[Federal Register Volume 67, Number 209 (Tuesday, October 29, 2002)]
[Notices]
[Pages 66005-66019]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-27243]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from, October 4, 2002, through October 17, 2002.
The last biweekly notice was published on October 15, 2002 (67 FR
63687).
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 Rgister 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
[[Page 66006]]
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 November 29, 2002, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714,\1\ which is
available at the Commission's PDR, located at One White Flint North,
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly
available records will be accessible from the Agencywide Documents
Access and Management System's (ADAMS) Public Electronic Reading Room
on the Internet at the NRC web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to
intervene is filed by the above date, the Commission or an Atomic
Safety and Licensing Board, designated by the Commission or by the
Chairman of the Atomic Safety and Licensing Board Panel, will rule on
the request and/or petition; and the Secretary or the designated Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
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\1\ The most recent version of Title 10 of the Code of Federal
Regulations, published January 1, 2002, inadvertently omitted the
last sentence of 10 CFR 2.714(d) and subparagraphs (d)(1) and (2),
regarding petitions to intervene and contentions. Those provisions
are extant and still applicable to petitions to intervene. Those
provisions are as follows: ``In all other circumstances, such ruling
body or officer shall, in ruling on--
(1) A petition for leave to intervene or a request for hearing,
consider the following factors, among other things:
(i) The nature of the petitioner's right under the Act to be
made a party to the proceeding.
(ii) The nature and extent of the petitioner's property,
financial, or other interest in the proceeding.
(iii) The possible effect of any order that may be entered in
the proceeding on the petitioner's interest.
(2) The admissibility of a contention, refuse to admit a
contention if:
(i) The contention and supporting material fail to satisfy the
requirements of paragraph (b)(2) of this section; or
(ii) The contention, if proven, would be of no consequence in
the proceeding because it would not entitle petitioner to relief.''
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As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, 11555 Rockville Pike (first floor),
Rockville, Maryland, by the above date. Because of continuing
disruptions in delivery of mail to United States Government offices, it
is requested that petitions for leave to intervene and requests for
hearing be transmitted to the Secretary of the Commission either by
means of facsimile transmission to 301-415-1101 or by e-mail to
[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
[[Page 66007]]
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].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station (PVNGS),
Units 1, 2, and 3, Maricopa County, Arizona
Date of amendments request: September 6, 2002.
Description of amendments request: The amendments would replace the
peak linear heat safety limit, in Technical Specification (TS) 2.1.1.2,
``Reactor Core SLs [Safety Limits],'' by a peak fuel centerline
temperature safety limit to have a safety limit in the TSs that would
not be exceeded during normal operation or anticipated operational
occurrences (AOOs), in accordance with Section 50.36(c)(1)(ii)(A) of
Title 10 of the Code of Federal Regulations (10 CFR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change does not require any physical change to
plant systems, structures, or components nor does it require any
change in systems or plant operations. The proposed change does not
result in any change to safety analysis methods or results. The
change to establish peak fuel centerline temperature as the Safety
Limit is consistent with the PVNGS Units 1, 2 and 3 licensing bases
for ensuring that the fuel design limits are met. Operations and
analysis will continue to be in accordance with the PVNGS Units 1, 2
and 3 licensing bases. The peak fuel centerline temperature is the
basis for protecting the fuel and is consistent with the safety
analysis. [The peak linear heat rate and peak fuel centerline
temperature safety limits are not initiators of accidents.]
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The PVNGS Units 1, 2 and 3 Updated Final Safety Analysis Report
(UFSAR) Chapter 15 accident analyses for AOOs where the peak linear
heat rate may exceed the existing Safety Limit of 21 kW/ft are the
control element assembly (CEA) Withdrawal events at Subcritical and
Low Power conditions. The analyses for these AOOs indicate that the
peak fuel centerline temperature is not exceeded. The existing
safety analyses, which remain unchanged, do not affect any accident
initiators that would create a new accident. [The peak linear heat
rate and peak fuel centerline temperature safety limits are not
initiators of accidents.]
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change does not result in any change to safety
analysis methods or results. Therefore, by changing the Safety Limit
from peak linear heat rate to peak fuel centerline temperature the
margins as established in the PVNGS Units 1, 2 and 3 Technical
Specifications and UFSAR are unchanged.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, APS [Arizona Public Service Company]
concludes that the activities associated with the proposed
amendment[s] presents no significant hazards consideration under the
standards set forth in 10 CFR 50.92(c) and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three 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: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Section Chief: Stephen Dembek.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendment request: September 20, 2002.
Description of amendment request: The proposed amendment revises
Technical Specification (TS) 3.9.5, Shutdown Cooling (SDC) and Coolant
Circulation--Low Water Level, for Unit Nos. 1 and 2 to add two notes to
allow operational changes in the Shutdown Cooling System to support
operations and testing. The changes would allow the SDC pumps to be
deenergized for less than or equal to 15 minutes when switching from
one train to another. The second change would allow one SDC loop to be
inoperable for up to 2 hours for surveillance testing, provided that
the other loop was operable and in operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The system affected by this proposed amendment is the Shutdown
Cooling (SDC) System. This system mitigates the consequences of a
boron dilution event and removes decay heat from the Reactor Coolant
System when the unit is in Mode 6. This proposed amendment revises
the Technical Specification to allow the SDC pumps to be deenergized
for less than or equal to 15 minutes to allow swapping from one
operating train to another, and would allow one SDC loop to be
inoperable for up to two hours for surveillance testing. Because
this system is used for the mitigation of an accident, it is not an
accident initiator. Therefore, the probability of an accident
previously evaluated is not increased.
The only design basis accident considered in this Mode is a
boron dilution event. Consideration is also given to a loss of decay
heat removal in this Mode as well. Both of these conditions are
evaluated in the Updated Final Safety Analysis Report (UFSAR). The
evaluations consider operation of the SDC system to mitigate these
conditions. Removing this system from service for a limited amount
of time, with other operational restrictions, limits the
consequences to those already assumed in the UFSAR. Thus, no
increase in offsite dose occurs under this conditions. Therefore,
the consequences of an accident previously evaluated have not
increased.
Therefore, the probability or consequences of an accident
previously evaluated have not significantly increased.
2. Would not create the possibility of a new or different [kind]
of accident from any accident previously evaluated.
The proposed changes do not involve a significant change in the
operation of the plant and no new accident initiation mechanism is
created by the proposed changes. The SDC System is not being altered
by this amendment request. No substantial changes are made in the
way in which the SDC System is operated. The only change made would
allow both SDC pumps to be
[[Page 66008]]
deenergized to swap operating trains, and one SDC inoperable for
less than two hours to allow for surveillance testing. Since the SDC
System is an accident mitigating system only, changes in when this
system is needed to operate cannot create a new [kind] of accident.
Therefore, the possibility of a new or different [kind] of
accident from any previously evaluated is not created.
3. Would not involve a significant reduction in a margin of
safety.
