[Federal Register Volume 66, Number 229 (Wednesday, November 28, 2001)]
[Notices]
[Pages 59498-59518]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-29446]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
Note: The publication date for this notice will change from
every other Wednesday to every other Tuesday, effective January 8,
2002. The notice will contain the same information and will continue
to be published biweekly.)
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 November 5 through November 16, 2001. The
last biweekly notice was published on November 14, 2001 (66 FR 57116).
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
[[Page 59499]]
within 30 days after the date of publication of this notice will be
considered in making any final determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC Public Document
Room, 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 December 28, 2001, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714, which is
available at the NRC's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852.
Publicly available records will be accessible electronically 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/NRC/ADAMS/index.html. If a request for a hearing or
petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Branch, or may be delivered to the Commission's Public
Document Room, located at One White Flint North, 11555 Rockville Pike
(first floor), Rockville, Maryland 20852, by the above date. A copy of
the petition should also be sent to the Office of the General Counsel,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to
the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for
[[Page 59500]]
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 Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/NRC/ADAMS/index.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].
AmerGen Energy Company, LLC,. et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: April 4, 2001, as supplemented on
October 12, 2001.
Description of amendment request: The proposed amendment request
would delete Technical Specifications (TSs) 5.3.1.B and 5.3.1.C. These
TSs restrict the handling of heavy loads over irradiated fuel stored in
the storage pool. The basis for deleting these TSs is the upgrade of
the reactor building crane and associated handling systems to a single-
failure proof system.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The staff's review is
presented below:
The proposed amendment does not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Until August 2000, the reactor building crane was not single-
failure-proof. For heavy load handling associated with the spent fuel
pool, Oyster Creek was consistent with Section 5.1.4(2) of NUREG-0612:
``The effects of heavy load drops in the reactor building should be
analyzed to show that the evaluation criteria of Section 5.1 are
satisfied.'' An alternative to this is Section 5.1.4(1): ``The reactor
building crane, and associated lifting devices used for handling of
heavy loads, should satisfy the single-failure-proof guidelines of
Section 5.1.6 of this report.'' The upgraded crane and handling systems
satisfy the guidelines of Section 5.1.6. Therefore, the licensing basis
for the reactor building crane with regard to its use in handling heavy
loads above the spent fuel storage pool is being revised to include
Section 5.1.4(1) of NUREG-0612 in addition to 5.1.4(2).
The cask drop protection system was required with the original
crane because the load drop analysis will yield unacceptable
consequences to the spent fuel storage pool (SFSP) structure. The cask
drop protection system (CDPS) serves to mitigate the consequences of a
cask drop accident involving the original crane which complied with
NUREG-0612 Phase I. The upgraded single-failure-proof crane satisfies
the criteria of NUREG-0612 Section 5.1.6. Therefore, the reactor
building crane eliminates reliance on the design function of the CDPS
because the probability of a heavy load drop is very low.
With the proposed revisions to the TSs, the evaluation criteria of
NUREG-0612, Section 5.1 is met with a single-failure-proof crane that
satisfies the guidelines of Section 5.1.6 or with consequence analyses
that satisfies Section 5.1.4(2).
The proposed TS revisions do not significantly change the potential
for unacceptable consequences to the plant in conducting heavy load
handling above the SFSP because the probability of a load drop accident
caused by use of the reactor building crane has been reduced to where
it is very unlikely, and therefore, can be considered not credible
within regulatory accepted standards.
Therefore, the proposed TS revisions do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
(2) Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The CDPS was installed in the Oyster Creek SFSP to mitigate the
effects of a cask drop when the reactor building crane was not single-
failure proof. The CDPS acts as a hydraulic dashpot to limit the
velocity of a falling cask to attenuate impact forces to within
acceptable levels. The CDPS structure cannot be removed from the spent
fuel pool without eliminating its functional requirement. The use of
the CDPS increases the duration of cask lifts and exposure to
personnel. Therefore, eliminating the complications caused by the use
of the CDPS together while improving the reliability of the crane and
associated systems does not create the possibility of a new or
different kind of accident.
Therefore, operation of the facility in accordance with the
proposed license amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
(3) Involve a significant reduction in a margin of safety.
The proposed TS change will remove the load limit over the SFSP and
CDPS restrictions when the reactor building crane is used with single-
failure-proof handling systems that comply with criteria in Section
5.1.6 of NUREG-0612.
The reactor building crane was upgraded to single-failure-proof in
compliance with NUREG-0554. The upgraded crane and handling system is
in compliance with NUREG-0612, Sections 5.1.1 and 5.1.6. The NRC in
NUREG-0612, Section 5.2 documented their review of the potential
consequences of a load drop when handled by a single-failure-proof
crane using single-failure-proof rigging compared with other
alternatives and concluded as follows:
``The likelihood for unacceptable consequences in terms of
excessive releases of gap activity or potential for criticality due to
accidental dropping of postulated heavy loads after Receptionist (OWFN
and TWFN) implementation of the guidelines of Section 5.1 is very
low.''
Therefore, there is a very minimal chance of a load drop that could
result in consequences that exceed the regulatory accepted standards
when the load is handled by a single-failure-proof crane and handling
system, and performed in accordance with Section 5.1 of NUREG-0612. A
single-failure-proof crane design incorporates the applicable design
basis event that in this case is a seismic event. A load drop is of
such low probability that it is considered unlikely when it is handled
with the reactor building crane because the crane and its handling
systems satisfy the NUREG-0612 criteria for a single-failure-proof
crane. Therefore, any load lifted over the SFSP using the reactor
building crane, and adhering to NUREG-612 Phase I guidelines has a very
low probability of falling into the spent fuel pool accidentally or as
a result of a design basis event.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
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: Kevin P. Gallen, Morgan, Lewis & Bockius,
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: L. Raghavan, Acting.
[[Page 59501]]
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: September 10, 2001.
Description of amendment request: The proposed change would revise
the requirement for the source range monitor (SRM) operability during
core operations. The proposed change would require two SRM channels to
be operable, one with its detector located in the core quadrant where
core alterations are being performed, and another with its detector
located in an adjacent quadrant.
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 TS [Technical Specification] change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed change revises Technical Specification 3.9.D for
source range monitor operability requirements during core
alterations. The only accident described in the Final Safety
Analysis Report (FSAR) while the plant is in Cold Shutdown or
Refueling is a fuel handling (dropped bundle) accident. The proposed
change involves equipment that is not involved in the mitigation or
prevention of a fuel handling accident as described in FSAR.
Therefore, the change to SRM operability requirements does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change to TS 3.9.D does not involve any physical
alteration of plant equipment or system configuration. Core
reactivity and reactivity control functions are not affected, and
adequate reactivity monitoring capability is maintained. Therefore,
the proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
The proposed change to TS 3.9.D affects the operability
requirements for source range monitors during core alterations. The
SRMs do not perform any required functions for mitigating the
consequences of an accident. The current specification only requires
one operable SRM. The proposed specification will ensure redundant
monitoring is available to detect changes in the reactivity
condition of the core by requiring the operability of at least two
source range monitors. This will provide adequate capability for
detecting an inadvertent criticality. 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: Kevin P. Gallen, Morgan, Lewis & Bockius,
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: L. Raghavan, Acting.
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: September 11, 2001
Description of amendment request: The purpose of the proposed
revision to the Technical Specifications (TSs) is to delete the cycle-
specific footnote for the Safety Limit Minimum Power Critical Ratio
(SLMCPR).
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. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The derivation of the cycle specific SLMCPR limit for
incorporation into the Technical Specification, and its use to
determine cycle specific thermal limits, has been performed using
the methodology discussed in ``General Electric Standard Application
for Reactor Fuel,'' NEDE-24011-P-A-13, and Amendment 25. Amendment
25 was approved by the NRC in a Safety Evaluation Report dated March
11, 1999. The footnote to Technical Specification 2.1.A is being
deleted. The footnote associated with the Technical Specification
2.1.A was originally included to ensure that the SLMCPR was only
applicable for the identified cycle because Amendment 25 was not yet
NRC approved. Amendment 25 has subsequently been approved.
Therefore, this footnote is no longer necessary. The footnote was
for information only, and has no impact on the design or operation
of the plant. Cycle-specific SLMCPR values will continue to be
developed in accordance with NRC approved methods, which ensures
that applicable regulatory requirements are met [. . .]
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed change deletes the footnote contained in Technical
Specification 2.1.A as the result of the NRC approval of Amendment
25 to NEDE-24011-P-A. This change does not affect the design or
operation of any plant structures, systems or components. Cycle-
specific SLMCPR values will continue to be developed in accordance
with NRC approved methods, which ensures that applicable regulatory
requirements are met. Changes to the SLMCPR value specified in the
Technical Specification will require prior NRC approval [. . .]
