetrations in operational
modes 1-4 will not be affected by this proposed change.
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 modify similar required Actions previously
reviewed and approved by the NRC in HNP License Amendment 104. The
proposed change to containment penetrations does not significantly
affect any of the parameters that relate to the margin of safety as
described in the Bases of the TS or the FSAR. Accordingly, NRC
Acceptance Limits are not significantly affected by this change.
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: 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.
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: August 30, 2002.
Description of amendment request: The amendment would revise
Technical Specifications Definitions 1.13, Engineered Safety Features
(ESF) Response Time and 1.29, Reactor Trip System (RTS) Response Time.
Also proposed in this change request are revisions to Surveillance
Requirements 4.3.1.2 and 4.3.2.2 and Bases Sections B 3/4.3.1 and B 3/
4.3.2. These changes will revise the definition and surveillance
requirements for response
[[Page 61677]]
time testing of the Engineered Safety Feature Actuation System (ESFAS)
and the RTS. These changes are in conformance with changes approved in
WCAP-13632-P-A, Revision 2, and WCAP-14036-P-A, Revision 1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The change to the Harris Nuclear Plant (HNP) Technical
Specification (TS) does not result in a condition where the design,
material, and construction standards that were applicable prior to
the change are altered. The same RTS and ESFAS instrumentation is
being used; the time response allocations/modeling assumptions in
the Final Safety Analysis Report (FSAR) Chapter 15 analyses are
still the same; only the method of verifying the time response is
changed. The proposed change will not modify any system interface
and could not increase the likelihood of an accident since these
events are independent of this change. The proposed change will not
change, degrade or prevent actions or alter any assumptions
previously made in evaluating the radiological consequences of an
accident described in the FSAR.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
This change does not alter the performance of process protection
racks, Nuclear Instrumentation, and logic systems used in the plant
protection systems. Replacement transmitters will still have
response time verified by testing before being placed in operational
service. Changing the method of periodically testing these systems
(assuring equipment operability) from response time testing to
calibration and channel checks will not create any new accident
initiators or scenarios. Periodic surveillance of these systems will
continue and may be used to detect degradation that could cause the
response time to exceed the total allowance. The total time response
allowance for each function bounds all degradation that cannot be
detected by periodic surveillance. Implementation of the proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
This change does not affect the total system response time
assumed in the safety analysis. The periodic system response time
verification method for the process protection racks, Nuclear
Instrumentation, and logic systems is modified to allow the use of
actual test data or engineering data. The method of verification
still provides assurance that the total system response is within
that defined in the safety analysis, since calibration tests will
continue to be performed and may be used to detect any degradation
which might cause the system response time to exceed the total
allowance. The total response time allowance for each function
bounds all degradation that cannot be detected by periodic
surveillance. Based on the above, it is concluded that the proposed
change does not result in a significant reduction in margin with
respect to plant safety.
Pursuant to 10 CFR 50.91, the preceding analysis provides a
determination that the proposed Technical Specifications change
poses no significant hazard as delineated by 10 CFR 50.92.
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.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of amendment request: August 12, 2002.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.8.2.3, ``Electrical Power
Systems, D.C. Distribution--Operating,'' TS 3.8.2.4, ``Electrical Power
Systems, D.C. Distribution--Shutdown,'' and TS 3.8.2.5, ``Electrical
Power Systems, D.C. Distribution Systems (Turbine Battery)--Operating''
to use standard technical specification terminology in order to provide
enhanced readability and usability. The proposed amendment would also
provide additional criteria for determining battery operability upon
restoration from a recharge or equalizing charge.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed Technical Specifications changes for relocation of
information which defines the operability of the D.C. electrical
power subsystems will not create any new failure modes, will not
cause an accident to occur, and will not result in any change in the
operation of accident mitigation equipment. Relocation of this
information will not have an adverse impact on any accident
initiators. Proper operation of the D.C. electrical power subsystems
will still be verified. As a result, the design basis accidents will
remain the same postulated events described in the Millstone Unit
No. 2 Final Safety Analysis Report, and the consequences of the
design basis accidents will remain the same. Therefore, the proposed
changes will not increase the probability or consequences of an
accident previously evaluated.
The proposed changes for deletion of redundant actions
requirements and reformatting of surveillance requirements
associated with the D.C. electrical power subsystems will not cause
an accident to occur and will not result in any change in the
operation of associated accident mitigation equipment. The proposed
changes will not have an adverse impact on any accident initiators.
Proper operation of the D.C. electrical power subsystems will still
be verified. As a result, the design basis accidents will remain the
same postulated events described in the Millstone Unit No. 2 Final
Safety Analysis Report, and the consequences of the design basis
accidents will remain the same. Therefore, the proposed changes will
not increase the probability or consequences of an accident
previously evaluated.
The proposed changes to the surveillance requirements for the
D.C. electrical power subsystems to add additional criteria relating
to physical damage or deterioration and its impact on battery
performance do not affect any existing accident initiators or
precursors. The proposed changes will not create any adverse
interactions with other systems that could result in initiation of a
design basis accident. Proper operation of the D.C. electrical power
subsystems batteries will still be verified. As a result, the design
basis accidents will remain the same postulated events described in
the Millstone Unit No. 2 Final Safety Analysis Report, and the
consequences of the design basis accidents will remain the same.
Therefore, the proposed changes will not increase the probability or
consequences of an accident previously evaluated.
The proposed changes to the surveillance requirements for the
D.C. electrical power subsystems to add additional criteria relating
to demonstrating battery operability following a recharge or
equalizing charge will not have an adverse affect on battery
operability. The proposed changes will not create any adverse
interactions with other systems that could result in initiation of a
design basis accident. Proper operation of the D.C. electrical power
subsystems batteries will still be verified. As a result, the design
basis accidents will remain the same postulated events described in
the Millstone Unit No. 2 Final Safety Analysis Report, and the
consequences of the design basis accidents will remain the same.
Therefore, the proposed changes will not increase the probability or
consequences of an accident previously evaluated.
[[Page 61678]]
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes do not create any new or different accident
initiators or precursors. The proposed changes do not create any new
failure modes for the components of the D.C. electrical power
subsystems and do not affect the interaction between the D.C.
electrical power subsystems and any other system. The proposed
changes do not alter the plant configuration (no new or different
type of equipment will be installed) or require any new or unusual
operator actions. The proposed changes do not alter the way any
structure, system, or component functions and do not alter the
manner in which the plant is operated. The components of the D.C.
electrical power subsystems will continue to function as before, and
will continue to be declared inoperable if their ability to perform
a safety function is impaired. Therefore, the proposed changes do
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes will not reduce the margin of safety since
they have no impact on any accident analysis assumption. The
proposed changes do not decrease the scope of equipment currently
required to be operable or subject to surveillance testing, nor do
the proposed changes affect any instrument setpoints or equipment
safety functions. The Technical Specifications will continue to
require that a battery be declared inoperable if physical damage or
abnormal deterioration of the cells, cell plates, or racks that
would degrade battery performance is observed. The proposed changes
do not alter the requirements of the Technical Specification with
respect to the capacity of any battery. The effectiveness of
Technical Specifications will be maintained since the changes will
not alter the operation of any component or system, nor will the
proposed changes affect any safety limits or safety system settings
which are credited in a facility accident analysis. Therefore, there
is no reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Section Chief: James W. Andersen, Acting.
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit No. 3, New London County, Connecticut
Date of amendment request: August 14, 2002.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) related to Containment
Systems. Specifically, the proposed changes would: (1) Add
clarification to TS 1.7 ``Definitions--Containment Integrity'' (2) add
clarifying information as well revise a portion of Surveillance
Requirement (SR) 4.6.1.1 associated with the affected section of TS
3.6.1.1 ``Containment Integrity;'' (3) revise TS 3.6.3, ``Containment
Isolation Valves,'' to make editorial changes, to add clarifying
information and to add an Action item that would increase the allowed
outage time (AOT) from 4 hours to 72 hours for Containment Isolation
Valves (CIVs) in closed systems, and (4) other changes that are
clarifying and/or administrative in nature. In addition, the TS Bases
would be revised to address the proposed changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed Technical Specification changes associated with
both containment integrity and CIVs that will remove ambiguity,
improve usability, and increase AOT for CIVs in closed systems, will
not cause an accident to occur. Operability requirements for
containment integrity and CIVs will remain the same. The ability of
the equipment associated with the proposed changes to mitigate the
design basis accidents will not be affected. The proposed Technical
Specification requirements are sufficient to ensure the required
accident mitigation equipment will be available and function
properly for design basis accident mitigation. The proposed allowed
outage time is reasonable and consistent with standard industry
guidelines to ensure the accident mitigation equipment will be
restored in a timely manner. In addition, the design basis accidents
will remain the same postulated events described in the Millstone
Unit No. 3 Final Safety Analysis Report, and the consequences of
those events will not be affected. Therefore, the proposed changes
will not increase the probability or consequences of an accident
previously evaluated.
The additional proposed changes to the Technical Specifications
(e.g., relocating information to the Bases, renumbering of
footnotes, renumbering a requirement) will not result in any
technical changes to the current requirements. Therefore, these
additional changes will not increase the probability or consequences
of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes to the Technical Specifications do not
impact any system or component that could cause an accident. The
proposed changes will not alter the plant configuration (no new or
different type of equipment will be installed) or require any
unusual operator actions. The proposed changes will not alter the
way any structure, system, or component functions, and will not
alter the manner in which the plant is operated. The response of the
plant and the operators following an accident will not be different.