The margin of safety provided by the SDC System is to provide
boration control and to remove decay and sensible heat from the
Reactor Coolant System as described in the UFSAR. Removal of system
components from service as described above, and with limitations in
place to prevent boron dilution and loss of decay and sensible heat
removal, does not significantly impact the margin of safety. The SDC
System will continue to be able to provide its safety function under
this conditions. Operators will continue to have adequate time to
respond to any off-normal events. Removing the system from service,
for a limited period of time, with other operational restrictions
limits the consequences to those already assumed in the UFSAR.
Therefore, no reduction in [a] margin of safety has occurred because
the event results in the UFSAR are not changed by operation in the
proposed conditions.
Therefore, the proposed changes do not involve a significant
reduction in [a] margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Richard J. Laufer.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, and Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, located in Mecklenburg County, North
Carolina and York County, South Carolina
Date of amendment request: August 26, 2002.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) for diesel fuel oil for the
plant's onsite diesel-generator power sources. The proposed changes
would allow the use of an optional water and sediment content test,
would relocate the specific version of certain American Society for
Testing and Materials (ASTM) references to licensee controlled
documents, would add several new ASTM references, and would relocate
the requirement for a 10-year diesel fuel oil tank inspection and
cleaning to licensee controlled documents. The licensee stated that the
changes are consistent with the Standard Technical Specification
Travelers (TSTF) 374, Revision 0 and TSTF 2, Revision 1. Associated
changes are also proposed for the TS Bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The following discussion is a summary of the evaluation of the
change contained in this proposed amendment against the 10 CFR 50.92
(c) requirements to demonstrate that all three standards are
satisfied. A no significant hazards consideration is indicated if
operation of the facility in accordance with the proposed amendments
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.
First Standard
The proposed changes relocate the specific American Society for
Testing and Materials (ASTM) Standard references from the
Administrative Controls Section of Technical Specifications (TS) to
a licensee-controlled document. Since any changes of the licensee-
controlled document will be evaluated to the requirements of 10 CFR
50.59, ``Changes, tests, and experiments,'' no increase in the
probability or consequences of an accident previously evaluated is
involved. In addition, the ``clear and bright'' test used to
establish the acceptability of new fuel oil for use prior to
addition to the storage tanks has expanded to allow a water and
sediment content test to be performed to establish the acceptability
of new fuel oil. The Bases for SR 3.8.3.3 (CNS) and 3.8.3.2 (MNS)
are revised to indicate that the API gravity is tested in accordance
with ASTM D1298 or D287.
Relocating the specific ASTM Standard references from the TS to
a licensee-controlled document, allowing a water and sediment test
to be performed to establish the acceptability of new fuel oil, and
revising the TS Bases will not affect or degrade the ability of the
emergency diesel generators (DGs) to perform their specified safety
function. Fuel oil quality will continue to meet ASTM requirements.
In addition Surveillance Requirements (SR) 3.8.3.5 for McGuire
and 3.8.3.6 for Catawba are revised to remove the requirement for a
10-year tank inspection and cleaning. This requirement will be moved
to a licensee-controlled document. Any changes of the licensee-
controlled document will be evaluated to the requirements of 10CFR
50.59 ``Changes, tests, and experiments,''.
This change will not affect or degrade the ability of the
emergency diesel generators (DGs) to perform their specified safety
function. Fuel oil quality will continue to meet ASTM requirements.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, and
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, or components (SSCs) from performing their
intended function to mitigate the consequences of an initiating
event within the assumed acceptance limits. The proposed change does
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 and
amounts of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures.
Therefore, the changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Second Standard
The proposed changes relocate the specific ASTM Standard
references from the Administrative Controls Section of TS to a
licensee-controlled document. In addition, the ``clear and bright''
test used to establish the acceptability of new fuel oil for use
prior to addition to storage tanks has been expanded to allow a
water and sediment content test to be performed to establish the
acceptability of new fuel oil. The proposed changes revise Bases B
3.8.3 to reference the current specific ASTM standards. The Bases
for SRs 3.8.3.3 (CNS) and 3.8.3.2 (MNS) are revised to indicate that
the API gravity is tested in accordance with ASTM D1298 or D287.
In addition Surveillance Requirements (SR) 3.8.3.5 for McGuire
and 3.8.3.6 for Catawba are revised to remove the requirement for a
10-year tank inspection and cleaning. This requirement will be moved
to a licensee-controlled document. Any changes of the licensee-
controlled document will be evaluated to the requirements of
10CFR50.59 ``Changes, tests, and experiments,''.
The changes do not involve a physical alteration to 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 or licensing basis.
Therefore, the changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Third Standard
The proposed changes relocate the specific ASTM Standard
references from the Administrative Control Section of TS to a
licensee-controlled document. Instituting the proposed changes will
continue to ensure the use of the current applicable ASTM
[[Page 66009]]
Standards to evaluate the quality of both new and stored fuel oil
designated for use in the emergency diesels. The detail associated
with the specific ASTM Standard references is not required to be in
the TS to provide adequate protection of the public health and
safety, since the TS still retains the requirement for compliance
with the applicable ASTM standard. Changes to the licensee-
controlled document are performed in accordance with the provisions
of 10 CFR 50.59. Should it be determined that future changes involve
a potential reduction in a margin of safety, NRC review and approval
would be necessary prior to the implementation of the changes. This
approach provides an effective level of regulatory control and
provides for a more appropriate change control process.
The ``clear and bright'' test used to establish the
acceptability of new fuel oil for use prior to the addition to
storage tanks has been expanded to allow a water and sediment
content test to be performed to establish the acceptability of new
fuel oil. The proposed changes revise Bases B 3.8.3 to allow
reference to the current ASTM standard. The Bases for SR 3.8.3.3 is
revised to indicate that the API gravity is tested in accordance
with ASTM D1298 or D287. The level of safety of facility operation
is unaffected by the proposed changes since there is no change in
the intent of the TS requirements of assuring fuel oil is of the
appropriate quality for emergency DG use.
In addition Surveillance Requirements (SR) 3.8.3.5 for McGuire
and 3.8.3.6 for Catawba are revised to remove the requirement for a
10-year tank inspection and cleaning. This requirement will be moved
to a licensee-controlled document. Any changes of the licensee-
controlled document will be evaluated to the requirements of
10CFR50.59 ``Changes, tests, and experiments''. The level of safety
of the facility operation is unaffected by the proposed changes
since there is no change in the intent of the SR to clean and
inspect the fuel tanks.
Therefore, the proposed changes listed above do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 15, 2001, as supplemented by
letter dated August 27, 2002.
Description of amendment request: The proposed amendment request
provides additional information to support a modification to Technical
Specification 3.4.7 and limits Reactor Coolant System activity
permitted by the ACTION statement to 60 microcuries per gram ([mu]Ci/
gm) at all power levels. The letdown line break accident analysis in
the Final Safety Analysis Report is also changed to reflect revised
dose consequences. This notice supercedes the biweekly Federal Register
notice dated November 28, 2001 (66 FR 59504), based on the original
application dated October 15, 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will the operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: The proposed change to the Technical Specifications
(TS) conservatively limits Reactor Coolant System (RCS) activity
permitted by Action Statement 3.4.7.a to 60 [mu]Ci/gm at all reactor
power levels. The proposed change to the Final Safety Analysis
Report (FSAR) Section 15.6.3.1 revises the letdown line break
accident analyses.