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed change deletes the footnote contained in Technical
Specification 2.1.A as the result of the NRC approval of Amendment
25 to NEDE-24011-P-A. Cycle-specific SLMCPR values will continue to
be developed in accordance with NRC approved methods as specified in
the Technical Specifications. These methods ensure that applicable
regulatory requirements are met. Changes to the SLMCPR value
specified in the Technical Specifications will require prior NRC
approval [. . .]
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: Kevin P. Gallen, Morgan, Lewis & Bockius,
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: L. Raghavan, Acting.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendments request: September 11, 2001.
Description of amendments request: The amendments would allow the
non-operating shutdown cooling loop to be declared inoperable for a
period up to 2 hours for surveillance testing in MODE 6. The request is
based on Technical Specification Task Force Traveler Number 361,
Revision 2.
[[Page 59502]]
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 amendment would add a note to the limiting
condition of operation (LCO) of Technical Specification 3.9.5,
Shutdown Cooling (SDC) and Coolant Circulation--Low Water Level,
that would permit one required SDC loop to be declared inoperable
for a period of up to 2 hours for surveillance testing, provided the
other SDC loop is OPERABLE and in operation.
Allowing the non-operating SDC loop to be declared inoperable in
accordance with the proposed amendment does not involve a
significant increase in the probability of an accident previously
evaluated because the SDC system is not an accident initiator of any
previously evaluated accidents. Because the SDC system does not
initiate any previously analyzed accidents, it cannot increase the
probability of these accidents occurring.
Furthermore, allowing the non-operating SDC loop to be declared
inoperable in accordance with the proposed amendment does not
involve a significant increase in the consequences of an accident
previously analyzed because only one operating SDC loop is necessary
to perform the SDC system function of removing decay heat from the
reactor core.
The proposed amendment does not represent a change to the design
of the facility. Nor does the proposed amendment prevent the safety
function of the shutdown cooling system from being performed. The
proposed amendment does not alter, degrade, or prevent actions
described or assumed in any accident described in the PVNGS Updated
Final Safety Analysis Report (UFSAR) from being performed.
Therefore, since the SDC system is not an accident initiator and
because only one SDC loop is necessary to perform the design
function, the proposed amendment would not significantly increase
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 amendment would add a note to the limiting
condition of operation (LCO) of Technical Specification 3.9.5,
Shutdown Cooling (SDC) and Coolant Circulation--Low Water Level,
that would permit one required SDC loop to be declared inoperable
for a period of up to 2 hours for surveillance testing, provided the
other SDC loop is OPERABLE and in operation. Allowing the non-
operating SDC loop to be declared inoperable in accordance with the
proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated
because: (1) The proposed amendment does not represent a change to
the design of the plant, (2) the proposed amendment does not involve
the installation of new or different equipment, (3) the proposed
amendment does not alter the methods for operating plant equipment,
and (4) the proposed amendment does not affect any other safety
related equipment. Therefore, the proposed amendment does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed amendment would add a note to the limiting
condition of operation (LCO) of Technical Specification (TS) 3.9.5,
Shutdown Cooling (SDC) and Coolant Circulation--Low Water Level,
that would permit the non-operating SDC loop to be declared
inoperable for a period of up to 2 hours for surveillance testing in
MODE 6, when the water level is less than 23 feet above the top of
the reactor vessel flange, provided the other SDC loop is OPERABLE
and in operation. Allowing the non-operating SDC loop to be declared
inoperable in accordance with the proposed amendment does not
involve a significant reduction in a margin of safety because the
operating SDC loop provides sufficient decay heat removal capacity.
The proposed change does not impact the operating SDC loop. In the
unlikely event that the operating SDC loop becomes inoperable
concurrent with the inoperability of the non-operating SDC loop
allowed by the proposed note, adequate controls exist within the TS
3.9.5 Required Actions to ensure adequate decay heat removal. In
addition, if the operating SDC loop fails, operator action to
restore the SDC loop being tested to OPERABLE status and place that
SDC loop in operation will be timely such that adequate decay heat
removal capability is maintained. Therefore, the proposed amendment
does not involve a significant reduction in a margin of safety.
Based on the responses to these three criteria, Arizona Public
Service Company (APS) has concluded that the proposed amendment
involves no significant hazard consideration.
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.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: November 7, 2001.
Description of amendments request: The proposed license amendments
would revise Technical Specification (TS) 3.1.4, ``Control Rod Scram
Times'' and TS 5.5.10, ``Technical Specifications Bases Control
Program.'' TS 3.1.4 would be revised to better delineate the
requirements for testing control rod scram times following refueling
outages. TS 5.1.10 would be revised to reference Title 10 of the Code
of Federal Regulations (10 CFR) Section 50.59. This license amendment
application incorporates the NRC-approved Technical Specification Task
Force (TSTF) Item 222, Revision 1, ``Control Rod Scram Time Testing,''
and TSTF Item 364, Revision 0, ``Revision to TS Bases Control Program
to Incorporate Changes to 10 CFR 50.59.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed license amendments do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change to adopt TSTF-222, Revision 1, is an
administrative clarification of existing Technical Specification
requirements regarding scram time testing requirements for control
rods. The current wording of Surveillance Requirement 3.1.4.1
requires each control rod to be tested if any fuel movement occurs
in the reactor pressure vessel. Surveillance Requirements 3.1.4.3
and 3.1.4.4 require only the affected control rods to be tested. The
NRC-approved TSTF-222, Revision 1, clarifies that post-refueling
scram time testing of control rods only applies to control rods
affected by work activities. The requirement to test all control
rods following routine refueling outages remains unchanged. As such,
there is no effect on initiators of analyzed events or assumed
mitigation of accidents or transients. Therefore, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to adopt TSTF-364, Revision 0, is an
administrative change to provide consistency between the Technical
Specification requirements for the Technical Specification Bases
Control Program and the regulatory requirements of Title 10, Section
50.59 of the Code of Federal Regulations, as revised by the NRC on
October 4, 1999. The change will have no affect on the initiators of
analyzed events or assumed mitigation of accidents or transients.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed license amendments will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
[[Page 59503]]
The proposed changes to adopt TSTF-222, Revision 1 and TSTF-364,
Revision 0, do not involve a physical alteration of the plant, add
any new equipment, or require any existing equipment to be operated
in a manner different from the present design. Therefore, the
proposed changes do not create a new or different kind of accident
from any accident previously evaluated.
3. The proposed license amendments do not involve a significant
reduction in a margin of safety.
The proposed change to adopt TSTF-222, Revision 1, will not
reduce a margin of safety because it has no effect on any safety
analysis assumptions. The proposed license amendment implements an
administrative clarification to better delineate the requirements
for scram time testing control rods following refueling outages and
for control rods requiring testing due to work activities. The
requirement to test all control rods following a routine refueling
outage remains unchanged. As such, the proposed change does not
involve a significant reduction in the margin of safety.
The proposed change to adopt TSTF-364, Revision 0, is an
administrative change to provide consistency between the Technical
Specification requirements for the Technical Specification Bases
Control Program and the regulatory requirements of Title 10, Section
50.59 of the Code of Federal Regulations, as revised by the NRC on
October 4, 1999. The change will not reduce the margin of safety
because the change has no effect on any safety analyses assumptions.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Richard P. Correia.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: August 24, 2001.
Description of amendment request: The proposed amendment would
delete the Technical Specification (TS)-required action which, in the
event of inoperability of the oscillation power range monitor (OPRM)
trip function, limits plant operation above 25-percent power to 120
days. Instead, continued plant operation would be allowed if a TS-
required action is taken to implement an alternate method to detect and
suppress thermal-hydraulic instability oscillations.
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 OPRM function is not considered as an initiator of any
previously analyzed accident. Therefore, this proposed change does
not significantly increase the probability of such accidents. This
proposed change would allow the use of existing well-established
alternate methods to detect and suppress the thermal hydraulic
instability oscillations. Considering that multiple Boiling Water
Reactor plants, including Fermi 2, have satisfactorily operated
using alternate stability monitoring methods for extended periods of
time prior to the installation of OPRM systems, it is concluded that
these measures are adequate. Therefore, the consequences of a
previously analyzed accident would not be significantly increased.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed amendment does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The change does not involve a significant reduction in the
margin of safety.
This proposed change would allow the use of an existing
alternate method to detect and suppress thermal hydraulic
instability oscillations to continue to operate the reactor above
25% power in the event of the inoperability of the OPRM system.
Considering that multiple Boiling Water Reactor plants, including
Fermi 2, have satisfactorily operated using alternate stability
monitoring methods for extended periods of time, it is concluded
that these measures are adequate, and that the proposed change does
not significantly reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB,
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
NRC Section Chief: William D. Reckley.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: October 23, 2001.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) surveillance requirement (SR)
3.8.4.1 to change limits for the battery terminal voltage when on a
float charge for 125 VDC station battery 31 following the replacement
of this battery in early 2002. The proposed amendment would also revise
the applicable TS Bases section.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed License Amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
No. The proposed TS SR change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The newly installed battery 31 will consist of
59 cells, instead of the presently installed 58-cell battery. An
additional cell will be added to 31 Battery in order to provide an
acceptable design margin for future load addition to this battery.