In addition, the proposed changes do not introduce any new failure
modes. Therefore, the proposed changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed Technical Specification changes associated with
both containment integrity and CIVs that will remove ambiguity,
improve usability, and increase AOT for CIVs in closed systems, will
not cause an accident to occur. Operablity requirements for
containment integrity and CIVs will remain the same. The equipment
associated with the proposed Technical Specification changes will
continue to be able to mitigate the design basis accidents as
assumed in the safety analysis. The proposed allowed outage time is
reasonable and consistent with standard industry guidelines to
ensure the accident mitigation equipment will be restored in a
timely manner. In addition, the proposed changes will not affect
equipment design or operation, and there are no changes being made
to the Technical Specification required safety limits or safety
system settings. The proposed Technical Specification changes will
provide adequate control measures to ensure the accident mitigation
functions are maintained. Therefore, the proposed changes will not
result in a reduction in a margin of safety.
The additional proposed changes to the Technical Specifications
(e.g., relocating information to the Bases, renumbering of
footnotes, renumbering a requirement) will not result in any
technical changes to the current requirements. Therefore, these
additional changes will not result in a reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Section Chief: James W. Andersen, Acting.
[[Page 61679]]
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: August 15, 2002.
Description of amendment request: The proposed amendment would
revise the River Bend Station (River Bend or RBS) reactor vessel
surveillance program required by Title 10 of the Code of Federal
Regulations (10 CFR) part 50, appendix H, section IIIB.3. The change
will incorporate the Boiling Water Reactor Vessel & Internals Project
Integrated Surveillance Program into the RBS licensing basis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Pressure-temperature (P/T) limits (RBS Technical Specifications
Figure 3.4.11-1) are imposed on the reactor coolant system to ensure
that adequate safety margins against nonductile or rapidly
propagating failure exist during normal operation, anticipated
operational occurrences, and system hydrostatic tests. The P/T
limits are related to the nil-ductility reference temperature,
RTNDT, as described in ASME [American Society of
Mechanical Engineers Boiler and Pressure Vessel Code (Code)] Section
III, Appendix G. Changes in the fracture toughness properties of RPV
[reactor pressure vessel] beltline materials, resulting from the
neutron irradiation and the thermal environment, are monitored by a
surveillance program in compliance with the requirements of 10CFR50,
Appendix H. The effect of neutron fluence on the shift in the nil-
ductility reference temperature of pressure vessel steel is
predicted by methods given in RG [Regulatory Guide] 1.99, Rev[ision]
2.
River Bend's current P/T and Power Uprate limits were
established based on adjusted reference temperatures developed in
accordance with the procedures prescribed in RG 1.99, Rev 2,
Regulatory Position 1. Calculation of adjusted reference temperature
by these procedures includes a margin term to ensure conservative,
upper-bound values are used for the calculation of the P/T limits.
When permitted (two or more credible surveillance data sets
available), Regulatory Position 2 (or other NRC [U.S. Nuclear
Regulatory Commission]-approved) methods for determining adjusted
reference temperature will be followed.
This change is not related to any accidents previously
evaluated. This change will not affect P/T limits as given in RBS
Technical Specifications Figure 3.4.11-1 or USAR [Updated Safety
Analysis Report] Figures 5.3-4a and 5.3-4b. This change will not
affect any plant safety limits or limiting conditions of operation.
The proposed change will not affect reactor pressure vessel
performance as no physical changes are involved and RBS vessel P/T
limits will remain conservative in accordance with Reg[ulatory]
Guide 1.99, Rev 2 requirements. The proposed change will not cause
the reactor pressure vessel or interfacing systems to be operated
outside of their design or testing limits. Also, the proposed change
will not alter any assumptions previously made in evaluating the
radiological consequences of accidents. Therefore, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the RBS license basis to reflect
participation in the ISP [Integrated Surveillance Program]. This
proposed change does not involve a modification of the design of
plant structures, systems, or components. The proposed change will
not impact the manner in which the plant is operated as plant
operating and testing procedures will not be affected by the change.
The proposed change will not degrade the reliability of structures,
systems, or components important to safety as equipment protection
features will not be deleted or modified, equipment redundancy or
independence will not be reduced, supporting system performance will
not be downgraded, the frequency of operation of equipment will not
be increased, and increased or more severe testing of equipment will
not be imposed. No new accident types or failure modes will be
introduced as a result of the proposed change. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from that previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
As stated in the River Bend SER [Safety Evaluation Report],
``Appendices G and H of 10CFR50 describe the conditions that require
pressure-temperature limits and provide the general bases for these
limits. These appendices specifically require that pressure-
temperature limits must provide safety margins at least as great as
those recommended in the ASME Code, Section III, Appendix G. * * *
Until the results from the reactor vessel surveillance program
become available, the staff will use Regulatory Guide (RG) 1.99,
Revision 1 [now Revision 2], to predict the amount of neutron
irradiation damage. * * * The use of operating limits based on these
criteria--as defined by applicable regulations, codes, and
standards--will provide reasonable assurance that nonductile or
rapidly propagating failure will not occur, and will constitute an
acceptable basis for satisfying the applicable requirements of
General Design Criteria (GDC) 31.''
Bases for RBS Technical Specification 3.4.11 states: ``The P/T
limits are not derived from Design Basis Accident (DBA) analyses.
They are prescribed during normal operation to avoid encountering
pressure, temperature, and temperature rate of change conditions
that might cause undetected flaws to propagate and cause nonductile
failure of the RCPB [Reactor Coolant Pressure Boundary], a condition
that is unanalyzed. * * * Since the P/T limits are not derived from
any DBA, there are no acceptance limits related to the P/T limits.
Rather, the P/T limits are acceptance limits themselves since they
preclude operation in an unanalyzed condition.''
The proposed change will not affect any safety limits, limiting
safety system settings, or limiting conditions of operation. The
proposed change does not represent a change in initial conditions,
or in a system response time, or in any other parameter affecting
the course of an accident analysis supporting the Bases of any
Technical Specification. The proposed change does not involve
revision of the P/T limits but rather a revision to the surveillance
capsule withdrawal schedule. The current P/T limits were established
based on adjusted reference temperatures for vessel beltline
materials calculated in accordance with Regulatory Position 1 of RG
1.99, Rev 2. P/T limits will continue to be revised as necessary for
changes in adjusted reference temperature due to changes in fluence
according to Regulatory Position 1 until two or more credible
surveillance data sets become available. When two or more credible
surveillance data sets become available, P/T limits will be revised
as prescribed by Regulatory Position 2 of RG 1.99, Rev 2, or other
NRC-approved guidance. Therefore, the proposed change does not
involve a significant reduction in any margins 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 Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: August 21, 2002.
Description of amendment request: The proposed amendment would
revise Surveillance Requirement (SR) 3.0.3 to extend the delay period,
before entering a Limiting Condition for Operation, following a missed
surveillance. The delay period would be extended from the current limit
of ``* * * up to 24 hours or up to the limit of the specified
Frequency, whichever is less'' to ``* * * up to 24 hours or up to the
limit of the
[[Page 61680]]
specified Frequency, whichever is greater.'' In addition, the following
requirement would be added to SR 3.0.3: ``A risk evaluation shall be
performed for any Surveillance delayed greater than 24 hours and the
risk impact shall be managed.''
The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice
of opportunity for comment in the Federal Register on June 14, 2001 (66
FR 32400), on possible amendments concerning missed surveillances,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on September 28, 2001 (66 FR
49714). The licensee affirmed the applicability of the following NSHC
determination in its application dated August 21, 2002.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC 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 proposed change relaxes the time allowed to perform a missed
surveillance. The time between surveillances is not an initiator of
any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The
equipment being tested is still required to be operable and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected. Any reduction
in confidence that a standby system might fail to perform its safety
function due to a missed surveillance is small and would not, in the
absence of other unrelated failures, lead to an increase in
consequences beyond those estimated by existing analyses. The
addition of a requirement to assess and manage the risk introduced
by the missed surveillance will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. A
missed surveillance will not, in and of itself, introduce new
failure modes or effects and any increased chance that a standby
system might fail to perform its safety function due to a missed
surveillance would not, in the absence of other unrelated failures,
lead to an accident beyond those previously evaluated. The addition
of a requirement to assess and manage the risk introduced by the
missed surveillance will further minimize possible concerns. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The extended time allowed to perform a missed surveillance does
not result in a significant reduction in the margin of safety. As
supported by the historical data, the likely outcome of any
surveillance is verification that the LCO [Limiting Condition for
Operation] is met. Failure to perform a surveillance within the
prescribed frequency does not cause equipment to become inoperable.
The only effect of the additional time allowed to perform a missed
surveillance on the margin of safety is the extension of the time
until inoperable equipment is discovered to be inoperable by the
missed surveillance. However, given the rare occurrence of
inoperable equipment, and the rare occurrence of a missed
surveillance, a missed surveillance on inoperable equipment would be
very unlikely. This must be balanced against the real risk of
manipulating the plant equipment or condition to perform the missed
surveillance. In addition, parallel trains and alternate equipment
are typically available to perform the safety function of the
equipment not tested. Thus, there is confidence that the equipment
can perform its assumed safety function.
Therefore, this change does not involve a significant reduction
in a margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
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 Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: May 14, 2002, as supplemented by letter
dated September 9, 2002. The May 14, 2002, application was originally
noticed in the Federal Register on July 23, 2002 (67 FR 48216).