The probability of a previously evaluated accident is not
affected by this change because the pre-existing iodine spike is not
an accident initiator and the new letdown line break accident
analysis does not affect any plant Structure, Systems, or Component
(SSC) but merely determines the consequences of the previously
evaluated accident.
This TS change is conservative in that it will reduce the
accident consequences for events occurring at lower power levels.
The new letdown line break accident analysis meets the original
Safety Evaluation Report (SER) and the current Standard Review Plan
(SRP) acceptance criteria of a small fraction of the 10 CFR [Part]
100 limits.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. Will the operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
Response: The probability of a new or different accident is not
affected by this change because the new letdown line break analysis
does not affect any plant Structure, Systems, or Component but
merely determines the consequences of the previously evaluated
accident.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. Will the operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: The TS change is more limiting in that it will reduce
the accident consequences for events occurring at lower power
levels.
The new letdown line break accident analysis, assuming one
operating charging pump, meets the original SER and current SRP
acceptance criteria of a small fraction of the 10 CFR [Part] 100
limits. This single pump analysis provides a suitable licensing
basis analysis and has sufficient conservatism to accommodate two
and three pump operating scenarios that may exist during the
operating cycle.
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: N. S. Reynolds, Esquire, Winston & 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.
50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle
County, Illinois; Docket Nos. 50-254 and 50-265, Quad Cities Nuclear
Power Station, Units 1 and 2, Rock Island County, Illinois
Date of application for amendment request: September 27, 2002.
Description of amendment request: The proposed amendment changes
Appendix B, ``Environmental Protection Plan (Non-Radiological),'' of
the license by removing a parenthetical reference to a superseded
section of 10 CFR 51.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change deletes a reference to a superseded section
of 10 CFR 51, ``Environmental Protection Regulations for Domestic
Licensing and Related Regulatory Functions,'' found in the non-
radiological Environmental Protection Plans (EPPs) for Byron
Station, LaSalle County Station and Quad Cities Nuclear Power
Station, Units 1 and 2. The EPP (Non-Radiological) is Appendix B to
the Facility Operating License. The change is administrative in
nature. No physical changes to the facilities will result from the
proposed change. The initial conditions and methodologies used in
accident analyses remain unchanged. The
[[Page 66010]]
proposed change does not revise or alter the design assumptions for
systems or components used to mitigate the consequences of
accidents. Thus, accident analyses results are not impacted by this
proposed change.
Therefore, this proposed amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change deletes a reference to a superseded section
of 10 CFR 51.5. The change is administrative in nature. No physical
or operational changes to the facilities will result from the
proposed change.
The proposed change does not affect the design or operation of
any system, structure, or component (SSC) in the plant. The safety
functions of the related SSCs are not changed in any manner, nor is
the reliability of any SSC reduced. The change does not affect the
manner by which the facility is operated and does not change any
facility, structure, system, or component. No new or different type
of equipment will be installed by this proposed change.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change is administrative in nature and has no
impact on the margin of safety of any Technical Specification. There
is no impact on safety limits or limiting safety system settings.
The change does not affect any plant safety parameters or setpoints.
The proposed change deletes an inaccurate reference to a section of
10 CFR 51 that has been superseded. No physical or operational
changes to the facility will result from the proposed changes.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County,
Pennsylvania
Date of amendment request: January 16, 2002.
Description of amendment request: The amendment would make
administrative, editorial, and format (including repagination) changes
to the technical specification (TS) Bases index and the Administrative
Control section of TSs. Specifically, the amendments would relocate the
TS Bases page listings from the TS index to a TS Bases index, and
remove certain duplicative administrative requirements from Section 6,
``Administrative Controls,'' of the TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed administrative changes to the TS index and to
Section 6 of the TSs do not result in changes being made to
structures, systems, or components (SSCs), or to event initiators or
precursors. Also, the proposed changes do not impact the design of
plant systems such that previously analyzed SSCs would now be more
likely to fail. The initiating conditions and assumptions for
accidents described in the Updated Final Safety Analysis Report
(UFSAR) remain as previously analyzed. Thus, the proposed changes do
not involve a significant increase in the probability of an accident
previously evaluated.
The previously analyzed SSCs are unaffected by the proposed
changes and continue to provide assurance that they are capable of
performing their intended design function in mitigating the effects
of design basis accidents (DBAs). As such, the consequences of
accidents previously evaluated in the UFSAR will not be increased
and no additional radiological source terms are generated.
Therefore, there will be no reduction in the capability of those
SSCs in limiting the radiological consequences of previously
evaluated accidents and reasonable assurance that there is no undue
risk to the health and safety of the public will continue to be
provided. Thus, the proposed changes do not involve a significant
increase in the consequences of an accident previously evaluated.
Therefore, the proposed administrative changes do not
significantly increase the probability or consequences of any
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. The proposed administrative changes do not involve physical
changes to analyzed SSCs or changes to the modes of plant operation
defined in the technical specification. The proposed changes do not
involve the addition or modification of plant equipment (no new or
different type of equipment will be installed) nor do they alter the
design or adversely affect operation of any plant systems. No new
accident scenarios, accident or transient initiators or precursors,
failure mechanisms, or limiting single failures are introduced as a
result of the proposed changes.
The proposed administrative changes do not cause the malfunction
of safety-related equipment assumed to be operable in accident
analyses. No new or different mode of failure has been created and
no new or different equipment performance requirements are imposed
for accident mitigation. As such, the proposed changes have no
effect on previously evaluated accidents.
Therefore, the proposed administrative changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
No. The proposed administrative changes do not affect any
previously evaluated accident. The proposed changes do not adversely
affect the TS requirements and will continue to ensure that the
necessary plant equipment is operable in the plant conditions where
these systems are required to operate to mitigate a DBA as described
in the analyses presented in the UFSAR. Thus, the proposed
administrative, editorial, and format changes do not affect plant
safety.
Therefore, the proposed administrative changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit 2, Berrien County, Michigan
Date of amendment request: July 23, 2002.
Description of amendment request: The proposed amendment would
revise the Unit 2 reactor coolant system (RCS) pressure-temperature
curves in Technical Specification (TS) Figures 3.4-2 and 3.4-3 and
associated TS Bases. The revised curves will bound operation of the
unit for the remainder of its current license duration and bound
operation with planned license amendments to increase the power level
at which the unit is allowed to operate.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
[[Page 66011]]
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
Response: No.