The resulting change in the minimum 31 Battery terminal voltage
on float charge to 125.7 V is due to the additional cell added. This
new value will ensure that the 31 Battery is properly verified to be
functional to meet its design requirements. Calculations
demonstrated in IP3-ECCF-845 indicate that 31 Battery DC circuit
coordination is not affected by the proposed replacement of the
existing battery with a 59-cell battery. The proposed TS SR change
does not affect accident initiators or precursors, nor do they alter
design assumptions for the systems or components used to mitigate
the consequences of an accident as analyzed in Chapter 14 of the IP3
UFSAR [Indian Point 3 Updated Final Safety Analysis Report].
2. Does the proposed License Amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
No. This TS SR change for 31 Battery is based upon replacement
of the 31 Battery with a new 59-cell battery. This new battery 31 is
at least equivalent to the existing 58-cell 31 Battery. This new 31
battery, with the added cell, provides an acceptable design margin
to the 31 Battery. Battery 31 circuit coordination is not adversely
affected by the addition of this new battery with 59 cells. The
proposed changes to this TS SR do not introduce any new accident
initiators or precursors, or any new design assumptions for those
components used to mitigate the consequences of an accident.
[[Page 59504]]
3. Does the proposed License Amendment involve a significant
reduction in a margin of safety?
No. During the replacement of the existing 31 battery with a new
59-cell battery and the subsequent TS SR change that verifies higher
minimum terminal voltage on float charge, the new 31 battery and the
requirements associated with verifying its design functionality will
not involve a significant reduction in the margin of safety. The
replacement 31 Battery is at least equivalent to the existing
battery. The additional cell in the proposed new 59-cell battery
provides an acceptable design margin, which will be 120% for 31
battery with 59 cells. The increase in the number of cells from 58
to 59 will result in a higher 31 Battery terminal voltage on float
charge. This proposed TS SR simply documents the verification of
this new minimum voltage value. The minimum terminal voltage value
for the new 32 Battery will not change nor be impacted by this TS
change. Accordingly, there is no significant reduction in [a] margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: L. Raghavan (Acting).
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 15, 2001.
Description of amendment request: The proposed change to Technical
Specification 3.4.7 limits Reactor Coolant System activity permitted by
the ACTION statement to 60 microcuries per gram (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.
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 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 FSAR change 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 proposed FSAR change meets the original SER [Safety
Evaluation Report] acceptance criteria with the exception of the
Exclusion Area Boundary (EAB) accident induced iodine spiking
thyroid dose. The SRP [Standard Review Plan] acceptance criteria for
the EAB accident induced iodine spiking thyroid dose is a small
fraction of the 10 CFR [Part] 100 limits (30 rem). The proposed
change falls well within 10 CFR [Part] 100 limits (75 rem).
The EAB accident induced iodine spiking thyroid dose
consequences are considered acceptable and reasonable for the
following reasons:
The letdown line break event starting from the most
limiting parameters allowed by the TS LCO [Limiting Conditions for
Operation] on RCS activity, pressure, temperature, primary to
secondary leakage, and proceeding unmitigated for 30 minutes is
highly unlikely. The additional use of conservative assumptions such
as an iodine spiking factor of 500, maximum bounding letdown flow,
worst case 95 percentile atmospheric dispersion factors, flashing
fraction based on 560 deg.F even though the break flow would travel
through the regenerative heat exchanger and cool down, no activity
plate out, no ground deposition, and no activity decay in the
transit to the exclusion area boundary significantly increases the
overall conservative nature of the calculation.
Currently, FSAR Table 15.6-4 lists the 'Realistic' EAB
thyroid dose as 0.46 rem. The realistic dose is based upon no iodine
spike, 50 percentile X/Q [atmospheric dispersion factor], and 0.12%
failed fuel RCS activity. The best estimate dose consequences using
the new analysis methodology with the normal plant operating
parameters would remain below 0.46 rem even for the accident induced
iodine spiking event.
The new analysis accident induced iodine spiking
results would remain below the SRP acceptance criteria if any one of
the following normal plant operating parameters were used: RCS
steady state activity, iodine spiking factor, letdown flow, or
atmospheric dispersion factors.
The letdown line break consequences are considered acceptable
due to the unlikeliness of the event and conservative nature of the
analyses. The `no iodine spike' results remain within a small
fraction of the 10 CFR [Part] 100 limits; the `accident induced
iodine spike' results fall well within the 10 CFR [Part] 100 limits;
and the `pre-existing iodine spike' results are within the 10 CFR
[Part] 100 limits.
Therefore, this change does not involve a significant increase
in the probability or consequence 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 pre-existing iodine spike is not an
accident initiator and the FSAR change does not affect any plant
Structure, Systems, or Components (SSC) 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 plant levels.
The proposed FSAR change meets the original SRP acceptance
criteria with the exception of the Exclusion Area Boundary (EAB)
accident induced iodine spiking thyroid dose. The SRP acceptance
criteria for the EAB accident induced iodine spiking thyroid dose is
a small fraction of the 10 CFR [Part] 100 limits (30 rem). The
proposed change falls well within 10 CFR [Part] 100 limits (75 rem).
The EAB accident induced iodine spiking thyroid dose
consequences are considered not to be a significant reduction in the
margin of safety for the following reasons.
The letdown line break event starting from the TS LCO
on RCS activity, pressure, temperature, primary to secondary
leakage, and proceeding unmitigated for 30 minutes is highly
unlikely. The additional use of conservative assumptions such as an
iodine spiking factor of 500, maximum bounding letdown flow, worst
case 95 percentile atmospheric dispersion factors, flashing fraction
based on 560 deg.F even though the break flow would travel through
the regenerative heat exchanger and cool down, no activity plate
out, no ground deposition, and no activity decay in the transit to
the exclusion area boundary significantly increases the overall
conservative nature of the calculation.
The FSAR Table 15.6-4 lists the ``Realistic'' EAB
thyroid dose as 0.46 rem. The realistic dose is based upon no iodine
spike, 50 percentile X/Q, and 0.12% failed fuel RCS activity. The
best estimate dose consequences using the new analysis methodology
with the normal plant operating parameters would remain below 0.46
rem even for the accident induced iodine spiking event.
The new analysis accident induced iodine spiking
results would remain below the SRP acceptance criteria if any one of
the following normal plant operating parameters were used: RCS
steady state activity, iodine spiking factor, letdown flow, or
atmospheric dispersion factors.
[[Page 59505]]
The letdown line break consequences are considered acceptable
due to the unlikeliness of the event and conservative nature of the
analyses. The ``no iodine spike'' results remain within a small
fraction of the 10 CFR [Part] 100 limits; the ``accident induced
iodine spike'' results fall well within the 10 CFR [Part] 100
limits; and the ``pre-existing iodine spike'' results are within the
10 CFR [Part] 100 limits.
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. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois]
Date of amendment request: September 21, 2001.
Description of amendment request: The proposed amendment would
revise the Reactor Core Safety Limit (SL) for peak fuel centerline
temperature from less than or equal to 4700 deg.F (i.e., the current
TS limit) to the design basis fuel centerline melt temperature of less
than 5080 deg.F, for unirradiated fuel, decreasing by 58 deg.F per
10,000 Megawatt-Days per Metric Tonne Uranium (MWD/MTU) burnup.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The use of high burnup rods or assemblies will not increase the
probability of any accident previously evaluated. These high burnup
rods or assemblies will continue to satisfy all fuel mechanical,
nuclear, thermal-hydraulic, and transient analysis design criteria.
Fuel type is not directly related to the probability of any
previously evaluated accidents; however, adhering to applicable
design criteria and standards precludes challenges to components and
systems that could increase the probability of an accident. The high
burnup fuel rods will continue to satisfy the Specified Acceptable
Fuel Design Limits (SAFDLs) specified in the Westinghouse Topical
Report, WCAP-12488-A, ``Westinghouse Fuel Criteria Evaluation
Process,'' which was approved by the Nuclear Regulatory Commission
(NRC) on July 27, 1994. The clad integrity of the four high burnup
rods in the LTA will be maintained as the LTAs will be placed in
non-limiting core locations as permitted by TS 4.2.1 and will
continue to meet the safety parameter requirements. In addition, the
acceptability of using the four high burnup rods in an LTA is
evaluated in the Byron Station, Unit 2, Cycle 10 Reload Safety
Evaluation which is supported by Westinghouse Topical Report,
``Extended Burnup Operation Assessment for the VANTAGE+ Design in
Byron, Unit 2, Cycle 10,'' dated March 2001.