Description of amendment request: The proposed amendment would
revise Surveillance Requirement (SR) 4.0.3 to extend the delay period,
before entering a Limiting Condition for Operation, following a missed
surveillance. The delay period would be extended from the current limit
of ``* * * up to 24 hours to permit the completion of the surveillance
when the allowable outage time limits of the ACTION requirements are
less than 24 hours'' to ``* * *up to 24 hours or up to the limit of the
specified interval, whichever is greater.'' In addition, the following
requirement would be added to SR 4.0.3: ``A risk evaluation shall be
performed for any Surveillance delayed greater than 24 hours and the
risk impact shall be managed.'' Also, the addition of a Bases Control
Program is proposed as Technical Specification (TS) 6.5.14,
clarifications are proposed for SR 4.0.1, and other minor changes are
proposed for SR 4.0.3, consistent with NUREG-1432, Revision 2,
``Standard Technical Specifications, Combustion Engineering Plants.''
The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice
of opportunity for comment in the Federal Register on June 14, 2001 (66
FR 32400), on possible amendments concerning missed surveillances,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on September 28, 2001 (66 FR
49714). The licensee affirmed the applicability of the model NSHC
determination in its application dated May 14, 2002, as supplemented by
letter dated September 9, 2002. The NRC staff has augmented the model
NSHC to address the ANO-2 plant-specific items regarding the addition
of a Bases Control Program, clarifications for SR 4.0.1, and other
minor changes for SR 4.0.3 (because the model NSHC assumes a plant's
TSs already have these improvements), as presented below.
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:
[[Page 61681]]
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change relaxes the time allowed to perform a missed
surveillance. The time between surveillances is not an initiator of
any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The
equipment being tested is still required to be operable and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected. Any reduction
in confidence that a standby system might fail to perform its safety
function due to a missed surveillance is small and would not, in the
absence of other unrelated failures, lead to an increase in
consequences beyond those estimated by existing analyses. The
addition of a requirement to assess and manage the risk introduced
by the missed surveillance will further minimize possible concerns.
The addition of a Bases Control Program formalizes a means for
processing changes to the Bases of the TSs and does not change the
meaning of any TS. The clarifications proposed for SR 4.0.1
regarding surveillances that are not met, do not change the current
intent or practice of the TSs. The other minor changes to SR 4.0.3
regarding the discovery of surveillances that were not performed,
address the delay time period and make other editorial changes that
do not change the current intent or practice of the TSs. As such,
none of these changes affects the initiator of any accident
previously evaluated nor the ability of safety systems to mitigate
any accident previously evaluated.
Therefore, the changes discussed above do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. A
missed surveillance will not, in and of itself, introduce new
failure modes or effects and any increased chance that a standby
system might fail to perform its safety function due to a missed
surveillance would not, in the absence of other unrelated failures,
lead to an accident beyond those previously evaluated. The addition
of a requirement to assess and manage the risk introduced by the
missed surveillance will further minimize possible concerns.
Likewise, formalizing a program to control changes to the Bases,
clarifying SR 4.0.1, and the other minor changes to SR 4.0.3, do not
change the meaning of any TS and thus do not involve a physical
alteration of the plant or change the methods governing normal plant
operation.
Therefore, the changes discussed above do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The extended time allowed to perform a missed surveillance does
not result in a significant reduction in the margin of safety. As
supported by the historical data, the likely outcome of any
surveillance is verification that the LCO [Limiting Condition for
Operation] is met. Failure to perform a surveillance within the
prescribed frequency does not cause equipment to become inoperable.
The only effect of the additional time allowed to perform a missed
surveillance on the margin of safety is the extension of the time
until inoperable equipment is discovered to be inoperable by the
missed surveillance. However, given the rare occurrence of
inoperable equipment, and the rare occurrence of a missed
surveillance, a missed surveillance on inoperable equipment would be
very unlikely. This must be balanced against the real risk of
manipulating the plant equipment or condition to perform the missed
surveillance. In addition, parallel trains and alternate equipment
are typically available to perform the safety function of the
equipment not tested. Thus, there is confidence that the equipment
can perform its assumed safety function.
Likewise, formalizing a program to control changes to the Bases,
clarifying SR 4.0.1, and the other minor changes to SR 4.0.3, do not
change the meaning of any TS and thus will not cause equipment that
is relied upon to perform a safety function, to become inoperable.
Therefore, the changes discussed above do not involve a
significant reduction in a margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff has reviewed the above 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: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of amendment request: August 7, 2002.
Description of amendment request: The proposed amendment would
revise the Limiting Condition for Operation (LCO), the associated
Conditions and Required Actions of TS 3.7.1, and the values in Table
3.7.1-1. The proposed changes would revise the LCO by requiring five
MSSVs per steam generator to be operable consistent with the accident
analyses assumptions. The proposed change would modify the associated
Required Actions of TS 3.7.1 by adding a requirement to reduce the
Power Range Neutron Flux--High reactor trip setpoint when one or more
steam generators with one or more MSSVs are inoperable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change adds a requirement to appropriately reduce
the Power Range Neutron Flux--High reactor trip setpoint when one or
more steam generators with one or more MSSVs are inoperable. The
proposed TS change does not affect the design of the MSSV or
increase the likelihood of MSSV failures. Reducing the Power Range
Neutron Flux--High reactor trip setpoint does not affect initiators
of any accident sequence analyzed in the Byron/Braidwood Stations'
Updated Final Safety Analysis Report (UFSAR). Therefore, the
probability of occurrence of a previously evaluated accident is not
increased.
The design basis for the MSSVs is to limit the secondary system
pressure to <= 110% of steam generator design pressure for any
Anticipated Operational Occurrence (AOO) or accident considered in
the Design Basis Accident (DBA) and transient analyses. If there are
inoperable MSSVs, it is necessary to limit the primary system power
during steady-state operation and Anticipated Operational
Occurrences (AOOs) to a value that does not result in exceeding the
combined steam flow capacity of the turbine (if available) and the
remaining operable MSSVs. It has been demonstrated that for those
events that challenge the relieving capacity of the MSSVs, i.e.,
decreased heat removal events resulting in a Reactor Coolant System
(RCS) heatup and reactivity insertion events, it is necessary to
limit the AOO by reducing the setpoint of the Power Range Neutron
Flux--High reactor trip function. For example, with one or more
MSSVs on one or more steam generators inoperable, during an RCS
heatup event (e.g., turbine trip) when the Moderator Temperature
Coefficient (MTC) is positive, the reactor power may increase above
the value assumed in the analysis at the start of the transient.
Likewise, a reactivity insertion event, such as an uncontrolled rod
cluster control assembly (RCCA) withdrawal from partial power level,
may result in an increase in reactor power that exceeds the combined
steam flow
[[Page 61682]]
capacity of the turbine and the remaining operable MSSVs. Thus, for
any number of inoperable MSSVs on one or more steam generators it is
necessary to prevent a power increase by lowering the Power Range
Neutron Flux--High reactor trip setpoint to an appropriate value.
This change will ensure that the consequences of previously
evaluated accidents remain bounding. Currently administrative
controls are in place to address the current non-conservative TS in
accordance with the direction provided in NRC Administrative Letter
98-10, ``Dispositioning of Technical Specifications that are
Insufficient to Assure Plant Safety.''
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?
The proposed change does not involve a physical alteration of
the units. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. The
design and operation of the MSSVs are unaffected by the proposed
change. The proposed change will not alter the manner in which
equipment operation is initiated, nor will the functional demands on
equipment be changed. No change is being made to procedures relied
upon to respond to off-normal events. As such, no new failure modes
are being introduced. The proposed change appropriately revises the
setpoints at which protective actions are initiated. The proposed
change also prevents operating the plant in a configuration that
could challenge the safety analyses limiting initial condition
assumptions, thereby ensuring previously evaluated accidents remain
bounding. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The primary purpose of the MSSVs is to provide overpressure
protection for the secondary system. The MSSVs must have sufficient
capacity to limit the secondary pressure to <= 110% of the steam
generator design pressure in order to meet the requirements of the
American Society of Mechanical Engineers (ASME) Boiler and Pressure
Vessel (B&PV) Code, Section III, ``Rules for Construction of Nuclear
Power Plant Components.'' The proposed change precludes operation in
a configuration that could challenge the design requirement of the
MSSVs by requiring a reduction in the Power Range Neutron Flux--High
reactor trip setpoint, in addition to a reduction in Thermal Power,
when one or more steam generators with one or more MSSVs are
inoperable. The maximum allowable power specified in TS Table 3.7.1-
1 was calculated using a simple heat balance calculation as
described in the attachment to NRC Information Notice 94-60,
``Potential Overpressurization of the Main Steam Safety System,''
dated August 22, 1994, assuming uprated power conditions with an
appropriate allowance for Nuclear Instrumentation System reactor
trip channel uncertainties. Precluding operation in a configuration
that could challenge the design requirement of the MSSVs and
appropriately revising the values in Table 3.7.1-1 preserves the
margin of safety. This change assures the design basis limit will
not be exceeded. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
Exelon Generation Company, LLC, Docket No. 50-171, Peach Bottom Atomic
Power Station, Unit 1, York County, Pennsylvania
Date of application for amendment: May 21, 2002.
Brief description of amendment: This proposed amendment will revise
the Peach Bottom Atomic Power Station, Unit 1, Technical Specifications
(TS) to: (1) delete License Condition C(4) to reflect satisfaction of
the minimum decommissioning trust fund amount at the time of transfer
of the Facility Operating License; 2) revise License Condition C(5)(d)
to reflect 30 days prior written notification to the Director of
Nuclear Material Safety and Safeguards before modification of the
decommissioning trust agreement in any material respect; 3) delete TS
2.1(B)3 and TS 2.4(b) to eliminate inconsistencies with reporting
requirements in Title 10 U.S. Code of Federal Regulations (10 CFR)
20.2202, 10 CFR 50.73, and 10 CFR 73.71; 4) revise TS 2.2 to refer to
the Facility Operating License; and 5) revise TS 2.3 to refer to the
radiological hazards associated with the facility.