Probability of Occurrence of an Accident Previously Evaluated
The proposed change will revise the RCS pressure-temperature
curves to bound operation of the reactor for up to 32 EFPY at a
power level of up to 3800 MW for the current fuel cycle and beyond,
to reflect new fluence analysis methodology, to reflect the use of
ASME [American Society of Mechanical Engineers] Code Case N-641, to
include boltup limits, and to no longer include instrument
uncertainty margins.
The proposed change will not result in physical changes to
structures, systems, or components (SSCs), or to event initiators or
precursors. The proposed change will not affect the ability of
personnel to control RCS [Reactor Coolant System] pressure at low
temperatures and, thereby, ensure the integrity of the RCPB [Reactor
Coolant Pressure Boundary]. Use of ASME Code Case N-641 will be
approved by the NRC through approval of a Donald C. Cook Nuclear
Plant-specific exemption to requirements in 10 CFR 50.60(a) and 10
CFR 50, Appendix G. Therefore, the proposed revision to the RCS
pressure-temperature curve changes will have been determined in
accordance with NRC accepted methodologies. These methodologies
provide adequate assurance that the reactor vessel will withstand
the effects of normal cyclic loads due to temperature and pressure
changes, and provide an acceptable level of protection against
brittle failure. Additionally, the proposed changes will not impact
the design or operation of plant systems such that previously
analyzed SSCs will be more likely to fail. The initiating conditions
and assumptions for accidents described in the UFSAR will remain as
previously analyzed. Therefore, the proposed changes will not
involve a significant increase in the probability of an accident
previously evaluated.
Consequences of an Accident Previously Evaluated
The proposed change does not reduce the ability of any SSC to
limit the radiological consequences of accidents described in the
UFSAR. The proposed change will not alter any assumptions made in
the analysis of radiological consequences of previously evaluated
accidents, nor does it affect the ability to mitigate these
consequences. No new or different radiological source terms will be
generated as a result of the proposed change. Therefore, the
proposed changes do not involve a significant increase in the
consequences of an accident previously evaluated.
The format changes will improve the appearance of the affected
pages but will not affect any requirements. In summary, the
probability of occurrence and the consequences of an accident
previously evaluated will not be significantly increased.
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 result in physical changes to SSCs.
The proposed change will not involve the addition or modification of
plant equipment (no new or different type of equipment will be
installed) nor will it alter the design of any plant systems. The
proposed change solely involves RCS pressure-temperature limits. The
types of potential accidents associated with these limits have been
previously identified and evaluated. No new accident scenarios,
accident or transient initiators or precursors, failure mechanisms,
or single failures will be introduced as a result of the proposed
changes. No new or different modes of failure will be created. The
format changes will improve the appearance of the affected pages but
will not affect any requirements. Therefore, the proposed change
will 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 RCS pressure-temperature curves will continue to
provide adequate margins of protection for the RCPB. The proposed
changes have been determined, through supporting analyses, to be in
accordance with the methodologies and criteria set forth in the
applicable regulations, or in accordance with technically adequate
alternatives. Compliance with these methodologies provides adequate
margins of safety and ensures that the RCPB will withstand the
effects of normal cyclic loads due to temperature and pressure
changes as well as the loads associated with postulated faulted
events as described in the UFSAR. The format changes will improve
the appearance of the affected pages but will not affect any
requirements. Therefore, the proposed change will not significantly
reduce the margin of safety.
In summary, based upon the above evaluation, [Indiana Michigan
Power Company] I&M has concluded that the proposed changes involve
no significant hazards consideration under the standards set forth
in 10 CFR 50.92(c), and, accordingly, a finding of ``no significant
hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of amendment request: September 30, 2002.
Description of amendment request: The proposed amendment requests
permission to change Kewaunee Nuclear Power Plant (KNPP) Facility
Operating License DRP-43 to use an upgraded computer code for design
basis accident containment integrity analyses. KNPP is currently
licensed to use code for Generation of Thermal-Hydraulic Information
for Containment (GOTHIC) version 6.0a. The proposed amendment requests
to use GOTHIC 7.0p2 (GOTHIC 7).
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.
Accident analyses affected by GOTHIC have each been evaluated
and found to show good agreement between the GOTHIC 7 analysis and
the current analysis of record (AOR). Safety analysis results using
GOTHIC 7 are shown to satisfy all applicable design and safety
analysis acceptance criteria. Since GOTHIC 7 conforms to design
bases and its results are bounded by the existing safety analyses,
its use within limits of the bounding accident analyses will not
cause an increase in the probability or consequences of an accident
previously evaluated. Adherence to safety analysis acceptance
criteria prevents use of GOTHIC 7 from creating new challenges to
components and systems that could adversely affect their ability to
mitigate accident consequence or diminish integrity of any fission
product barrier.
Thus, the requested upgrade to GOTHIC 7 with [mist diffusion
layer] MDL modeling option will not increase probability or the
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
Upgrade to GOTHIC 7 is a change in analysis methods applied to
Kewaunee [design basis accident] DBA. Analysis methods are not
accident initiators. GOTHIC 7 will be applied in the same manner
currently licensed and it is consistent with current plant design
bases and licensed accident analysis methodologies. It does not
adversely affect any fission product barrier, nor does it alter the
safety function of safety related systems, structures, and
components depended upon for accident prevention or mitigation.
Equipment important to safety will continue to function within
design. As demonstrated by the [Numerical
[[Page 66012]]
Applications Inc.] NAI report, GOTHIC 7 yields a representation of
expected plant response for affected design basis accidents that is
more accurate but remains conservative. GOTHIC 7 predicted results
for affected DBA remain bounded by the limiting analyses of record.
Thus, the requested upgrade to GOTHIC 7 does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Involve a significant reduction in the margin of safety.
Upgrade to GOTHIC 7 affects Kewaunee design basis [loss of
coolant accident] LOCA and [main steamline break] MSLB DBA
containment analyses. The results predicted by GOTHIC 7 for these
DBA analyses remain within limiting design basis accidents of
record. GOTHIC 7 accuracy and conservatism in this application has
been verified through benchmark analyses against the current
analyses of record, validated against recognized standard data, and
found to be appropriate for application to Kewaunee DBA. Safety
analysis acceptance criteria are satisfied and adherence to safety
analysis acceptance criteria using GOTHIC 7 assures that Technical
Specification limits will not be exceeded during normal operation.
Thus, upgrade to GOTHIC 7 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: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: September 19, 2002.
Description of amendment request: The proposed amendment would
delete Surveillance Requirement (SR) 4.6.B.2, ``Primary System
Boundary--Reactor Vessel Temperature and Pressure,'' from the
Monticello Technical Specifications (TSs) on the basis of the
licensee's commitment to (1) relocate the current requirements to the
Updated Safety Analysis Report (USAR) and (2) implement the Boiling
Water Reactor Vessel and Internals Program Integrated Surveillance
Program as approved by the Nuclear Regulatory Commission (NRC) in a
letter dated February 1, 2002. SR 4.6.B.2 currently states: ``Test
specimens representing the reactor vessel, base weld, and weld heat
affected zone metal shall be installed in the reactor vessel adjacent
to the vessel wall at the core midplane level. The material sample
program shall conform to ASTM [American Society for Testing and
Materials] E 185-66.'' The licensee would also make related changes to
the TS Bases 3.6/4.6.