It has been shown in Westinghouse Topical Report, WCAP-12610-P-
A, ``VANTAGE+ Fuel Assembly Reference Core Report,'' approved by the
NRC in April 1995, that even though there are variations in core
inventories of isotopes due to extended burnup up to 75,000 MWD/MTU,
there are no significant increases of isotopes that are major
contributors to accident doses. It is worthy to note that, at higher
burnups, there is a reduction in certain isotopes that are major
dose contributors under accident situations (e.g., Kr-88). With only
four high burnup rods in the entire core, any variation of isotopes
will be extremely small. Thus, the radiation dose limitations of 10
CFR [Part] 100, ``Reactor Site Criteria,'' will not be exceeded.
The bases for establishing the fuel centerline melt temperature
are discussed in WCAP-12610-P-A, noted above, and implemented by
Westinghouse Topical Report WCAP-14483-A, ``Generic Methodology for
Expanded Core Operating Limits Report,'' approved by the NRC on
January 19, 1999. These methodologies and associated analyses
confirm that the present analytical limits for all accidents will be
maintained.
Based on this evaluation, it is concluded that the proposed TS
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Do the proposed TS changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
As required by WCAP-12488-A, the LTA with the four high burnup
rods must satisfy the five guidelines accepted by the NRC. These
guidelines are as follows:
Design of LTAs are mechanically and hydraulically
compatible with existing fuel
Peaking factors meet the TS limits
NRC approved/accepted safety/design methods and codes
are used
No SAFDLs are exceeded
Not more than eight LTAs per core are inserted
As previously noted, TS 4.2.1 allows the use of a limited number
of LTAs in nonlimiting core regions.
The use of high burnup rods or assemblies will comply with WCAP-
12488-A and TSs. All safety evaluations in support of using high
burnup rods or assemblies have been performed in accordance with
accepted methodologies.
In support of proposed High Burnup LTA Programs in the industry,
the NRC has requested fuel characterization inspections prior to
high burnup irradiation. LTA M09E, (i.e., the assembly containing
the high burnup fuel rods at Byron Station) was subjected to fuel
characterization inspections prior to operation in Byron Station,
Unit 2, Cycle 10. These inspections included assembly growth, rod
growth, assembly bow, peripheral rod oxidation, grid growth, grid
oxidation, guide thimble inner diameter oxidation, grid cell size,
crud scraping, single rod exams for the high burnup rods,
profilometry, and pellet-to-pellet gap measurements using a Gamma
Scanner instrument. All parameters inspected were found to be
acceptable.
By performing the above inspection regimen, the demonstrated
adherence to the inspection standards and acceptance criteria
precludes the potential for new risks to components and systems that
could introduce a new type of accident.
Based on this evaluation, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
There is no significant reduction in the margin of safety due to
the proposed change. The current TS Safety Limit (SL) 2.1.1.3 states
that ``In MODES 1 and 2, the peak fuel centerline temperature shall
be maintained 4700 deg.F.'' The TS Safety Limit Bases
states that overheating of the fuel is prevented by maintaining the
steady state peak Linear Heat Rate (LHR) below the level at which
fuel centerline melting occurs. Fuel centerline melting occurs when
the local LHR, or power peaking, in a region of fuel is high enough
to cause the fuel centerline temperature to reach the fuel melting
point.
WCAP-14483-A conservatively states that the fuel centerline
temperature limit has been established based on the melting
temperature for Uranium Dioxide (UO2) fuel of 5080
deg.F, decreasing by 58 deg.F per 10,000 MWD/MTU of burnup. Based
on the WCAP-14483-A equation, a burnup of approximately 65,500 MWD/
MTU could be accrued before the melting temperature would
academically reach the current TS SL of 4700 deg.F.
Westinghouse has evaluated the fuel centerline temperatures for
the Byron Station and Braidwood Station reactor cores under uprated
power conditions. This evaluation shows that the high burnup rods'
temperatures would remain below both the current SL of 4700 deg.F
and the proposed WCAP-14483-A equation (i.e., the proposed SL) for
fuel melting temperatures under extended burnup conditions past
75,000 MWD/MTU. Thus, fuel melting will not occur in the LTA high
burnup rods.
The insertion of the four high burnup rods does not impact any
other TS. The LTA has been designed to operate within the SAFDLs and
will therefore have sufficient safety margins. Furthermore, the high
burnup LTA will satisfy the five guidelines specified in WCAP-12488-
A approved by the NRC. The high burnup LTA will comply with TS 4.2.1
by being placed in a nonlimiting core region.
Based on the above discussion, changing the fuel centerline melt
temperature from the
[[Page 59506]]
existing 4700 deg.F to an equation consistent with the design basis
for fuel melt temperature will not significantly reduce the margin
of safety. The analysis shown in WCAP-12610-P-A indicates that the
minimum margin to safety occurs at fuel assembly Beginning of Life
(BOL). The evaluation in WCAP-12610-P-A demonstrates that margin of
safety with respect to the proposed SL equation remains sufficient
for fuel burnups up to 75,000 MWD/MTU.
Based on this evaluation, the proposed TS changes do not involve
a significant reduction in a margin of safety.
Conclusion: Based upon the above analyses and evaluations, we
have concluded that the proposed change to the TS involves no
significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Vice President,
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way,
Kennett Square, PA 19348.
NRC Section Chief: Anthony J. Mendiola.
Exelon Generation Company, LLC, Docket No. 50-352, Limerick Generating
Station, Unit 1, Montgomery County, Pennsylvania
Date of amendment request: September 14, 2001.
Description of amendment request: Exelon proposed to extend the use
of the pressure temperature limits specified in Technical Specification
(TS) Figure 3.4.6.1-1, ``Minimum Reactor Vessel Metal Temperature vs.
Reactor Vessel Pressure,'' through Cycle 10 of operation, currently
scheduled to end April 2004. Exelon also proposed to modify TS Table
4.4.6.1.3-1, ``Reactor Vessel Material Surveillance Program--Withdrawal
Schedule,'' with a note clarifying that surveillance capsule
withdrawals are to be scheduled for the nearest vessel refueling outage
date subsequent to the withdrawal time specified in the TS Table.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensees analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Extended Use of Pressure-Temperature Limits
The proposed change to the TSs to extend the use of the P-T limits
does not affect the operation or configuration of any plant equipment.
Thus, no new accident initiators are created by this change. The
proposed change extends the use of the pressure-temperature (P-T)
limits for an additional cycle of operation. The P-T curves prohibit
operational conditions in which brittle fracture of the reactor vessel
materials is possible. The P-T limits are based on the projected
reactor vessel neutron fluence at 32 effective full power years (EFPY)
of operation. At the end of the next cycle of operation, Cycle 10,
Limerick Generating Station (LGS) Unit 1 will have attained a maximum
of 48.1 percent of the 32 EFPY operating time which provides
significant margin to ensure that the current 32 EFPY fluence
projection will not be exceeded. This ensures that the basis for
proposed applicability of the P-T limits is conservative and that the
reactor vessel integrity is protected under all operating conditions.
Therefore, neither the probability nor the consequences of an accident
are increased.
Deferral of Withdrawal of Vessel Surveillance Specimens
Deferring the withdrawal of the vessel surveillance capsules will
not initiate or is not a precursor to any of the accident scenarios
presented in the Updated Final Safety Analysis Report (UFSAR). This
schedular adjustment will not increase the likelihood of equipment
failure, will not defeat the design reactor protection functions, and
will not increase the likelihood of failure of any plant structure,
system or component. Therefore, neither the probability nor the
consequences of an accident are increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Extended Use of Pressure-Temperature Limits
The proposed change to the technical specifications to extend the
use of the P-T limits does not affect the operation or configuration of
any plant equipment. The current P-T limits will remain valid and
conservative during the proposed extension. Thus, no new or different
accidents are created by this proposed change.
Deferral of Withdrawal of Vessel Surveillance Specimens
The proposed deferral of the withdrawal of the vessel surveillance
capsule does not involve a change to the plant design or operation. No
new equipment will be installed or utilized, and no new operating
conditions will be initiated as a result of this change. Because the P-
T limit curves are not impacted, the safety function of the reactor
vessel to mitigate the release of radioactive steam and limit reactor
inventory loss under normal, accident, and transient conditions is not
affected. Therefore, the proposed changes do not create the possibility
of a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Extended Use of Pressure-Temperature Limits
The proposed change extends the use of the current P-T limits for
an additional cycle of operation. No changes to the P-T limits are
proposed. The current P-T limit curves are based on the projected
reactor vessel neutron fluence after 32 EFPY of operation. At the end
of the next operating cycle, Cycle 10, LGS Unit 1 will have attained a
maximum of 48.1 percent of the 32 EFPY reactor vessel neutron fluence
projection upon which the current P-T curves are based. The maximum
operating time at the end of Cycle 10, when compared with the maximum
operating time assumed for the P-T limits curves, ensures that the P-T
limits will remain conservative and will ensure that the current
margins for reactor pressure vessel integrity are unchanged. The
proposed change maintains the relative margin of safety commensurate
with that which existed at the time the American Society of Mechanical
Engineers Boiler & Pressure Vessel Code, Section XI, Appendix G, was
approved in 1974. No plant safety limits, setpoints, or design
parameters are adversely affected by the proposed TS change. Therefore,
the proposed change does not involve a significant reduction in a
margin of safety.