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:
a. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed changes do not impact the SAFSTOR status of
Peach Bottom Atomic Power Station, Unit 1, or the design of any
plant system, structure, or component. These changes are
administrative in nature. They do not affect security at Unit 1 or
the potential of radioactive material being released. Therefore, the
proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
b. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The changes do not alter the plant configuration. These
changes are administrative in nature and do not alter assumptions
made in the safety analysis and licensing basis. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
c. Do the proposed changes involve a significant reduction in a
margin of safety?
No. These changes are administrative in nature. The changes will
not reduce a margin of safety because they have no impact on any
safety analysis assumptions. Therefore, the proposed changes will
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve 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: Claudia M. Craig.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: August 22, 2002.
Description of amendment request: The proposed change modifies the
required surveillance interval for calibration of the trip units
associated with the instrumentation channels of the Anticipated
Transient Without Scram-Recirculation Pump Trip (ATWS-RPT) system from
monthly to quarterly.
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 change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed TS [Technical Specification] change increases a STI
[surveillance test interval] for ATWS-RPT System actuation
[[Page 61683]]
instrumentation based on generic analyses completed by the Boiling
Water Reactor Owners' Group (BWROG). The NRC has reviewed and
approved these generic analyses and has concurred with the BWROG
that the proposed changes do not significantly affect the
probability of failure or availability of the affected
instrumentation systems. EGC [Exelon Generation Company, LLC] has
determined these studies are applicable to QCNPS [Quad Cities
Nuclear Power Station], Units 1 and 2.
TS requirements that govern operability or routine testing of
plant instruments are not assumed to be initiators of any analyzed
event because these instruments are intended to prevent, detect, or
mitigate accidents. Therefore, this change will not involve an
increase in the probability of occurrence of an accident previously
evaluated. Additionally, this change will not increase the
consequences of an accident previously evaluated because the
proposed change does not involve any physical changes to ATWS-RPT
System components or the manner in which the ATWS-RPT System is
operated. This change will not alter the operation of equipment
assumed to be available for the mitigation of accidents or
transients specified in the ATWS analysis contained in the QCNPS
Updated Final Safety Analysis Report (UFSAR). As justified and
approved in licensing topical reports endorsing extended AOTs
[allowed out-of-service times] and STIs, the proposed change
establishes or maintains adequate assurance that components are
operable when necessary for the prevention or mitigation of
accidents or transients, and that plant variables are maintained
within limits necessary to satisfy the assumptions for initial
conditions in the safety analyses. Furthermore, there will be no
change in the types or significant increase in the amounts of any
effluents released offsite. For these reasons, the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change does 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 changes to the
ATWS-RPT System or associated components, or the manner in which the
ATWS-RPT System functions. Therefore, this change will not create
the possibility of a new or different kind of accident from any
accident previously evaluated. There is no change being made to the
parameters within which the plant is operated. There are no
setpoints at which protective or mitigative actions are initiated
that are affected by the proposed change. This proposed change will
not alter the manner in which equipment operation is initiated nor
will the function demands on credited equipment be changed. The
change in methods governing normal plant operation is consistent
with the current ATWS analysis assumptions specified in the UFSAR.
Therefore, this change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Does the proposed change involve a significant reduction in a
margin of safety?
Margins of safety are established in the design of components,
the configuration of components to meet certain performance
parameters, and in the establishment of setpoints to initiate alarms
or actions. The proposed change increases a STI for ATWS-RPT System
actuation instrumentation based on generic analyses completed by the
BWROG. The analyses determined that there is no significant change
in the availability and/or reliability of ATWS-RPT instrumentation
as a result of the proposed change in STI. The extended STI does not
result in significant changes in the probability of ATWS-RPT
instrument failure. Furthermore, the proposed change will not reduce
the probability of test-induced ATWS-RPT transients and equipment
failures. Therefore, it is concluded that the proposed change will
not result in a reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: August 26, 2002.
Description of amendment request: The proposed amendment would
revise Surveillance Requirement (SR) 3.0.3 to extend the delay period
before entering a Limiting Condition for Operation (LCO) following a
missed surveillance. The delay period would be extended from the
current limit of ``* * * up to 24 hours or up to the limit of the
specified Frequency, whichever is less'' to ``* * * up to 24 hours or
up to the limit of the specified Frequency, whichever is greater.'' In
addition, the following requirement would be added to SR 3.0.3: ``A
risk evaluation shall be performed for any Surveillance delayed greater
than 24 hours and the risk impact shall be managed.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments
concerning missed surveillances, including a model safety evaluation
and model no significant hazards consideration (NSHC) determination,
using the consolidated line-item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on September 28, 2001 (66 FR 49714). The licensee affirmed the
applicability of the following NSHC determination in its application
dated August 26, 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 proposed change relaxes the time allowed to perform a missed
surveillance. The time between surveillances is not an initiator of
any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The
equipment being tested is still required to be operable and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected. Any reduction
in confidence that a standby system might fail to perform its safety
function due to a missed surveillance is small and would not, in the
absence of other unrelated failures, lead to an increase in
consequences beyond those estimated by existing analyses. The
addition of a requirement to assess and manage the risk introduced
by the missed surveillance will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. A
missed surveillance will not, in and of itself, introduce new
failure modes or effects and any increased chance that a standby
system might fail to perform its safety function due to a missed
surveillance would not, in the absence of other unrelated failures,
lead to an accident beyond those previously evaluated. The addition
of a requirement to assess and manage the risk introduced by the
missed surveillance will further minimize possible concerns. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The extended time allowed to perform a missed surveillance does
not result in a
[[Page 61684]]
significant reduction in the margin of safety. As supported by the
historical data, the likely outcome of any surveillance is
verification that the LCO is met. Failure to perform a surveillance
within the prescribed frequency does not cause equipment to become
inoperable. The only effect of the additional time allowed to
perform a missed surveillance on the margin of safety is the
extension of the time until inoperable equipment is discovered to be
inoperable by the missed surveillance. However, given the rare
occurrence of inoperable equipment, and the rare occurrence of a
missed surveillance, a missed surveillance on inoperable equipment
would be very unlikely. This must be balanced against the real risk
of manipulating the plant equipment or condition to perform the
missed surveillance. In addition, parallel trains and alternate
equipment are typically available to perform the safety function of
the equipment not tested. Thus, there is confidence that the
equipment can perform its assumed safety function.
Therefore, this change does not involve a significant reduction
in a margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff has reviewed the 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: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: L. Raghavan.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: August 20, 2002.
Description of amendment request: The proposed change will modify
action statements and surveillance requirements associated with the
diesel generators and make various editorial 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:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not affect the operational limits or the
physical design of the emergency diesel generators.
The emergency diesel generator system is not an accident
initiator. The proposed changes will minimize unnecessary testing
that can result in accelerated degradation and will reduce the
burden on plant operating personnel while continuing to ensure
emergency diesel generator reliability. The editorial and
administrative changes do not change the intent of any Technical
Specification requirement.
Since the proposed changes do not affect any accident initiator
and since the emergency diesel generators will remain capable of
performing their design function, the proposed change does not
involve a significant increase in the probability or off-site and
on-site radiological consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not affect the operational limits or the
physical design of the emergency diesel generators. The diesel
generators will remain capable of performing their design function.
No new failure mechanisms, malfunctions, or accident initiators are
being introduced by the proposed changes. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not affect the operational limits or the
physical design of the emergency diesel generators. The diesel
generators will remain capable of performing their design function.
Unnecessary testing that can result in accelerated degradation will
be minimized by the proposed changes. Therefore, the proposed change
does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Andersen, Acting.
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: August 9, 2002.
Description of amendment request: The proposed amendments would
incorporate the Boiling Water Reactor Vessel and Internals Project
(BWRVIP) Integrated Surveillance Program for the surveillance of the
Plant Hatch material capsules. The schedule for removal of the capsules
is provided in the Units 1 and 2 Final Safety Analysis Reports. The
proposed amendment is consistent with the NRC's Regulatory Issue
Summary 2002-05, ``NRC Approval of Boiling Water Reactor Pressure
Vessel Integrated Surveillance Program,'' dated April 8, 2002 (ADAMS
Accession No. ML020660522).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change to the material surveillance program will
involve implementing the BWRVIP Integrated Surveillance Program
(ISP). The purpose of the program is to monitor the reactor pressure
vessel beltline materials for neutron embrittlement. The existing
program for Hatch Units 1 and 2 includes removal and evaluation of
existing material capsules in the Hatch Unit 1 and 2 Reactor
vessels. The ISP combines all the individual surveillance programs
for participating U.S. BWRs into a single integrated program. To
insure the program is adequate, similar heats of materials are used
to represent the limiting materials of the RPVs. A test matrix was
developed to identify the specimens that best meet the needs of each
BWR, including the Hatch units. The material associations for the
ISP were chosen to best represent the limiting plate and weld
materials for each plant using specimens from the entire BWR fleet.
As a result, the Plant Hatch RPVs [reactor pressure vessels] will be
adequately monitored for neutron embrittlement and thus the
probability or consequences of RPV embrittlement are not
significantly increased.
Implementing the ISP does not affect the assumptions of any
previously evaluated accident, neither does it affect any of the
systems designed for the prevention or mitigation of previously
evaluated accidents. Therefore, their consequences are not
significantly increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated.