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:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change relocates the requirement of the TS
Surveillance Requirement to a Licensee controlled document and
implements an integrated surveillance program that has been
evaluated by the NRC staff as meeting the requirements of paragraph
III.C of Appendix H to 10 CFR [Part] 50. The proposed change of
relocating a TS Surveillance Requirement to the Monticello USAR and
implementing an integrated surveillance program is not considered a
precursor or initiator of an accident previously evaluated. The
proposed change does not impact current plant operations or the
design function of any structure, system or component. Consequently,
the proposed change does not significantly increase the probability
of any accident previously evaluated.
The proposed change provides the same assurance of Reactor
Pressure Vessel integrity as has always been assured. The relocation
of the TS Surveillance Requirement provides an acceptable method for
implementing the integrated surveillance program which was evaluated
by the NRC staff as meeting the requirements of 10 CFR [Part] 50,
Appendix H, paragraph III.C. The relocation of the TS Surveillance
or the implementation of an integrated surveillance program is not
an input or consideration in any accident previously evaluated, thus
the proposed change will not increase the probability of any such
accident occurring. The proposed amendment does not involve any
change to the configuration or method of operation of any plant
equipment that is used to mitigate the consequences of an accident,
nor does it affect any assumptions or conditions in the accident
analysis. As a result, the consequences of any accident previously
evaluated are not significantly increased.
Therefore, operation of the facility in accordance with the
proposed change will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed amendment does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in methods governing normal plant operation. No equipment
interfaces are modified and no changes to any equipment function or
the method of operating the equipment are being made. The proposed
change, to relocate the TS Surveillance and implement an integrated
surveillance program, maintains an equivalent level of RPV [reactor
pressure vessel] material surveillance and does not introduce any
new accident initiators. The proposed change will not change the
design, configuration or operation of the plant.
Therefore, operation of the facility in accordance with the
proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed change will not involve a significant reduction
in a margin of safety.
The proposed amendment has been evaluated as providing an
acceptable alternative to the plant-specific RPV material
surveillance program that meets the requirements of the regulations
for RPV material surveillance. The proposed change does not exceed
or alter a design basis or safety limit. The change relocates a TS
Surveillance Requirement and implements an integrated surveillance
program and as such does not significantly reduce the margin of
safety.
Therefore, operation of the facility in accordance with the
proposed change does not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Section Chief: L. Raghavan.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: September 23, 2002.
Description of amendment request: The proposed amendments would
change the SSES 1 and 2 Technical Specifications (TSs) by revising
limiting condition for operation (LCO) 3.6.2.3 to add a new Condition
B, which permits both residual heat removal (RHR) suppression pool
cooling subsystems to be inoperable for 8 hours, rather than
immediately initiating a unit shutdown. By making this change, the
licensee is incorporating Technical Specifications Task Force change
traveler number 230 into its TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
[[Page 66013]]
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
The proposed change relaxes the Required Actions of [LCO
3.6.2.3] by allowing 8 hours to restore one RHR suppression pool
cooling subsystem to OPERABLE status when both subsystems have been
determined to be inoperable. Required Actions and their associated
Completion Times are not initiating conditions for any accident
previously evaluated. The proposed 8 hour Completion Time provides
some time to restore required subsystem(s) to OPERABLE status, yet
is short enough that operating an additional 8 hours is not a
significant risk. Consequently, this change in Required Actions does
not significantly increase the probability of occurrence of any
accident previously evaluated. The Required Actions in the proposed
change have been developed to provide assurance that appropriate
remedial actions are taken in response to the degraded condition,
considering the operability status of the RHR Suppression Pool
Cooling System and the capability of minimizing the risk associated
with continued operation. As a result, the consequences of any
accident previously evaluated are not significantly increased.
Therefore, this 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 modification or
alteration of plant equipment (no new or different type of equipment
will be installed) or a change in the methods governing normal plant
operation. The Required Actions and associated Completion Times in
the proposed change have been evaluated to ensure that no new
accident initiators are introduced. Thus, this 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 relaxed Required Actions do not involve a significant
reduction in a margin of safety. The proposed change has been
evaluated to minimize the risk of continued operation with both RHR
suppression pool cooling subsystems inoperable. The operability
status of the RHR Suppression Pool Cooling System, a reasonable time
for repair or replacement of required features, and the low
probability of a design basis accident occurring during the repair
period have been considered in the evaluation. Therefore, this
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: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: Richard J. Laufer.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: April 16, 2002.
Description of amendment request: The proposed change would revise
the Technical Specifications to delete the primary containment
isolation valves and instrumentation associated with the permanent
removal of the reactor vessel head spray piping.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does not involve a significant increase in the probability or
consequences of an accident previously analyzed?
Response: No.
The proposed changes to Technical Specification Tables 3.3.2-1,
3.3.7.4-2, 3.4.3.2-1, and 3.6.3-1 do not involve a change in
structures, systems, or components that would affect the probability
or consequences of any accident previously evaluated in the Hope
Creek Updated Final Safety Analysis Report.
The proposed changes involve eliminating piping and valves
associated with the reactor head spray. The reactor head spray
system was initially provided to cool down the steam dryer and
separator during shutdown. The head spray system is not credited for
the prevention or mitigation of any accident. Therefore, neither the
offsite or control room radiological consequences are affected. The
head spray piping removal and addition of a bolted flange on the
reactor coolant pressure boundary enhances plant safety by
eliminating a source of pipe whip and potential leakage. In
addition, the drywell penetration will be capped and welded closed.
This will maintain primary containment integrity and will be
periodically tested in conjunction with the containment integrated
leak rate test.
Therefore, as discussed above, this modification does not
involve a significant increase in the probability or consequences
from any accident previously analyzed.
2. Does not create the possibility of a new or different kind of
accident from any accident previously analyzed?
Response: No.
The proposed changes to Technical Specification Tables 3.3.2-1,
3.3.7.4-2, 3.4.3.2-1, and 3.6.3-1 do not involve a change in
structures, systems, or components that would create a new or
different kind of accident from any accident previously evaluated in
the Hope Creek Updated Final Safety Analysis Report.
The proposed change to eliminate the head spray piping and the
addition of a bolted flange on the reactor coolant pressure boundary
enhances plant safety by eliminating a source of pipe whip and
potential leakage. In addition, the drywell penetration will be
capped and welded closed. This will maintain primary containment
integrity and will be tested in conjunction with the containment
integrated leak rate test.
Therefore, as discussed above, this modification does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does not involve a significant reduction in [a] margin of
safety?
Response: No.
The proposed change to delete the head spray valves from Tables
3.3.2-1, 3.3.7.4-2, 3.4.3.2-1, and 3.6.3-1 does not reduce any
margin of safety as defined in the Technical Specifications or
Bases. The bolted flange that will be installed on the head spray
penetration will maintain the integrity of the reactor coolant
pressure boundary. This flange would then be tested as part of the
reactor pressure vessel hydrostatic test. In addition, the drywell
penetration will be capped and welded closed. This will maintain
primary containment integrity and will be tested as part of the
containment integrated leak rate test.