Deferral of Withdrawal of Vessel Surveillance Specimens
No plant safety limits, set points, or design parameters are
adversely affected by the proposed deferral of withdrawal of vessel
surveillance specimens. The deferral of the withdrawal of the vessel
surveillance specimens does not affect the current P-T limit curves,
and therefore, does not affect a margin of safety.
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 59507]]
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Edward Cullen, Vice President & General
Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett
Square, PA 19348.
NRC Section Chief: James W. Clifford.
Florida Power and Light Company, et al. (FPL), Docket Nos. 50-335 and
50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: October 18, 2001.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) for St. Lucie, Units 1 and 2,
regarding Engineered Safety Feature Actuation System (ESFAS)
instrumentation. Specifically, it would limit the period of time that
inoperable recirculation actuation signal (RAS), containment spray
actuation signal (CSAS), and auxiliary feedwater actuation signal
(AFAS) input channels could be in the bypassed and/or tripped
condition. Generally, the proposed TS employ a 48-hour completion time
to restore an inoperable channel, which, in most cases, is more
restrictive than the existing TS, and is comparable to the value used
in the Standard TS for Combustion Engineering plants.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Would operation of the facility in accordance with the
proposed amendments involve a significant increase in the
probability or consequences of an accident previously evaluated?
No, facility operation under the new Technical Specification
(TS) restrictions would not increase the probability of occurrence
of any accident previously evaluated. The proposed changes only
affect the ESFAS functions of RAS, CSAS, and AFAS; generally
limiting the time that any instrument channel may be inoperable in a
bypassed or tripped condition. No physical plant changes are
proposed in conjunction with these revisions. The proposed changes
to RAS and AFAS channel operability greatly reduce the time that
actuation systems are vulnerable to spurious, inadvertent actuation.
The proposed changes do allow a new unlimited time for trip of one
CSAS channel on Unit 1. Although this increases the possibility of a
spurious channel trip with a potential for causing an inadvertent
spray actuation, this is offset by the increased reliability of
spray in this configuration. Unit 2 already contains provision for
the indefinite single channel trip of CSAS, and this change will
also make the two units similar. Additionally, it is important to
note that inadvertent actuation of any of these functions (RAS,
CSAS, or AFAS) during plant operation is not an accident initiating
event. Therefore, with no physical effects on the plant and no
increase in probability that the subject ESFAS functions will
initiate an accident, there is no increased probability that any
previously evaluated accident will occur. The changes provided in
this safety evaluation do not affect the assumptions or results of
any accident evaluated in the UFSAR [Updated Final Safety Analysis
Report].
Likewise, the consequences of any accident previously evaluated
have not been increased. The proposed changes, by limiting the time
that ESFAS functions are inoperable, will increase the reliability
of the associated ESFAS functions to respond to accidents. In
particular, the revision to the RAS TS will limit the time that the
RAS will be vulnerable to single failure and will therefore improve
the system reliability during an accident. As these proposed changes
constitute no physical change to the facility and only serve to
increase ESF function reliability, FPL concludes that the
consequences of previously evaluated accidents are not increased.
The ability of the ESFAS to respond to accident conditions as
assumed in any accident analysis has not been affected.
(2) Would operation of the facility in accordance with the
proposed amendments create the possibility of a new or different
kind of accident from any accident previously evaluated?
No, the proposed activity does not create the possibility of an
accident of a different type than any previously evaluated. The
proposed changes only affect the ESFAS functions of RAS, CSAS, and
AFAS; generally limiting the time that any instrument channel may be
inoperable in a bypassed or tripped condition. No physical plant
changes are proposed in conjunction with these revisions. Thereby,
the proposed changes do not create any new equipment interfaces,
equipment response characteristics, or operating configurations.
Without creation of a new interaction of materials, operating
configuration, or operating interface, there is no possibility that
the proposed changes can introduce a new or different kind of
accident.
(3) Would operation of the facility in accordance with the
proposed amendments involve a significant reduction in a margin of
safety?
The margin of safety as defined in the basis for any Technical
Specification or in any licensing document has not been reduced. The
TS Bases for the associated ESF LCO [Limiting Condition for
Operation] do not explicitly discuss a related margin of safety.
However, by virtue of the increased ESFAS reliability provided by
the proposed amendments, it is evident that the margin of safety
will not be reduced in any manner.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Richard P. Correia.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: October 17, 2001.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 6.8.4.h, ``Containment Leakage Rate
Testing Program,'' to allow only one-time deviation from the 10-year
frequency of the performance-based leakage rate testing program for
Type A tests as recommended by Nuclear Energy Institute, NEI 94-01,
Revision 0, ``Industry Guideline for Implementing Performance-Based
Option of 10 CFR part 50, Appendix J,'' and endorsed by Regulatory
Guide 1.163, ``Performance-Based Containment Leak-Rate Program.'' The
one-time deviation would allow intergrated leak rate testing (ILRT) at
no more than 15 years after the last ILRTs, performed in November 1992
and October 1991 for Units 3 and 4 respectively.
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) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendment to the Technical Specifications adds a
one-time extension to the current interval for Type A (ILRT)
testing. The current test interval of ten years, based on past
performance, would be extended on a one-time basis to 15 years from
the last Type A test. The proposed extension to Type A testing
cannot increase the probability of an accident previously evaluated
since the containment Type A testing extension is not a
modification, nor a change to the operation of the plant, and the
test extension is not a type that could lead to equipment failure or
accident initiation. The proposed extension of Type A testing does
not involve a significant increase in the consequences of an
accident since research documented in NUREG-1493 has found that,
generically, very few potential containment leakage paths are not
identified with Type B and C tests. In fact, an analysis of 144 ILRT
results, including 23 failures, found that no failures
[[Page 59508]]
were due to containment liner breech. The NUREG concluded that
reducing the Type A frequency to one per twenty years was found to
lead to an imperceptible increase in risk.
Florida Power & Light provides a high degree of assurance
through testing and inspection that the containment will not degrade
in a manner detectable only by Type A testing. The last four Type A
tests for both Turkey Point Units 3 and 4 show leakage rates well
below acceptance criteria, indicating a leak-tight containment.
Inspections required by the Maintenance Rule [10 CFR 50.65] and ASME
[American Society of Mechanical Engineers] code, will identify
indications of containment structure degradation that could affect
that leak tightness. Type B and C testing required by Technical
Specifications will identify any containment openings, such as
valves, that would otherwise be detected by the Type A tests. These
factors show that the Turkey Point Units 3 and 4 Type A test
extension will not represent a significant increase in the
consequences of an accident.
Based on the above, it is concluded that the proposed amendments
to extend the Type A test frequency does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any previously evaluated.
The proposed change does not create a new or different type of
accident for Turkey Point because no physical plant changes are
being made, and no compensatory measures are imposed that would
create a new failure scenario. The proposed change only requests a
one-time extension to the current interval for Type A testing. The
current test interval of 10 years, based on past performance, would
be extended on a one-time basis to 15 years from the last Type A
test.
The proposed extension to Type A testing cannot create the
possibility of a new or different type of accident because there are
no physical changes being made to the plant, and there are no
changes to the operation of the plant that could introduce a new
failure mode creating an accident or affecting the mitigation of an
accident.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed license amendment requests a one-time extension to
the current interval for Type A testing. The current test interval
of ten years, based on past performance, would be extended on a one-
time basis to 15 years from the last Type A test. The proposed
extension to Type A testing will not significantly reduce the margin
of safety. The NUREG-1493 generic study of the effects of extending
containment leakage testing found that a 20-year test interval for
Type A leakage testing resulted in an imperceptible increase in risk
to the public. NUREG-1493 found that, generically, the design
containment leakage rate contributed about 0.1 percent to the
individual risk and that the decrease in Type A testing frequency
would have minimal effect on this risk, since 95 percent of the
potential leakage paths are detected by Type B and C testing. A
Turkey Point plant-specific risk calculation is consistent with the
generic conclusions identified in NUREG-1493.
Therefore, the proposed changes in this license amendment will
not result in a significant reduction in the plant's margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Richard P. Correia.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: August 7, 2001.