Implementing the ISP will not affect the operation of any plant
system designed for the prevention or mitigation of accidents. As a
result, no new modes of operation are introduced which may result in
the need to consider a new type of event. As described above in the
answer to question 1, the ISP will continue to adequately
monitor the RPV materials; therefore, the possibility of an RPV
embrittlement event is not created.
3. Does the proposed change involve a significant decrease in
the margin of safety.
[[Page 61685]]
The ISP will use materials that adequately represent a
particular RPV, including Plant Hatch. A test matrix, as provided in
BWRVIP-86: [``]BWR Vessel and Internals Project, BWR Integrated
Surveillance Program Implementation Plan,'' includes representative
materials from other plants to be used for the Hatch Units. A
representative material is a plate or weld that is selected from
among all the existing plant surveillance programs to represent the
corresponding limiting plate or weld material in a plant. The choice
of material considers chemistry, heat number, fabricator and the
welding process. These are factors that determine the best
representative material. As a result, the Hatch RPV will be
adequately monitored for radiation embrittlement and the margin of
safety is not significantly reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: John A. Nakoski.
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 19, 2002.
Description of amendment request: The proposed amendment revises
Technical Specification (TS) Section 3/4.3.2, ``Engineered Safety
Features Actuation System Instrumentation.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Pursuant to 10 CFR 50.91, this analysis provides a determination
that the proposed change to the Technical Specifications described
previously, does not involve any significant hazards consideration
as defined in 10 CFR 50.92, as described below:
[(1)] Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This change to the Technical Specifications will not result in a
condition where the design, material, and construction standards
that were applicable prior to the change are altered. The same ESFAS
[engineered safety features actuation system] instrumentation will
be used and the same ESFAS system reliability is expected. The
proposed change will not modify any system interface or function and
could not increase the likelihood of an accident because these
events are independent of this change. The proposed activity will
not change, degrade, or alter any assumptions previously made in
evaluating the radiological consequences of an accident described in
the safety analysis report.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[(2)] Does the proposed change create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change will not alter the performance of the ESFAS
mitigation systems assumed in the plant safety analysis. Changing
the interval for periodically verifying ESFAS slave relays (assuring
equipment operability) will not create any new accident initiators
or scenarios. Only the testing frequency is changed. No physical
changes will be made to the Solid State Protection System or the ESF
Actuation System as a result of this change.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
[(3)] Does the proposed change involve a significant reduction
in a margin of safety?
Response: No.
The proposed change will not affect the total ESFAS response
assumed in the safety analysis because the reliability of the slave
relays will not be significantly affected by the increased
surveillance interval. The relays have demonstrated a high
reliability and insensitivity to short term wear and aging effects.
The overall reliability, redundancy, and diversity assumed available
for the protection and mitigation of accident and transient
conditions is unaffected by this proposed Technical Specification
change.
Therefore, the proposed change does not involve a reduction in a
margin of safety.
Based on the above safety evaluation, the South Texas Project
concludes that the change proposed by this License Amendment Request
satisfies the no significant hazards consideration standards of 10
CFR 50.92(c) and, accordingly, a finding of no significant hazards
is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis, &
Bockius, 1111 Pennsylvania NW., Washington, DC 20004.
NRC Section Chief: Robert A. Gramm.
STP Nuclear Operating Company (STPNOC), Docket Nos. 50-498 and 50-499,
South Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: August 20, 2002.
Description of amendment request: The proposed amendment would
delete the Appendix C of the Operating License, regarding antitrust
conditions.
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:
STPNOC has determined whether a significant hazards
consideration is involved with the proposed amendment by focusing on
the three criteria set forth in 10 CFR 50.92 as discussed below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This request involves an administrative change only. The
Operating Licenses are being changed to remove unnecessary and
outdated antitrust conditions. No actual plant equipment or accident
analyses will be affected by the proposed changes. Therefore, this
request will have no impact on the probability or consequences of
any type of accident: new, different, or previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This request involves an administrative change only. The
Operating Licenses are being changed to remove unnecessary and
outdated antitrust conditions. No actual plant equipment or accident
analyses will be affected by the proposed change and no failure
modes not bounded by previously evaluated accidents will be created.
Therefore, this request will have no impact on the possibility of
any type of accident: new, different, or previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel and fuel cladding, Reactor
Coolant System pressure boundary, and containment structure) to
limit the level of radiation dose to the public. This request
involves an administrative change only. The Operating Licenses are
being changed to remove unnecessary and outdated antitrust
conditions.
No actual plant equipment or accident analyses will be affected
by the proposed change. Additionally, the proposed change will not
relax any criteria used to establish safety limits, safety systems
settings, or any limiting conditions of operations. Therefore, this
request will not impact [a] margin of safety.
Based on the above, STPNOC concludes that the proposed amendment
involves no significant hazards consideration under the criteria set
forth in 10 CFR 50.92 and, accordingly, a finding of ``no
significant hazards consideration'' is justified.
[[Page 61686]]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis, &
Bockius, 1111 Pennsylvania NW., Washington, DC 20004.
NRC Section Chief: Robert A. Gramm.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: August 21, 2002.
Description of amendment request: The proposed amendment revises
Technical Specifications (TS) 3/4.4.1.4.2 and 3/4.9.1.3 to delete the
specific reference to the valves required to be secured to isolate
uncontrolled boron dilution flow paths in MODE 5 with the loops not
filled and in MODE 6.
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:
STPNOC has evaluated whether or not a significant hazards
consideration is involved with the proposed amendment by focusing on
the three standards set forth in 10 CFR 50.92, ``Issuance of
amendment,'' as discussed below.
[(1)] Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
There is no technical change in the requirements imposed by the
Technical Specifications. The proposed changes to replace the TS
reference to the specific valves to be used to isolate boron
dilution flow paths with new Technical Specification requirements to
assure the flow paths are secured provides the same level of
assurance that the boron dilution event will be precluded.
[(2)] Does the proposed change create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change allows alternate, equally effective,
locations where the potential boron dilution flow paths can be
isolated to preclude an uncontrolled boron dilution event in MODE 5
with the loops not filled and in MODE 6. Consequently, the
possibility of the dilution event is unchanged. The proposed change
does not otherwise alter how the plant is operated or change its
design basis so that the possibility of a new accident is not
created.
[(3)] Does the proposed change involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes to replace the TS reference to the specific
valves to be used to isolate boron dilution flow paths with new
Technical Specification requirements to assure the flow paths are
secured provides the same level of assurance that the boron dilution
event will be precluded.
Based upon the analysis provided herein, the proposed amendments
do not involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis, &
Bockius, 1111 Pennsylvania NW., Washington, DC 20004.
NRC Section Chief: Robert A. Gramm.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: August 16, 2002.
Description of amendment request: The amendment would revise
Technical Specification 3.6.3, ``Containment Isolation Valves,'' by (1)
deleting the Note and adding the acronym ``(CIV)'' for containment
isolation valve in Condition A of the Actions for the Limiting
Condition for Operation, (2) revising the Completion Time for Required
Condition A.1 from 4 hours to as much as 7 days depending on the
category of the CIVs, (3) deleting Condition C, and (4) renumbering the
later Conditions D and E. The proposed amendment is based on Topical
Report WCAP-15791-P, ``Risk-Informed Evaluation of Extensions to
Containment Isolation Valve Completion Times.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to the Completion Times do not change the
response of the plant to any accidents and have an insignificant
impact on the reliability of the containment isolation valves. The
containment isolation valves will remain highly reliable and the
proposed changes will not result in a significant increase in the
risk of plant operation. This is demonstrated by showing that the
impact on plant safety as measured by the large early release
frequency (LERF) and incremental conditional large early release
probabilities (ICLERP) is acceptable. These changes are consistent
with the acceptance criteria in [the risk-informed] Regulatory
Guides 1.174 and 1.177. Therefore, since the containment isolation
valves will continue to perform their [safety] functions with high
reliability as originally assumed and the increase in risk as
measured by LERF and ICLERP is acceptable, there will not be a
significant increase in the consequences of any accidents.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, or
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not alter or
prevent the ability of structures, systems, and components (SSCs)
from performing their intended [safety] function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed changes do not affect the source term,
containment isolation, or radiological release assumptions used in
evaluating the radiological consequences of an accident previously
evaluated. Further, the proposed changes do not increase the types
or amounts of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures. The proposed changes are consistent with the
safety analysis assumptions and resultant consequences [in Chapter
15, ``Accident Analysis,'' of the Updated Final Safety Analysis
Report (USAR) for the plant].
Therefore, it is concluded that this change does not increase
the probability of occurrence of a malfunction of equipment
important to safety.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not result in a change in the manner in
which the containment isolation valves provide plant protection.
There are no design changes associated with the proposed changes.
The changes to Completion Times do not change any existing accident
scenarios, nor create any new or different accident scenarios.
The changes do not involve a physical alteration of the plant
(i.e., no new or different type of equipment will be installed) or a
change in the methods governing normal plant operation. In addition,
the changes do not impose any new or different requirements or
eliminate any existing requirements. The changes do not alter
assumptions made in the safety analysis. The proposed changes are
consistent with the safety analysis assumptions and current plant
operating practice.
Therefore, the possibility of a new or different malfunction of
safety related equipment is not created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
[[Page 61687]]
operation are determined. The safety analysis acceptance criteria
are not impacted by these changes. The proposed changes will not
result in plant operation in a configuration outside the design
basis. The calculated impact on risk is insignificant and is
consistent with the acceptance criteria contained in Regulatory
Guides 1.174 and 1.177.