Accordingly, based on the above, 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, Docket Nos. 50-348 and 50-364,
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama
Date of amendment request: November 7, 2001.
Description of amendment request: The proposed amendments would
remove license condition 2.C.3.f from the Unit 1 operating license and
license condition 2.C.4 from the Unit 2 operating license, and replace
them with a commitment in Section 9.1.4.2.2.5 of the Updated Final
Safety Analysis Report (UFSAR). Specifically, license conditions
2.C.3.f and 2.C.4 to FOLs NPF-2 and NPF-8, respectively, require NRC
approval of the lifting devices which attach the spent fuel cask to the
crane prior to use of the spent fuel cask crane for the purpose of
moving
[[Page 66014]]
spent fuel casks. Subsequent to issuance of FOLs NPF-2 and NPF-8, the
NRC issued NUREG-0612, ``Control of Heavy Loads at Nuclear Power
Plants,'' which endorsed the use of ANSI N14.6 for the design and
inspection of special lift devises thereby eliminating the need for
license conditions 2.C.3.f and 2.C.4. Accordingly, SNC proposes that
license conditions 2.C.3.f and 2.C.4 be removed from FOLs NPF-2 and
NPF-8, respectively, and replaced with a commitment in the FNP UFSAR to
ANSI N14.6 for the design, fabrication, testing, and quality assurance
requirements associated with the spent fuel cask lift device.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change replaces license conditions 2.C.3.f and2.C.4
to FOLs NPF-2 and NPF-8, respectively, with a commitment in the FNP
Updated Final Safety Analysis Report (UFSAR) to the requirements of
ANSI N14.6, as clarified by NUREG-0612, for the design, fabrication,
testing, maintenance, and quality assurance requirements applicable
to the spent fuel cask special lift device. The proposed change does
not involve a physical change to or require new or different
operability requirements for plant systems, structures, or
components. NUREG-0612, Control of Heavy Loads at Nuclear Power
Plants, provides methods acceptable to the NRC for assuring the safe
handling of heavy loads. NUREG-0612 endorses the use of ANSI N14.6
for the design, fabrication, testing, maintenance, and quality
assurance requirements applicable to special lifting devices used to
handle heavy loads in the proximity of safe shutdown equipment and
irradiated spent fuel, thereby eliminating the need for license
conditions 2.C.3.f and 2.C.4 to FOLs NPF-2 and NPF-8, respectively.
Accordingly, removal of license conditions 2.C.3.f and 2.C.4 to FOLs
NPF-2 and NFP-8, respectively, does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change replaces license conditions 2.C.3.f and
2.C.4 from FOLs NPF-2 and NPF-8, respectively, with a commitment in
the FNP UFSAR to the requirements of ANSI N14.6, as clarified by
NUREG-0612, for the design, fabrication, testing, maintenance, and
quality assurance requirements applicable to the spent fuel cask
special lift device. The proposed change does not involve: (1) A
physical change to plant systems, structures or components; or (2)
require new or different operability requirements for plant systems,
structures, or components. SNC's commitment to the guidance provided
in ANSI N14.6, as clarified by NUREG-0612, provides assurance that
the spent fuel cask special lift device, in conjunction with the use
of the single-failure proof spent fuel cask crane, will preclude the
possibility of a cask drop accident. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety?
The proposed change does not involve a physical change to the
plant or impact the operability requirements of systems, structures,
or components considered important to safety. As stated above, the
use of ANSI N14.6, as clarified by NUREG-0612, has been endorsed by
the NRC in NUREG-0612. The proposed change replaces license
conditions 2.C.3.f and 2.C.4 to FOLs NPF-2 and NPF-8, respectively,
with a commitment in the FNP UFSAR to the requirements of ANSI
N14.6, as clarified by NUREG-0612, for the design, fabrication,
testing, maintenance, and quality assurance requirements for the
spent fuel cask crane special lift device. 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: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Section Chief: John A. Nakoski.
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear (SQN)
Plant, Unit 1, Hamilton County, Tennessee
Date of amendment request: March 29, 2002 (TSC 02-02) as
supplemented by a letter dated October 10, 2002.
Description of amendment request: The proposed amendment deletes
several of the Unit 1 Technical Specification (TS) surveillance
requirements (SR) contained in TS 3/4.4.5, ``Steam Generators'' (SGs),
associated with the voltage-based SG alternative repair criteria (ARC).
In addition the proposed changes would delete License Condition 2.C.9.d
which references commitment letters associated with SG inspection
activities.
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. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Tennessee Valley Authority's [TVA's] proposed TS amendment does
not compromise limits associated with SG tube integrity. TVA's
proposed change removes existing SG tube plugging criteria (i.e.,
ARC) from the TS and reestablishes the standard TS criteria (40
percent through-wall criteria). This change is inherently more
conservative.
The proposed revision does not alter plant equipment, test
methods or operating practices. The proposed change continues to
provide controls for safe operation of SQN SGs within the required
limits. The proposed change does not contribute to events or
assumptions associated with postulated design basis accidents (i.e.,
SG tube rupture). The proposed change does not affect operator
indicators or actions required to diagnose or mitigate a SG tube
rupture accident. The proposed revisions continue to maintain the
required safety functions. Accordingly, the probability of an
accident or the consequences of an accident previously evaluated is
not increased.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
TVA's proposed amendment removes existing repair criteria and
incorporates the more conservative TS limit for SG tube plugging
(i.e., plug tubes with degradation depths equal to or greater than
40 percent through-wall). This change will not give rise to new
failure modes. The failure of a SG tube to maintain leakage
integrity during operation is an analyzed event in the SQN Updated
Final Safety Analysis Report. Accordingly, the proposed change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
TVA's proposed TS amendment is conservative with respect to the
margin of safety. The margin of safety is preserved through ensuring
structural integrity and leakage integrity of the SG tubes.
TVA's proposed change to remove ARC from the TS does not
compromise structural integrity or leakage integrity of SG tubes.
The proposed change invokes the standard TS tube plugging criteria
limit (40 percent through-wall criteria) which is inherently more
conservative.
The proposed change does not affect the plant conditions,
setpoints, or safety limits that could result in precursors to
accidents or degrade accident mitigation systems. Plant system
safety functions are not altered by the proposed change.
Consequently, the proposed TS revisions does not reduce the margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
[[Page 66015]]
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: September 6, 2002 (TS 00-14).