Description of amendment requests: The proposed amendments would
create Technical Specification (TS) 3.0.6 and associated bases to allow
equipment that was removed from service or declared inoperable to be
returned to service under administrative controls solely to perform the
testing required to demonstrate its operability or the operability of
other equipment. TS 3.0.6 would incorporate the administrative controls
currently approved for use as TS 3.0.5 in NUREG-1431, ``Standard
Technical Specifications Westinghouse Plants,'' Revision 2, dated April
30, 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. Does the change involve a significant increase in the probability of
occurrence or consequences of an accident previously evaluated?
Probability of Occurrence of an Accident Previously Evaluated
The potential impact of temporarily returning the equipment to
service is considered to be insignificant since the equipment will
either be expected to be able to perform its required safety
function or sufficient redundancy will exist such that the function
would still occur if required. This is addressed in Generic Letter
(GL) 87-09, ``Sections 3.0 and 4.0 of the Standard Technical
Specifications (STS) on the Applicability of Limiting Conditions for
Operation and Surveillance Requirements.'' GL 87-09 states, ``It is
overly conservative to assume that systems or components are
inoperable when a surveillance has not been performed because the
vast majority of surveillances do in fact demonstrate that systems
or components are operable.'' In addition, returning the equipment
to service for testing will promote timely restoration of the
equipment. Therefore, the proposed changes do not significantly
affect accident initiators or precursors.
The proposed change to create a Bases statement for TS 3.0.6
provides explanatory information regarding the intent of the
specification and how it is to be implemented. The proposed Bases
change does not alter requirements of the associated TS. Therefore,
the effect of the Bases change on accident initiators and precursors
of an accident is bounded by the effect of the TS change as
described above. The format changes are intended to improve
appearance and do not alter any requirements.
Therefore, the proposed changes do not adversely affect any
accident initiators or precursors and will not involve a significant
increase in the probability of an accident previously evaluated.
Consequences of an Accident Previously Evaluated
The proposed change will allow temporarily returning equipment,
that was previously declared inoperable, to service in a state in
which it is expected to function to mitigate the consequences of a
previously analyzed accident. The proposed change will also permit
temporarily restoring inoperable equipment to service in situations
where sufficient redundancy would exist for its function to mitigate
the consequences of a previously analyzed accident to be performed.
This will promote timely restoration of equipment and capabilities
to mitigate the consequences of an accident previously analyzed.
The proposed change to include a Bases statement for TS 3.0.6
provides explanatory information regarding the intent of the
specification and how it is to be implemented. The proposed Bases
change does not alter requirements of the associated TS. Therefore,
the effect of the Bases change on offsite dose consequences of an
accident previously analyzed is bounded by the effect of the TS
change as described above. The format changes are intended to
improve appearance and do not alter any requirements.
Therefore, the probability of occurrence or the consequences of
accidents previously evaluated are not significantly increased.
2. Does the change create the possibility of a new or different kind of
accident from any accident previously evaluated?
The proposed changes do not introduce a new mode of plant
operation and do not involve a physical modification to the plant.
Operation with the inoperable equipment temporarily restored to
service under administrative controls is not considered a new mode
of operation since the equipment is not being physically altered. As
such, the manner in which it can fail remains the same.
[[Page 59509]]
The proposed change to include a Bases statement for TS 3.0.6
provides explanatory information regarding the intent of the
specification and how it is to be implemented. The proposed Bases
change does not alter requirements of the associated TS. Therefore,
the effect of the Bases changes on accident initiators or precursors
is bounded by the effect of the associated TS as described above.
The format changes are intended to improve appearance and do not
alter any requirements.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin of
safety?
The proposed new TS 3.0.6 can be applied to any structures,
systems, and components that are governed by the TS. As such, the
proposed changes are applicable to every margin of safety imposed by
the TS.
The proposed change will allow temporarily returning equipment
that was previously declared inoperable to service in a state in
which it is expected to function to mitigate the consequences of a
previously analyzed accident. The proposed change will also permit
temporarily restoring inoperable equipment to service in situations
where sufficient redundancy would exist for its function to mitigate
the consequences of a previously analyzed accident to be performed.
The performance of the testing should confirm the expected
capability of the equipment and there is no significant impact on
any TS safety setting or setpoint.
There is no margin of safety pertinent to the proposed Bases
change. The format changes are intended to improve appearance and do
not alter any requirements.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety. In summary, based upon the above
evaluation, I&M has concluded that the proposed amendment involves
no significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: William D. Reckley, Acting Section Chief.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: May 9, 2001.
Description of amendment request: The proposed amendment would
change the Technical Specification (TS) to correct an error in TS Table
3.3.1.1-1 Function 2.b, correct a typographical error in labeling
surveillance requirement 3.3.1.1.13, and revise bases pages B 3.3-8 and
B 3.3-10.
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?
This change is to correct an error in documentation that was
introduced during implementation of Amendment 151 and retained in TS
during the conversion to ITS [Improved Technical Specifications] as
well as an error that was introduced into TS during the conversion
to ITS. Neither the design basis nor the functionality of the
instrumentation is being physically changed. The Neutron Monitoring
system performs a mitigating function and is not an accident
initiating system. The actual mitigating function of the Neutron
Monitoring is not changed. Only an implied but non-existent
mitigating capability is being removed from TS. This change does not
create or modify any accident initiators. Therefore, there is no
increase in the probability of an accident previously evaluated.
The APRM [Average Power Range Monitor] system is credited for
mitigating the consequences of the Control Rod Drop Accident. The
APRM system also provides protection for the reactor to mitigate the
consequences of such abnormal operational transients as loss of
feedwater heater, pressure regulator failure, or Main Steam
Isolation Valve closure. The proposed change will not change the
functionality or setpoints for either the APRM Flux-High (Fixed) or
the APRM Flux-High (Biased) functions. Additionally, the correction
of an incorrect Surveillance Requirement reference does not change
how any surveillance is performed. Therefore the consequences of an
accident previously evaluated will not be increased.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Since this change in the TS does not involve a physical change
to the instrumentation, to the setpoints, or to the design or
functionality of the circuitry for reactor scram on APRM Flux-High,
fixed or flow-biased, the change does not create a possibility of a
new or different kind of accident not previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The setpoints for the Neutron Flux-High instrumentation are not
changed by this proposed TS change. The safety function allowable
value setpoint remains at less than or equal to 120% RTP [rated
thermal power]. The formula for the APRM Flux-High (flow biased) is
not being changed. Since neither of these is being changed, the
margin of safety is not reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: August 6, 2001, as supplemented November
2, 2001.
Description of amendment request: The proposed amendment would
revise the Seabrook Station Technical Specifications (TS) Index, TS
1.0, ``Definitions,'' and TS Table 1.2, ``Operational Modes,'' to
reflect the improved Standard Technical Specifications for Westinghouse
plants.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes to TS Index, TS 1.0 and TS Table 1.2 are
changes that do not change any structures, systems or components
(SSCs) thus, the proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, and configuration of the facility. In addition, the
proposed changes do not affect the manner in which the plant
responds in normal operation, transient or accident conditions. The
proposed changes do not alter or prevent the ability of SSCs to
perform their intended function to mitigate the consequences of an
initiating event within the acceptance limits assumed in the Updated
Final Safety Analysis Report (UFSAR). Finally, while these changes
may afford North Atlantic operational flexibility, the changes are
an enhancement and do not affect plant safety.
The proposed changes do not affect the source term, containment
isolation or radiological release assumptions used in evaluating the
radiological consequences of an accident previously evaluated in the
Seabrook Station UFSAR. 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, it is concluded that these proposed revisions to TS
Index, TS 1.0 and TS Table 1.2 do not involve a significant increase
in the probability or consequence of an accident previously
evaluated.
[[Page 59510]]
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
This [sic] proposed changes to TS Index, TS 1.0 and TS Table 1.2
are changes that do not change the operation or the design basis of
any plant system or component during normal or accident conditions.
The proposed change incorporates definitions delineated in the
improved Standard Technical Specifications (NUREG-1431). The
proposed changes do not include any physical changes to the plant.
In addition, the proposed changes do not change the function or
operation of plant equipment or introduce any new failure
mechanisms. The plant equipment will continue to respond per the
design and analyses and there will not be a malfunction of a new or
different type introduced by the proposed changes.
The proposed changes are administrative in nature and only
correct, update and clarify the Seabrook Station Operating License
to reflect the definitions in the improved Standard Technical
Specifications. The proposed changes do not modify the facility nor
do they affect the plant's response to normal, transient or accident
conditions. The changes do not introduce a new mode of plant
operation. While these changes may afford North Atlantic operational
flexibility, the changes are an enhancement and do not affect plant
safety. The plant's design and design basis are not revised and the
current safety analyses remains in effect.
Thus, these proposed revisions to TS Index, TS 1.0 and TS Table
1.2 do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The proposed changes to TS Index, TS 1.0 and TS Table 1.2 are
administrative in nature and only correct, update and clarify the
Seabrook Station Operating License to reflect the improved Standard
Technical Specifications. While these changes may afford North
Atlantic operational flexibility, the changes are an enhancement and
do not affect plant safety. The safety margins established through
Limiting Conditions for Operation, Limiting Safety System Settings
and Safety Limits as specified in the Technical Specifications are
not revised nor is the plant design revised by the proposed changes.