Therefore, it is concluded that this 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: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit No. 3, New London County, Connecticut
Date of amendment request: July 19, 2002.
Brief description of amendment request: The proposed amendment
would revise Technical Specification Surveillance Requirement (SR)
4.0.3 to extend the delay period, before entering a Limiting Condition
for Operation, following a missed surveillance. The delay period would
be extended from the current limit of ``* * * up to 24 hours'' to ``* *
* up to 24 hours or up to the limit of the specified surveillance
interval, whichever is greater.'' In addition, the following
requirement would be added to SR 4.0.3: ``A risk evaluation shall be
performed for any surveillance delayed greater than 24 hours and the
risk impact shall be managed.'' The proposed amendment would also make
administrative changes to SRs 4.01 and 4.03 to be consistent with
NUREG-1431, Revision 2.
Date of publication of individual notice in Federal Register:
September 4, 2002 (67 FR 56604).
Expiration date of individual notice: October 4, 2002.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management Systems (ADAMS) Public Electronic
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/
reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by e-mail to [email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: September 19, 2001, as
supplemented on January 17 and July 1, 2002.
Brief description of amendment: The amendment revises Technical
Specifications Subsections 3.5.A.5.b and c, concerning operability of
suppression chamber-to-drywell vacuum breakers.
Date of Issuance: September 11, 2002.
Effective date: As of the date of issuance, to be implemented
within 30 days of issuance.
Amendment No.: 230.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 20, 2001 (66
FR 65749). The January 17 and July 1, 2002, letters provided clarifying
information within the scope of the original application and did not
change the staff's initial proposed no significant hazards
consideration determination. The Commission's related evaluation of
this amendment is contained in a Safety Evaluation dated September 11,
2002.
No significant hazards consideration comments received: No.
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: September 10, 2001.
Brief description of amendment: The amendment revised the
requirements in Technical Specifications, Sections 3.4.A.7.c and
3.4.A.8.c, changing confirmation of operability of core spray pumps and
system components from testing to verification.
Date of Issuance: September 10, 2002.
Effective date: As of the date of issuance, to be implemented
within 30 days of issuance.
Amendment No.: 231.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 5, 2002 (67 FR
10008). The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated September 10, 2002.
No significant hazards consideration comments received: No.
[[Page 61688]]
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: August 1, 2001, as supplemented
on June 19 and September 9, 2002.
Brief description of amendment: The amendment revised Technical
Specifications Section 6.3, ``Facility Staff Qualifications,'' deletes
Section 6.4, ``Training,'' and revises the Table of Contents to reflect
deletion of Section 6.4. These changes reflect updating of requirements
that had been outdated based on licensed operator training programs
being accredited by the Institute of Nuclear Power Operations, and
promulgation of applicable regulations.
Date of Issuance: September 18, 2002.
Effective date: September 18, 2002, and shall be implemented within
30 days of issuance.
Amendment No.: 232.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 31, 2001 (66 FR
55009). The June 19 and September 9, 2002, letters provided clarifying
information within the scope of the original application and did not
change the staff?s initial proposed no significant hazards
consideration determination. The Commission's related evaluation of
this amendment is contained in a Safety Evaluation dated September 18,
2002.
No significant hazards consideration comments received: No.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: May 15, 2002, as supplemented
by letter dated August 29, 2002.
Brief description of amendments: The amendments revise Limiting
Condition for Operation (LCO) 3.9.3, ``Containment Penetrations.'' The
amendments would (1) modify the requirement in LCO 3.9.3.b that one
door in each air lock is closed by adding the words ``capable of
being'' before the word ``closed'' and (2) add a note to LCO 3.9.3
stating that containment penetration flow paths providing direct access
from the containment to the outside atmosphere may be unisolated under
administrative controls. The amendments would allow the containment air
lock and other penetrations that provide direct access to the outside
atmosphere to be open during core alterations or movement of irradiated
fuel assemblies within containment.
Date of issuance: September 11, 2002.
Effective date: September 11, 2002, and shall be implemented within
60 days of the date of issuance, including completing the changes to
the Technical Specification Bases, as described in the licensee's
letters of May 15 and August 29, 2002.
Amendment Nos.: Unit 1--144, Unit 2--144, Unit 3--144.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 25, 2002 (67 FR
42816). The Commission's related evaluation of the amendments are
contained in a Safety Evaluation dated September 11, 2002.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: March 26, 2002, as supplemented
June 19 and August 8, 2002.
Brief description of amendment: This amendment extends the 10-year
performance-based Type A test interval on a one-time basis to require
the performance of a Type A test within 12.1 years from the last test,
which was performed on April 9, 1992.
Date of issuance: September 16, 2002.
Effective date: September 16, 2002.
Amendment No.: 193.
Facility Operating License No. DPR-23: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 28, 2002 (67 FR
36928). The June 19, and August 8, 2002, supplements contained
clarifying information only, and did not change the initial proposed no
significant hazards consideration determination or expand the scope of
the initial application. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated September 16, 2002.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: February 21, 2002, as
supplemented May 14 and August 2, 2002.
Brief description of amendment: The amendment modifies the
containment vessel spray nozzle testing frequency from testing every
``10 years'' to testing ``following activities which could result in
nozzle blockage.''
Date of issuance: September 19, 2002.
Effective date: September 19, 2002.
Amendment No.: 194.
Facility Operating License No. DPR-23: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 30, 2002 (67 FR
21285). The May 14 and August 2, 2002, supplements contained clarifying
information only and did not change the initial proposed no significant
hazards consideration determination or expand the scope of the initial
application. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 19, 2002.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket No. 50-400, Shearon Harris
Nuclear Plant, Unit 1, Wake and Chatham Counties, North Carolina
Date of application for amendment: July 8, 2002.
Brief Description of amendment: The amendment deleted the level
value in 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.''
Date of issuance: September 12, 2002.
Effective date: As of date of issuance and shall be implemented
within 60 days from date of issuance.
Amendment No.: 111.
Facility Operating License No. NPF-63: Amendment changes the
Technical Specifications.
Date of initial notice in Federal Register: August 6, 2002 (67 FR
50950). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 12, 2002.
No significant hazards consideration comments received: No.
Consumers Energy Company, Docket No. 50-155, Big Rock Point Nuclear
Plant, Charlevoix County, Michigan
Date of amendment request: June 11, 2002, as supplemented by letter
dated July 3, 2002.
Brief description of amendment: The amendment revises Defueled
Technical Specification (DTS) Section 5.2, ``Storage and Inspection of
Spent Fuel,'' and DTS Section 6.6.2.9, ``Spent Fuel Pool Water
Chemistry Program,'' by adding applicability statements that specify
that these specifications apply
[[Page 61689]]
only when irradiated fuel is stored in the spent fuel pool.
Date of issuance: September 11, 2002.
Effective date: The license amendment is effective as of the date
of issuance and shall be implemented within 45 days from the date of
issuance.
Amendment No.: 124.
Facility Operating License No. DPR-6: The amendment revised the
Defueled Technical Specifications.
Date of initial notice in Federal Register: July 9, 2002 (67 FR
45562). The July 3, 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 September 11, 2002.
No significant hazards considerations comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: May 23, 2002.
Brief description of amendment: The amendment deletes Technical
Specification 5.5.3, ``Post Accident Sampling System (PASS),'' and
thereby eliminates the requirements to have and maintain the PASS at
Fermi 2.
Date of issuance: September 5, 2002.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 150.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: June 25, 2002 (67 FR
42816). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 5, 2002.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423,
Millstone Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: June 6, 1998; April 5, 1999;
April 7, April 19, July 31, and September 28, 2000; March 19, June 11,
September 21, and December 20, 2001.
Brief description of amendment: The amendment revises the Millstone
Power Station, Unit No. 3 licensing basis related to operation of the
supplementary leak collection and release system after a postulated
accident. Specifically, the proposed revision to the Final Safety
Analysis Report (FSAR) would address: (1) The manual actions required
to trip the non-safety grade fans and the time requirements for control
room ventilation realignment, and (2) the input assumptions and results
of the loss-of-coolant accident/control rod ejection accident analyses.
Date of issuance: September 16, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 211.
Facility Operating License No. NPF-49: Amendment revised the FSAR.
Date of initial notice in Federal Register: July 1, 1998 (63 FR
35992). The April 5, 1999; April 7, April 19, July 31, and September
28, 2001; March 19, June 11, September 21, and December 20, 2001,
letters provide clarifying information that was within the scope of the
original application and did not change the staff's proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated September 16, 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 amendments: December 7, 2001, as
supplemented by letter dated July 22, 2002.
Brief description of amendments: The amendments revise the
Technical Specifications (TS) to permit implementation of containment
local leakage rate testing addressed by 10 CFR Part 50, Appendix J,
Option B, and to reference Regulatory Guide 1.163, ``Performance-Based
Containment Leak Test Program,'' dated September 1995. In addition, the
TS are revised regarding soap bubble testing and leak testing of
containment purge valves with resilient seals for upper and lower
compartments and instrument rooms.
Date of issuance: September 4, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 207 & 188.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 26, 2001 (67
FR 66464). The supplement dated July 22, 2002, provided clarifying
information that did not change the scope of the December 7, 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 September 4, 2002.
No significant hazards consideration comments received: No.
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 application of amendments: July 11, 2002.
Brief description of amendments: The amendments revised the
Technical Specifications to incorporate several administrative changes.
Date of Issuance: September 5, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 328, 328 & 329.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 6, 2002 (67 FR
50951). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 5, 2002.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear
Generating Unit No. 2, Westchester County, New York
Date of application for amendment: January 8, 2002, as supplemented
on August 22, 2002.