Brief description of amendments: The proposed amendments would
change the Sequoyah (SQN) Units 1 and 2 Technical Specification (TS) 3/
4.4.9.1, ``Pressure/Temperature [P-T] Limits, Reactor Coolant System''
and TS 3/4.4.12, ``Low Temperature Overpressure Protection [LTOP]
Systems.'' The proposed amendment provides two changes to the these
specifications as described below:
1. The proposed change relocates the information provided in these
TSs into a pressure temperature limit report (PTLR) format in
accordance with U.S. Nuclear Regulatory Commission (NRC) Generic Letter
(GL) 96-03, ``Relocation of the Pressure Temperature Limit Curves and
Low Temperature Overpressure Protection System Limits.''
2. The proposed change also upgrades these TSs to the standard TS
requirements for Westinghouse plants (NUREG-1431, Revision 2). In
addition, the Tennessee Valley Authority (TVA) proposed a change to SQN
TS 3/4.4.9.2, ``Pressurizer,'' to relocate the requirements of this TS
into the SQN Technical Requirements Manual (TRM).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), Tennessee Valley
Authority (TVA), the licensee, has provided its analysis of the issue
of no significant hazards consideration, which is presented below:
A. The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously evaluated.
The proposed revision does not affect plant equipment, test methods
or operating practices. The modification to SQN TSs is consistent with
the Standard Technical Specifications for Westinghouse Plants and
continues to provide controls for safe operation within the required
limits. The revised specifications provide appropriate administrative
controls for the RCS [reactor coolant system] P-T limits and LTOP
setpoints within the PTLR for future revisions as needed. The proposed
changes do not contribute to events or assumptions associated with
postulated design basis accidents (DBA). The proposed revisions
continue to maintain the required safety functions. Accordingly, the
probability of an accident or the consequences of an accident
previously evaluated is not increased.
B. The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The proposed revisions are not the result of changes to plant
equipment, test methods, or operating practices. The proposed revision
to the SQN RCS P-T limits, and LTOP setpoints continues to ensure that
conservative fracture toughness margins are maintained to protect
against reactor pressure vessel failure and overpressure conditions.
The modified P-T limits and LTOP setpoints are based on NRC approved
methodology in conjunction with alternative methods provided in ASME
Code Case N-640, ``Alternative Requirement Fracture Toughness for
Development of P-T Limit Curves for ASME [American Society of
Mechanical Engineers] Section XI, Division 1'' and WCAP-15315,
``Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for
Operating PWR [pressurized water reactor] and BWR [boiling water
reactor] Plants.''
The proposed changes to incorporate the PTLR format is
administrative in nature and provide controls for maintaining RCS P-T
limits and LTOP setpoints for future revisions as needed.
The reactor vessel P-T limits and LTOP setpoints are operational
limits and are not considered to be contributors to the generation of
postulated accidents. The safety functions of the associated systems
remain unchanged and do not affect the assumptions of DBAs. The
operational limits and setpoints continue to be governed within the
TSs/PTLR. Accordingly, the proposed changes do not create the
possibility of a new or different kind of accident.
C. The proposed amendment does not involve a significant reduction
in a margin of safety.
TVA's proposed TS amendment provides revised reactor pressure
vessel P-T limits and LTOP setpoints that are within the design
capabilities of the RCS Safety Structures, Systems and Components (SSC)
and pressure control systems. The limits are based on conservative
design margins that ensure that plant operation is within the design
capacity of the reactor vessel materials. Accordingly, the function of
the RCS to provide a fission product barrier is not compromised.
TVA's proposed change to include revised P-T and LTOP limits does
not result in a change to system design features. The proposed change
does not affect plant conditions that result in precursors to accidents
or cause degradation of accident mitigation systems. The plant system
safety functions are not altered by the proposed change.
The proposed changes to the P-T limits and LTOP setpoints change
the calculations and method from that described in the current TS Bases
to one based on ASME Code Case N-640 and WCAP-15315. The effect of this
change is to allow plant operation with different limits while
continuing to retain conservative margins for assuring integrity of the
reactor vessel and the RCS. Consequently, the proposed TS revisions do
not significantly reduce the margin of safety.
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: July 9, 2002.
[[Page 66016]]
Brief description of amendment: The proposed amendment would revise
the Technical Specifications to remove the cycle-specific allowances on
(1) Rod insertion limits during individual rod position indicator
channel calibrations and (2) rod position indicator channel accuracy
for operation at or below 50 percent power. The proposed amendment also
would revise the control rod indicated misalignment limits.
Date of publication of individual notice in Federal Register:
October 7, 2002 (67 FR 62500).
Expiration date of individual notice: November 6, 2002.
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: October 1, 2002.
Brief description of amendments: The amendments revise the
licensing basis as described in the Updated Final Safety Analysis
Report (UFSAR) to allow lifting heavier loads with the reactor building
crane during the Unit 1 refueling outage beginning in November 2002.
Date of publication of individual notice in Federal Register:
October 4, 2002 (67 FR 62270).
Expiration date of individual notice: November 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 email 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 11, 2001, as
supplemented on June 27 and September 19, 2002.
Brief description of amendment: The amendment revised the Technical
Specifications, Section 3.9, ``Refueling,'' and its corresponding bases
to permit the continuation of core alterations during refueling
operations with the refueling interlocks inoperable by providing
alternate actions which will preserve the intended design function of
the inoperable interlocks.
Date of Issuance: October 10, 2002.
Effective date: October 10, 2002, and shall be implemented within
30 days of issuance.
Amendment No.: 234.
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 June 27 and September 19, 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 October 10,
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 (HBRSEP2), Darlington County, South Carolina
Date of application for amendment: March 13, 2002, as supplemented
May 10, August 14, September 5, September 23, and October 4, 2002.
Brief description of amendment: The amendment revises the Technical
Specifications (TS) for HBRSEP2 to permit selective implementation of
alternative radiological source term and modify the TS requirement for
movement of irradiated fuel and performing core alterations.
Date of issuance: October 4, 2002.
Effective date: October 4, 2002.
Amendment No.: 195.
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 10, August 14, September 5, September 23, and October
4, 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 October 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: January 31, 2002, as
supplemented by letters dated June 12, June 25, July 22, September 16,
and October 2, 2002.
Brief description of amendment: This amendment increases the
licensed power level by approximately 1.7%, from 3,833 megawatts
thermal (MWt) to 3,898 MWt. These changes result from increased
feedwater flow measurement accuracy to be achieved by utilizing high
accuracy ultrasonic flow measurement instrumentation.
Date of issuance: October 10, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 156.
Facility Operating License No. NPF-29: The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 2, 2002 (67 FR
15622).
[[Page 66017]]
The June 12, June 25, July 22, September 16, and October 2, 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 October 10, 2002.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: May 30, 2001.
Description of amendment request: The amendment revises the Cooper
Nuclear Station's Technical Specifications (TS) 5.5.7, ``Ventilation
Filter Testing Program (VFTP),'' reflecting a correction of an
erroneous reference to American Society of Mechanical Engineers N510-
1980.
Date of issuance: September 30, 2002.
Effective date: The amendment is effective on the date of issuance,
to be implemented within 30 days from the date of issuance.
Amendment No.: 195.