Thus, it is concluded that these proposed revisions to TS Index,
TS 1.0 and TS Table 1.2 do not involve a significant reduction in a
margin of safety.
Based on the above evaluation, North Atlantic concludes that the
proposed changes to TS Index, TS 1.0 and TS Table 1.2 do not
constitute a significant hazard.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Section Chief: James W. Clifford.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: November 2, 2001.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Table 3.3.1-1, Item 1, ``Variable
High Power Trip'' (VHPT), by increasing the maximum allowable value for
the VHPT from 106.5 percent to 111 percent.
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:
Nuclear Management Company has evaluated whether or not a
significant hazards consideration is involved with the proposed
amendment by focusing on the three standards set forth in 10 CFR
50.92, ``Issuance of Amendment.'' The following evaluation supports
the finding that operation of the facility in accordance with the
proposed change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to the maximum Allowable Value for the
Variable High Power Trip (VHPT) function in the Technical
Specifications would not change or remove any considerations of
uncertainties from the FSAR [Final Safety Analysis Report] Chapter
14 Safety Analysis. The methodology that was utilized in determining
the recommended change in the maximum allowable value follows
standard ANSI/ISA-S67.04-1994, ``Setpoints for Nuclear Safety-
Related Instrumentation,'' and NRC Regulatory Guide 1.105,
``Setpoints for Safety-Related Instrumentation,'' Revision 3. With
the proposed changes to the maximum allowable value and calculated
setpoint of the VHPT in place, the reactor is still protected from
reaching the analytical limit of 115% reactor power.
Therefore, operation of the facility in accordance with the
proposed change to the Technical Specifications would not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed changes to the maximum Allowable Value and
Calculated Setpoint for the Variable High Power Trip function in the
Technical Specifications would not change or add a system function.
The proposed change alters the way the uncertainties (including
uncertainties of instrument measurement and calibration) are
accounted for without actually removing uncertainties from the
calculation. This proposed change follows the standard ANSI/ISA-
S67.04-1994 and NRC Regulatory Guide 1.105, Revision 3.
Thus, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change to the maximum Allowable Value for the
Variable High Power Trip function in the Technical Specifications
would account for all uncertainties in the VHP trip setpoint
calculation, instead of taking them into account in the maximum
allowable value calculation, as is currently done. In addition,
double accounting for nuclear instrumentation uncertainties has been
removed. The uncertainties will still be taken into account in
determining the calculated setpoint based on the maximum allowable
value of the VHPT, in accordance with the standard ANSI/ISA-S67.04-
1994 and NRC Regulatory Guide 1.105, Revision 3. This methodology
continues to assure that the Analytical Limit will not be exceeded.
Therefore, the proposed change to the Technical Specifications
would not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based upon
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: William D. Reckley (Acting).
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: November 1, 2001.
Description of amendment request: The proposed amendments would
change the Technical Specifications (TSs) to allow a one-time extension
of the allowed outage time for the control room emergency filtration
system (CREFS) from 7 days to 30 days. The licensee is requesting this
one-time change in order to implement modifications to the CREFS.
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. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant increase in
the probability or consequences of any accident previously
evaluated.
[[Page 59511]]
The operability of CREFS ensures that the control room will
remain habitable for operators during and following all credible
accident conditions. The inoperability or failure of CREFS is not an
accident initiator or precursor. Therefore, the probability of an
accident previously evaluated will not be significantly increased as
a result of the proposed change. Because design limitations continue
to be met and the integrity of the reactor coolant system pressure
boundary is not challenged, the assumptions employed in the
calculation of the offsite radiological doses remain valid. Control
room dose calculations are not affected outside the limited one-time
period when the CREFS modifications/upgrades are ongoing.
During the period that CREFS will be inoperable, temporary
ventilation will provide adequate filtration to the control room and
adequate cooling to the control and computer rooms. The
effectiveness of the temporary filtration provided during this 30
day period is not significantly less than that of the permanently
installed CREFS. Only the duration of a currently allowed outage
time is being changed, with commensurate compensatory measures being
taken. Therefore, the consequences of an accident previously
evaluated will not be significantly increased as a result of the
proposed change.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a new or different kind
of accident from any accident previously evaluated.
The possibility for a new or different type of accident from any
accident previously evaluated is not created as a result of this
amendment. The evaluation of the effects of the proposed changes
indicate that all design standards and applicable safety criteria
limits are met. These changes therefore do not cause the initiation
of any new or different accident nor create any new failure
mechanisms.
Equipment important to safety will continue to operate as
designed. Only the duration of a system's allowed outage time is
being changed. Component integrity is not challenged. The changes do
not result in any event previously deemed incredible being made
credible. The changes do not result in more adverse conditions or
result in any increase in the challenges to safety systems.
Therefore, operation of the Point Beach Nuclear Plant in accordance
with the proposed amendments will not create the possibility of a
new or different type of accident from any accident previously
evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant reduction
in a margin of safety.
The CREFS functions to mitigate the effects of accidents.
Implementation of the modifications/upgrades will require removing
the system from service for a period of time longer than presently
allowed by the Technical Specification. This results in a longer
period during which the consequences of a design basis accident,
affecting the dose of control room personnel, may be slightly
increased. During the period that CREFS will be inoperable, a
temporary ventilation system will provide adequate filtration to the
control room and adequate cooling to the control and computer rooms.
The effectiveness of the temporary filtration provided during this
30 day period is not significantly less than that of the permanently
installed CREFS. Only the duration of a currently allowed outage
time is being changed, with commensurate compensatory measures being
taken. There are no new or significant changes to the initial
conditions contributing to accident severity or consequences. The
proposed modification will not otherwise affect the plant protective
boundaries, will not cause a release of fission products to the
public, nor will it degrade the performance of any other SSCs
important to safety. The analysis for the limiting design basis
accident, the large break LOCA, has a significant amount of
conservatism built in to account for uncertainties in system
performance an analysis techniques. This conservative margin of
safety, along with the temporary filtration unit, provide a high
level of confidence that the health and safety of the operators will
be maintained, such that they will be able to prevent or mitigate an
event. Therefore, removing the CREFS from service for 30 days on a
one-time basis to permit system upgrading, will not significantly
reduce the margin of safety. The improvements to CREFS resulting
from the proposed modifications will enhance operator protection
against conditions resulting from a design basis accident and
therefore provide a net benefit to radiological health and reactor
safety.
Conclusion
Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not result in a significant increase in
the probability or consequences of any accident previously analyzed;
will not result in a new or different kind of accident from any
accident previously analyzed; and, does not result in a significant
reduction in any margin of safety. Therefore, operation of PBNP
[Point Beach Nuclear Plant] in accordance with the proposed
amendments does not result in a significant hazards determination.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: William Reckley (Acting).
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station (SSES), Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: October 18, 2001.
Description of amendment request: The proposed amendments would
modify the Technical Specification Surveillance Requirement (SR)
3.4.3.1 for testing of the main steam safety relief valves (MSRVs) so
that the setpoint tolerance for ``As-Found'' testing would be changed
from 1 percent to 3 percent. The requirements
for testing of the tolerances associated with ``As-left'' testing would
remain unchanged. An editorial change would also be made to remove a
note regarding an associated relief request.
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 action does not involve a significant increase
in the probability or consequences of an accident as previously
evaluated.
The proposed change allows an increase in the as-found MSRV
safety mode setpoint tolerance, determined by test after the valves
have been removed from service, from 1% to
3%. The proposed change does not alter the TS 3.4.3
Surveillance Requirements on the nominal MSRV safety mode lift
setpoints, the MSRV relief mode setpoints, the required frequency
for the MSRV lift setpoint tests, or the number of MSRVs required to
be operable.
Consistent with current requirements, this change continues to
require that these valves be adjusted to within 1% of
their nominal lift setpoints following testing. The proposed action
does not change any other behavior or operation of any MSRV, and
therefore, has no significant impact on the reactor operation. It
also has no significant impact on response to any perturbation of
reactor operation including transients and accidents previously
analyzed in the Final Safety Analysis Report (FSAR).
The proposed action does not involve physical changes to the
valves, nor does it change the safety function of the valves. The
proposed TS revision involves no significant changes to the
operation of any systems or components in normal or accident
operating conditions and no changes to existing structures, systems,
or components. Therefore, these changes will not increase the
probability of an accident previously evaluated.