Brief description of amendment: The amendment revised Technical
Specifications Section 3.7.C, ``Gas Turbine Generators,'' and Section
4.6, ``Emergency Power System Periodic Tests,'' to change the minimum
amount of fuel oil required to be stored from 54,200 gallons to 94,870
gallons. The amendment also revised the minimum electrical output of
the gas turbine generator that is required to be tested monthly to 2000
kilowatts from the previous value of 750 kilowatts.
Date of issuance: September 18, 2002.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 233.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 5, 2002 (67 FR
10012). The August 22, 2002, letter provided clarifying information
that did not
[[Page 61690]]
change the initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 18, 2002.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: June 7, 2002, supplemented July
17, 2002.
Brief description of amendment: The amendment changes the Technical
Specifications to allow relaxation of secondary containment operability
requirements while handling irradiated fuel in the secondary
containment. The amendment replaces the current accident source term
use in selected design basis radiological analyses with an alternative
source term pursuant to 10 CFR 50.67, ``Accident Source Term.''
Date of issuance: September 12, 2002.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 276.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 9, 2002 (67 FR
45568). The July 17, 2002, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated September 12, 2002.
No significant hazards consideration comments received: No.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of application for amendment: March 19, 2002, as supplemented
on June 4, July 16 and 24, August 22 and September 4, 2002.
Brief description of amendment: The amendment revises the technical
specifications to reflect the removal of the automatic reactor scram
and main steam isolation valve closure functions of the main steam line
radiation monitors (MSLRM). An explicit requirement for periodic
functional test and calibration of the MSLRM is added to maintain
operability of the mechanical vacuum pump trip function.
Date of Issuance: September 18, 2002.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 212.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 9, 2002 (67 FR
45573). The July 16 and 24, August 22, and September 4, 2002,
supplements were within the scope of the original application and did
not change the staff's proposed no significant hazards consideration
determination. The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated September 18, 2002.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station (GGNS), Unit 1, Claiborne
County, Mississippi
Date of application for amendment: November 15, 2001, as
supplemented by letters dated March 1 and June 19, 2002.
Brief description of amendment: This amendment revises the GGNS
Unit 1 Technical Specification Surveillance Requirements (SRs)
pertaining to testing of the standby emergency diesel generators (DGs)
to allow DG testing during reactor operation. The change removes the
restriction associated with these SRs that prohibits conducting the
required testing of the DGs during reactor operating Modes 1, 2, or 3.
Date of issuance: September 5, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No: 153.
Facility Operating License No. NPF-29: The amendment revises the
Technical Specifications and Surveillance Requirements.
Date of initial notice in Federal Register: December 26, 2001 (66
FR 66464). The supplemental letters dated March 1 and June 19, 2002,
provided clarifying information that did not change the scope of
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 September 5, 2002.
No significant hazards consideration comments received: None.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: February 25, 2002, as
supplemented by letters dated August 16 and 22, 2002.
Brief description of amendment: This amendment adds a new Technical
Specification 3.10.9, ``Suppression Pool Makeup-MODE 3,'' to allow
installation of reactor cavity gate 2 in the Upper Containment Pool
(UCP) and draining the reactor cavity pool portion of the UCP while
still in MODE 3, with the reactor pressure less than 230 pounds per
square inch gauge (psig). It also modifies the applicability of the UCP
gates surveillance requirement (TS Section 3.6.2.4, ``Suppression Pool
Makeup (SPMU) System,'') to allow installation of UCP gates in MODES 1,
2, and 3.
Date of issuance: September 6, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No: 154.
Facility Operating License No. NPF-29: The amendment revises the
Technical Specifications and Surveillance Requirements.
Date of initial notice in Federal Register: April 30, 2002 (67 FR
21289). The August 16 and 22, 2002, supplemental letters 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 September 4, 2002.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: February 19, 2002, as
supplemented by letter dated July 17, 2002.
Brief description of amendment: This amendment revises Technical
Specification 3.8.1, ``AC Sources--Operating,'' to remove all current
Mode restrictions associated with testing the High Pressure Core Spray
Diesel Generator 13 during normal operation. The proposed changes
remove the restriction associated with Surveillance Requirements (SRs)
that prohibit performing the required testing in
[[Page 61691]]
Modes 1, 2, or 3. The specific SRs addressed in this amendment are: SR
3.8.1.11, 3.8.1.12, 3.8.1.16, and 3.8.1.19.
Date of issuance: September 10, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No: 155.
Facility Operating License No. NPF-29: The amendment revises the
Technical Specifications and Surveillance Requirements.
Date of initial notice in Federal Register: April 30, 2002 (67 FR
21288). The supplemental letter dated July 17, 2002, provided
clarifying information that did not change the scope of 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 September 10, 2002.
No significant hazards consideration comments received: None.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of application for amendments: March 8, 2002.
Brief description of amendments: The amendments revise TS 3.8.4,
``DC Sources-Operating,'' 3.8.5, ``DC Sources-Shutdown,'' 3.8.6,
``Battery Cell Parameters,'' and 3.8.8, ``Inverter-Shutdown.'' The
changes also include the relocation of the following TS items to a
licensee-controlled program: (1) A number of Surveillance Requirements
(SRs) that require the performance of preventive maintenance, and (2)
TS Table 3.8.6-1, ``Battery Cell Parameter Requirements.'' The
amendments also add new actions and their associated completion times
to TS 3.8.6 for out-of-limits conditions for battery cell voltage,
electrolyte level, and electrolyte temperature. In addition, SRs are
added for verification of these parameters.
Date of issuance: September 19, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 129, 129, 124 & 124.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: May 14, 2002 (67 FR
34485). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 19, 2002.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: August 1, 2001, as supplemented
June 19 and September 9, 2002.
Brief description of amendments: The amendments revise Technical
Specification 5.3, ``Unit Staff Qualifications,'' concerning approval
of the education and experience eligibility requirements for operator
license applicants.
Date of issuance: September 17, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 194 & 187.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 31, 2001 (66 FR
55018). The supplements dated June 19 and September 9, 2002, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 17, 2002.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: August 01, 2001, as
supplemented June 19 and September 09, 2002.
Brief description of amendments: The amendments revise Technical
Specifications requirements regarding Facility Staff Qualifications for
licensed operator and non-licensed personnel training programs. The
changes revise requirements that have been superseded based on licensed
operator training programs being accredited by the Institute of Nuclear
Power Operations, promulgation of the revised 10 CFR part 55,
``Operators' Licenses,'' which became effective on May 26, 1987, and
adoption of a systems approach to training as required by 10 CFR
50.120, ``Training and qualification of nuclear power plant
personnel.''
Date of issuance: September 17, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 154 & 140.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 31, 2001 (66 FR
55018). The supplements dated June 19 and September 09, 2002, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 17, 2002.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
County, Pennsylvania
Date of application for amendments: August 1, 2001, as supplemented
June 19 and September 9, 2002.
Brief description of amendments: The amendments revised Technical
Specification 5.3.1 to state that the licensed operators shall comply
with the qualification requirements in 10 CFR part 55, rather than the
American National Standards Institute's (ANSI) standard ANSI N18.1-
1971.
Date of issuance: September 17, 2002.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendments Nos.: 245, 249.
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 31, 2001 (66 FR
55018). The June 19 and September 9, 2002, letters provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the application beyond
the scope of the original Federal Register notice. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated September 17, 2002.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: August 1, 2001, as supplemented
June 19 and September 9, 2002.
[[Page 61692]]
Brief description of amendments: The amendments revise Technical
Specification requirements that have been superceded based on the
licensed operator training program being accredited by the Institute of
Nuclear Power Operations, promulgation of the revised 10 CFR part 55,
and adoption of a systems approach to training as required by 10 CFR
50.120.
Date of issuance: September 18, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 208 & 203.
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 31, 2001 (66 FR
55018). The supplements dated June 19 and September 9, 2002, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 18, 2002.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: November 9, 2000.
Brief description of amendment: This amendment revises the allowed
outage time from 72 hours to 7 days for one low pressure injection
train, and one containment spray system train. The supporting analysis
for the request is based on the Babcock & Wilcox Owners Group (B&WOG)
Topical Report BAW-2295A, Revision 1 & 2, ``Justification for the
Extension of Allowed Outage Time for Low pressure Injection and Reactor
Building Spray Systems,'' and its review by the staff documented in a
Safety Evaluation Report. The Davis-Besse Nuclear Power Station is the
lead B&WOG plant requesting these changes to be made to the Technical
Specifications.
Date of issuance: September 17, 2002.
Effective Date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 253.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 27, 2000 (65
FR 81919). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 17, 2002.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: January 18, 2002.
Brief description of amendments: These amendments revised Technical
Specifications to relocate specific working hour limits and controls to
administrative procedures.
Date of issuance: September 10, 2002.
Effective Date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 185 and 128.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 19, 2002 (67
FR 7418). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 10, 2002.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: July 26, 2002, as supplemented
August 23, 2002
Brief description of amendments: The amendments will add a license
condition to the Operating Licenses for both units, allowing a one-time
140-hour allowed outage time for the essential service water (ESW)
system, to allow ESW pump replacement during plant operation.
Date of issuance: September 9, 2002.
Effective date: As of the date of issuance and shall be implemented
within 20 days.
Amendment Nos.: 270 and 251.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Facility Operating License.
Date of initial notice in Federal Register: August 8, 2002 (67 FR
51603). The August 23, 2002, letter provided clarifying information
within the scope of the original application and 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 September 9, 2002.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station, Unit No. 1, Oswego County, New York
Date of application for amendment: October 19, 2001, as
supplemented June 17, 2002.