Facility Operating License No. DPR-46: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 5, 2001 (66
FR 46480). The Commission related evaluation of the amendment is
contained in a Safety Evaluation dated September 30, 2002.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: June 28, 2002, as supplemented
on August 15, August 16, and October 2, 2002.
Brief description of amendments: The amendments change the Salem
Technical Specifications (TS) requirements for Fuel Decay Time prior to
commencing movement of irradiated fuel. TS Limiting Condition for
Operation 3/4.9.3, ``Decay Time,'' is revised to allow fuel movement in
the containment to commence 100 hours after the reactor has become
subcritical between October 15th through May 15th. Should refueling
occur between May 16th and October 14th, the current 168 hours decay
time limit will remain in place. These requirements are valid through
the year 2010.
Date of issuance: October 10, 2002.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment Nos.: 251 and 232.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 30, 2002 (67 FR
55887). The August 15, August 16, and October 2, 2002, letters provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated October 10, 2002.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: July 18, 2002.
Brief description of amendments: These amendments change the Salem
Technical Specifications (TSs) requirements associated with its
containment spray nozzles. The frequency of TS Surveillance Requirement
(SR) 4.6.2.1.d for verifying that the containment spray nozzles are
unobstructed is changed from a fixed 10-year frequency to after
activities that could result in nozzle blockage. In this case, PSEG
will be required to evaluate the work performed to determine the impact
to the containment spray system, or perform an air or smoke flow test.
The applicable Bases pages are also revised to reflect this change.
Date of issuance: October 10, 2002.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment Nos.: 252 & 233.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 20, 2002 (67 FR
53989). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 10, 2002.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: November 1, 2001, as
supplemented on October 1, 2002.
Brief description of amendments: The changes modify the provisions
under which equipment may be considered operable when either its normal
or emergency power source is inoperable. Technical Specifications (TS)
Section 3.0.5 was deleted and additional limiting conditions for
operation were incorporated into electrical power systems TS 3.8.1.1,
``A.C. Sources--Operating.'' The corresponding TS Bases were modified
accordingly. The proposed changes are consistent with the
recommendations contained in NUREG-1431, Rev. 2, ``Standard Technical
Specifications for Westinghouse Plants.''
Date of issuance: October 11, 2002.
Effective date: October 11, 2002.
Amendment Nos.: 253 and 234.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 5, 2002 (67 FR
5331). The October 1, 2002 supplement 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 amendments is contained in a Safety
Evaluation dated October 11, 2002.
No significant hazards consideration comments received: No.
Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco
Nuclear Generating Station, Sacramento County, California
Date of application for amendment: February 20, 2001.
Brief description of amendment: The amendment eliminates the
security plan requirements from the 10 CFR Part 50 licensed site after
the spent nuclear fuel has been transferred to the 10 CFR Part 72
licensed Independent Spent Fuel Storage Installation and is based in
part on exemptions from specific requirements set forth in 10 CFR Part
73 and 10 CFR 50.54(p).
Date of issuance: October 10, 2002.
Effective date: October 10, 2002, to be implemented within 30 days.
Amendment No.: 131.
Facility Operating License No. DPR-54: The amendment revised the
Operating License and the Technical Specifications.
Date of initial notice in Federal Register: March 21, 2001 (66 FR
15930). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 10, 2002.
No significant hazards consideration comments received: No.
[[Page 66018]]
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: July 25, 2002.
Brief description of amendment: The amendment revises Surveillance
Requirements (SRs) 3.3.1.2 and 3.3.1.3 of the technical specifications
on the reactor trip system (RTS) instrumentation. The change to SR
3.3.1.2 replaces the reference to the nuclear instrumentation system
channel output by a reference to the power range channel output and
deletes Note 1 to the SR. The change to SR 3.3.1.3 is editorial.
Date of issuance: October 2, 2002.
Effective date: October 2, 2002, and shall be implemented within 6
months of the date of issuance, including the incorporation of changes
to the Technical Specification Bases as described in the licensee's
application dated July 25, 2002.
Amendment No.: 148.
Facility Operating License No. NPF-42: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 20, 2002 (67 FR
53992). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 2, 2002.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, located at One White Flint North, 11555 Rockville Pike (first
floor), Rockville, Maryland. Publicly available records will be
accessible from the Agencywide Documents Assess and Management System's
(ADAMS) Public Electronic Reading Room on the Internet at the NRC web
site, http://www.nrc.gov/reading-rm/adams.html. If you do not have
access to ADAMS or if there are problems in accessing the documents
located in ADAMS, contact the NRC Public Document Room (PDR) Reference
staff at 1-800-397-4209, 304-415-4737 or by email to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By November 29, 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
[[Page 66019]]
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,\2\ which is available at the Commission's PDR, located at
One White Flint North, 11555 Rockville Pike (first floor), Rockville,
Maryland, and electronically on the Internet at the NRC web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1-800-397-4209, 301-415-4737, or by e-mail to
[email protected]. If a request for a hearing or petition for leave to
intervene is filed by the above date, the Commission or an Atomic
Safety and Licensing Board, designated by the Commission or by the
Chairman of the Atomic Safety and Licensing Board Panel, will rule on
the request and/or petition; and the Secretary or the designated Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
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\2\ ``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 petitoner's interest.
(2) The admissibility of a contention, refuse to admit a
contention if:
(i) The contention and supporting material fail to satisfy the
requirements of paragraph (b)(2) of this section; or
(ii) The contention, if proven, would be of no consequence in
the proceeding because it would not entitle petitioner to relief.''
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As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, 11555 Rockville Pike (first floor),
Rockville, Maryland, by the above date. Because of the continuing
disruptions in delivery of mail to United States Government offices, it
is requested that petitions for leave to intervene and requests for
hearing be transmitted to the Secretary of the Commission either by
means of facsimile transmission to 301-415-1101 or by e-mail to
[email protected]. A copy of the petition for leave to intervene
and request for hearing should also be sent to the Office of the
General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and because of continuing disruptions in delivery of mail
to United States Government offices, it is requested that copies be
transmitted either by means of facsimile transmission to 301-415-3725
or by e-mail to [email protected]. A copy of the request for
hearing and petition for leave to intervene should also be sent to the
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: September 26, 2002.
Description of amendment request: The amendments consist of a one-
time change to the Dresden Updated Final Safety Analysis Report (UFSAR)
to state that lifting heavy loads up to and including 116 tons is
allowed prior to and during the upcoming Dresden Unit 3 refueling
outage number 17.
Date of issuance: October 4, 2002.
Effective date: Immediately, to be implemented within 30 days.
Amendment No.: 196 and 189.
Facility Operating License Nos. DPR-19 and DPR-25: Amendment
revises the UFSAR.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. Joliet Herald News, dated October 1, 2002.
The notice provided an opportunity to submit comments on the
Commission's proposed NSHC determination. No comments have been
received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a Safety Evaluation dated October 4, 2002.
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.
Dated at Rockville, Maryland, this 18th day of October 2002.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 02-27243 Filed 10-28-02; 8:45 am]
BILLING CODE 7590-01-P