Brief description of amendment: The amendment revised the Technical
Specifications to implement programmatic controls for radiological
effluent technical specifications in the Administrative Controls
section, to relocate certain procedural details to licensee-controlled
documents, and to add new programs to accommodate existing NRC
requirements and guidance.
Date of issuance: September 11, 2002.
Effective date: September 11, 2002.
Amendment No.: 176.
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: January 8, 2002 (67 FR
928). The June 17, 2002, supplemental letter did not expand the scope
of the application as originally noticed and did not change the
proposed no significant hazards consideration determination. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated September 11, 2002.
No significant hazards consideration comments received: No.
North Atlantic Energy Service Corporation, et al., Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: August 9, 2001, as supplemented
September 17, 2001, and June 24, 2002.
Description of amendment request: The amendment combines Technical
Specifications (TSs) 3/4.9.9, ``Containment Purge and Exhaust Isolation
System,'' and 3/4.9.4, ``Containment Building Penetrations.'' By
combining these two TSs, the amendment updates the Seabrook TSs related
to refueling operations by adopting portions of NUREG-1431, ``Standard
Technical Specifications, Westinghouse Plants,'' Revision 2. The
amendment also changes the TS index pages and the associated TS Bases.
By letter dated June 24, 2002, the licensee withdrew that part of the
application associated with relocation of TS 3/4.9.4, ``Decay Time,''
to the Seabrook Station Technical Requirements Manual.
Date of issuance: September 5, 2002.
Effective date: As of its date of issuance, and shall be
implemented within 90 days.
[[Page 61693]]
Amendment No.: 85.
Facility Operating License No. NPF-86: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 19, 2001 (66
FR 48290). The supplements dated September 17, 2001, and June 24, 2002,
provided clarifying information that did not change the initial
proposed no significant hazards consideration determination. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated September 5, 2002.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of application for amendment: January 28, 2002.
Brief description of amendment: The amendment revises the Core
Operating Limits Report analytical methods referenced in Technical
Specification (TS) 5.6.5.b. Specifically, the amendment adds references
to two NRC-approved Framatome ANP, Inc., reports: (1) EMF-2310(P)(A),
Revision 0, ``SRP [Standard Review Plan] Chapter 15 Non-LOCA [loss-of-
coolant accident] Methodology for Pressurized Water Reactors [PWRs],''
dated May 2001, and (2) EMF-2328(P)(A), Revision 0, ``PWR Small Break
LOCA Evaluation Model, S-RELAP5 Based,'' dated March 2001. The
amendment also deletes previous references in TS 5.6.5.b describing
Exxon Nuclear Company's large-break LOCA evaluation model.
Date of issuance: September 13, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 209.
Facility Operating License No. DPR-20: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 19, 2002 (67
FR 7420). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 13, 2002.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: April 3, 2002.
Brief description of amendment: This amendment consists of changes
to the Technical Specifications (TSs) which allow the relocation of TS
3/4.4.4, ``Reactor Coolant System--Chemistry,'' and the associated
bases from the TSs to the Hope Creek Updated Final Safety Analysis
Report (UFSAR).
Date of issuance: September 18, 2002.
Effective date: September 18, 2002, and shall be implemented within
60 days.
Amendment No.: 140.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications and the UFSAR.
Date of initial notice in Federal Register: May 14, 2002 (67 FR
34492). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 18, 2002.
No significant hazards consideration comments received: No.
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: May 24, 2002.
Brief description of amendments: The amendments revised the
Technical Specifications to allow Mode 2 (startup) operation with two
out of four, rather than three out of four, required intermediate range
monitor channels per trip system.
Date of issuance: September 12, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 233/175.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 9, 2002 (67 FR
45572). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 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: May 31, 2001, as supplemented by letters
dated June 14, August 13, October 16, November 7, 2001, August 14,
2002, and September 4, 2002.
Brief description of amendments: The amendment grants conforming
amendments to the operating licenses to reflect the direct transfer of
Reliant Energy Incorporated's ownership interest to Texas Genco, LP.
Date of issuance: September 4, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1-142; Unit 2-130.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the facility operating licenses.
Date of initial notice in Federal Register: September 28, 2001 (66
FR 49711). The supplemental information did not expand the scope of the
application as originally noticed in the Federal Register. The
Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated September 4, 2002.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: April 1, 2002, as supplemented by letter
dated June 6, 2002.
Brief description of amendments: The amendments include addition of
topical report ERX-2001-005, ``ZIRLO\TM\ Cladding and Boron Coating
Models for TXU Electric's Loss of Coolant Accident Analysis
Methodologies,'' to the list of approved methodologies for use in
generating the Core Operating Limits Report in Technical Specification
(TS) 5.6.5, ``Core Operating Limits Report (COLR).'' In addition, the
proposed changes include ZIRLO\TM\ clad in the description of the fuel
assemblies in TS 4.2.1, ``Fuel Assemblies.''
Date of issuance: September 4, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 99 and 99.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 14, 2002 (67 FR
34493). The June 6, 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 amendments is
contained in a Safety Evaluation dated September 4, 2002.
No significant hazards consideration comments received: No.
[[Page 61694]]
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: March 27, 2002.
Brief description of amendments: The amendments revise Technical
Specification (TS) 5.3.1 to require that each member of the unit staff,
with the exception of licensed Reactor Operators (ROs) and licensed
Senior Reactor Operators (SROs), shall meet or exceed the minimum
qualifications of Regulatory Guide (RG) 1.8, ``Qualification and
Training of Personnel for Nuclear Power Plants,'' Revision 2, 1987.
Also, a new TS 5.3.2 is added to require that the ROs and SROs shall
meet or exceed the minimum qualifications of RG 1.8, Revision 3, May
2000, and the current TS 5.3.2 is renumbered to TS 5.3.3.
Date of issuance: September 4, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 100 and 1000.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 14, 2002 (67 FR
34493). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 4, 2002.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: June 17, 2002 (ULNRC-04684).
Brief description of amendment: The amendment revised Technical
Specification 3.3.1, ``Reactor Trip System (RTS) Instrumentation,'' by
adding Surveillance Requirement (SR) 3.3.1.16 to Function 3 of TS Table
3.3.1-1. SR 3.3.1.16 verifies that the reactor trip system response
times are within limits every 18 months on a staggered test basis.
Date of issuance: September 3, 2002.
Effective date: September 3, 2002, and shall be implemented within
60 days from the date of issuance.
Amendment No.: 151.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 23, 2002 (67 FR
48222). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 3, 2002.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: February 15, 2001, as
supplemented by letters dated April 20 and November 7, 2001, and March
1 and August 5, 2002.
Brief description of amendment: The amendment revises paragraph
d.1.j) 2) of Technical Specification (TS) 5.5.9, ``Steam Generator (SG)
Tube Surveillance Program,'' to (1) delete the requirement that all SG
tubes containing an Electrosleeve TM, a Framatome
proprietary process, be removed from service within two operating
cycles following installation of the first ElectrosleeveTM;
(2) add the requirement that ElectrosleevesTM will not be
installed in the outermost periphery tubes of the SG bundles where
potentially locked tubes would cause high axial loads; (3) revise the
references describing electrosleeving; and (4) add the requirement that
all sleeves with detected inside diameter flaw indications will be
removed from service upon detection. In addition, if an
ElectrosleeveTM tube pull is performed by the licensee, the
licensee has agreed to provide the results of the tube examination to
the NRC staff within 60 days of when the final results of the
examination are made available to the licensee.
Date of issuance: September 13, 2002.
Effective date: September 13, 2002, and shall be implemented within
60 days of the date of issuance.
Amendment No.: 153.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 14, 2002 (67 FR
34494). The supplemental letter of August 5, 2002, provided additional
information that clarified the application, 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 September 13, 2002.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: June 27, 2000, and its supplements dated
January 31, 2001, May 2, 2001, October 30, 2001, and May 10, 2002.
Brief description of amendment: The amendment revised the antitrust
conditions for Kansas Gas and Electric Company (KGE) in Appendix C to
the operating license. The revisions (1) add a statement that the
antitrust conditions do not restrict the rights of Kansas Electric
Power Cooperative, Inc. (KEPCo) or the duties of KGE, that may exist
beyond, and are not inconsistent with, the antitrust conditions, (2)
define ``KGE members in licensee's service area'' in the appendix to
include all KEPCo members with facilities in Western Resources' and
KGE's combined service area, (3) delete license conditions restricting
KEPCo's use of the power from WCGS, (4) remove out-of-date conditions,
and (5) update conditions to be consistent with the terms and
conditions of Western Resources' Federal Energy Regulatory Commission
open access transmission tariff. Western Resources is the parent
company of KGE.
Date of issuance: September 6, 2002.
Effective date: September 6, 2002, and shall be implemented within
90 days from the date of issuance.
Amendment No.: 147.
Facility Operating License No. NPF-42: The amendment revised
Appendix C, ``Antitrust Conditions for Kansas Gas and Electric
Company,'' to the operating license.
Date of initial notice in Federal Register: July 26, 2000 (65 FR
46010). The supplemental letters dated January 31, 2001, May 2, 2001,
October 30, 2001, and May 10, 2002, provided additional clarifying
information that did not expand the application beyond the scope of the
initial notice or change the staff's proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated September 6, 2002.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 20th day of September, 2002.
For the Nuclear Regulatory Commission.
Stuart A. Richards,
Acting Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 02-24616 Filed 9-30-02; 8:45 am]
BILLING CODE 7590-01-P