Generic considerations related to the change in setpoint
tolerance were addressed in NEDC-31753P, ``BWROG In-Service Pressure
Relief Technical Specification Revision Licensing Topical Report,''
and were reviewed and approved by the NRC in a Safety Evaluation
Report (SER) dated March 8, 1993. The plant specific evaluations,
required by the NRC's SER and performed to support this proposed
change, show that there is adequate margin to the design core
thermal limits and to the reactor vessel pressure limits using a
3% setpoint tolerance. These analyses also show that
[[Page 59512]]
operation of the high pressure coolant injection (HPCI) and reactor
core isolation cooling (RCIC) systems are not adversely affected and
the containment response from a loss of coolant accident is
acceptable. The plant systems associated with these proposed changes
are capable of meeting all applicable design basis requirements and
retain the capability to mitigate the consequences of accidents
described in the FSAR. Therefore, these changes do not involve an
increase in the consequences of any accident previously evaluated.
Therefore, the proposed amendment does not increase the
probability or consequences of an accident previously evaluated.
2. The proposed action does not create a possibility of a new or
different kind of accident than previously evaluated.
The proposed change was developed in accordance with the
provisions contained in the NRC SER, dated March 8, 1993, for the
``BWR Owners Group Inservice Pressure Relief Technical Specification
Revision Licensing Topical Report,'' NEDC-31753P. The revised MSRV
setpoint tolerance limit does not adversely impact the operation of
any safety-related component or equipment. Since the proposed action
does not involve hardware changes, significant changes to the
operation of any systems or components, nor changes to existing
structures, systems, or components, there is no possibility that a
new or different kind of accident is created.
The proposed change to allow an increase in the MSRV safety mode
setpoint tolerance from 1% to 3% does not
alter the nominal MSRV lift setpoints or the number of MSRVs
currently required to be operable by SSES Technical Specifications.
The proposed action does not involve physical changes to the valves,
nor does it change the safety function of the valves. The proposed
action does not involve a physical alteration of any existing plant
equipment. No new or different equipment is being installed. There
is no alteration to the parameters within which the plant is
normally operated. As a result no new failure modes are being
introduced. There are no changes in the procedures governing normal
plant operation, nor the procedures utilized to respond to plant
transients.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed action does not involve a significant reduction
in a margin of safety.
The proposed action does not involve a significant reduction in
a margin of safety. Establishment of the 3% MSRV safety
setpoint tolerance limit does not adversely impact the operation of
any safety-related component or equipment. Engineering evaluations
concluded that there are no significant impacts on fuel thermal
limits, safety related systems, structures or components, and no
significant impact on the accident analyses associated with the
proposed changes.
The margin of safety is established through the design of the
plant structures, systems, and components, the parameters within
which the plant is operated, and the establishment of the setpoints
for the actuation of equipment relied upon to respond to an event.
The proposed change does not significantly impact the condition or
performance of structures, systems, and components relied upon for
accident mitigation.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: Lakshminaras Raghaven.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: September 20, 2001.
Description of amendment request: The proposed amendments would
support extension of the operating cycle from 18 months to 24 months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
a. Surveillance Testing Interval Extensions.
The proposed Technical Specification (TS) change involves a
change in the surveillance testing intervals to facilitate a change
in the operating cycle from 18 months to 24 months. The proposed TS
change does not physically impact the plant, nor does it impact any
design or functional requirements of the associated systems. That
is, the proposed TS change neither degrades the performance of, nor
increases the challenges to, any safety systems assumed to function
in the plant safety analysis. The proposed TS change neither impacts
the TS SRs [surveillance requirements] themselves nor the manner in
which the surveillances are performed.
In addition, the proposed TS change does not introduce any
accident initiators, since no accidents previously evaluated relate
to the frequency of surveillance testing. Also, evaluation of the
proposed TS change demonstrates that the availability of equipment
and systems required to prevent or mitigate the radiological
consequences of an accident is not significantly affected because of
other, more frequent testing that is performed, the availability of
redundant systems and equipment, or the high reliability of the
equipment. Since the impact on the systems is minimal, it is
concluded the overall impact on the safety analysis is negligible.
Furthermore, an historical review of surveillance test results
and associated maintenance records indicate there is no evidence of
any failure that would invalidate the above conclusions. Therefore,
the proposed TS change does not significantly increase the
probability or consequences of an accident previously evaluated.
b. Allowable Value Changes.
A change in Allowable Values is proposed for Table 3.3.5.1-1,
Item 2.f. The proposed change is the result of application for the
Hatch Instrument Setpoint Methodology using plant-specific drift
values. Application of this methodology results in Allowable Values
that more accurately reflect total instrumentation loop accuracy, as
well as that of test equipment and calculated drift between
surveillances. The proposed change will not result in any hardware
changes. The instrumentation is not assumed to be an initiator of
any analyzed event. Existing operating margin between plant
conditions and actual plant setpoints is not significantly reduced
due to the proposed changes. The role of the instrumentation is in
mitigating and thereby, limiting the consequences of accidents.
The Allowable Values were developed to ensure the design and
safety analysis limits are satisfied. The methodology used for the
development of the Allowable Values ensures: 1) the affected
instrumentation remains capable of mitigating design basis events as
described in the safety analysis and 2) the results and radiological
consequences described in the safety analysis remain bounding.
Additionally, the proposed change does not alter the plant's ability
to detect and mitigate events. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
c. Surveillance Testing Interval Reduction to Semiannual.
The proposed TS change involves a reduction in the surveillance
testing interval from 18 months to 184 days for the instrumentation
associated with Table 3.3.8.2-1. The shorter intervals are based
upon the plant-specific results of a review of the surveillance test
history for the devices. The implementing procedures for these SRs
have been performed on a 184-day interval for a number of years, and
this change more accurately reflects actual plant maintenance
practices. The proposed, more restrictive TS change does not
physically impact the plant, nor does it impact any design or
functional requirements of the associated systems. That is, the
proposed TS change neither degrades the performance of, nor
increases the challenges to, any safety systems assumed to function
in the safety analysis. This proposed TS change neither impacts the
TS SRs
[[Page 59513]]
themselves nor the manner in which the surveillances are performed.
In addition, the proposed TS change does not introduce any
accident initiators, since no accidents previously evaluated relate
to the frequency of surveillance testing. The proposed TS intervals
demonstrate that the equipment and systems required to prevent or
mitigate the radiological consequences of an accident are continuing
to meet the assumptions of the setpoint evaluation on a more
frequent basis. Since the impact on the systems is minimal, and the
assumptions of the safety analyses are maintained, it is concluded
the overall impact on the plant safety analysis is negligible.
Furthermore, setpoint drift evaluations prepared for the subject
instrumentation show that the existing Allowable Values are
acceptable without change. Therefore, the proposed TS change does
not significantly increase the probability or consequences of an
accident previously evaluated.
d. Change of CHANNEL CALIBRATION to CHANNEL FUNCTIONAL TEST for
Float Switches.
The proposed TS change involves a change in the SRs from CHANNEL
CALIBRATIONS to CHANNEL FUNCTIONAL TESTS for float switches used in
Table 3.3.1.1-1, Item 7.b; Table 3.3.5.1-1, Item 3.d; and Table
3.3.5.2-1, Items 3 and 4. The float switches are mechanical devices
that require mechanical setting at the proper level only. Because
the devices cannot be significantly adjusted without a physical
change in the location of the installation, the CHANNEL FUNCTIONAL
TEST provides all the functionality of a CHANNEL CALIBRATION for
this type of device. Therefore, the change in type of SR does not
impact the actual testing requirements for the subject devices.
The proposed TS change does not physically impact the plant, nor
does it impact any design or functional requirements of the
associated systems. That is, the proposed TS change neither degrades
the performance of, nor increases the challenges to, any safety
systems assumed to function in the safety analysis. The proposed TS
change does not impact the manner in which the surveillances are
performed.
In addition, the proposed TS change does not introduce any
accident initiators, since the same functional requirements exist
with the proposed change. Also, evaluation of the proposed TS change
demonstrates the availability of equipment and systems required to
prevent or mitigate the radiological consequences of an accident is
not significantly affected because of the availability of redundant
systems and equipment and the high reliability of the equipment.
Since the impact on the systems is minimal, it is concluded the
overall impact on the plant safety analysis is negligible.
Furthermore, an historical review of surveillance test results
and associated maintenance records indicated that there was no
evidence of any failures that would invalidate the above
conclusions. Therefore, the proposed TS change does not
significantly increase the probability or consequences of an
accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
a. Surveillance Testing Interval Extensions.
The proposed TS change involves a change in the surveillance
testing intervals to facilitate a change in the operating cycle
length. The proposed TS change does not introduce any failure
mechanisms of a different type than those previously evaluated,
since there are no physical changes being made to the facility. No
new or different equipment is being installed. No installed
equipment is being operated in a different manner. As a result, no
new failure modes are introduced. In addition, the SRs themselves,
and the manner in which surveillance tests are performed, remain
unchanged.
Furthermore, an historical review of surveillance test results
and associated maintenance records indicate there is no evidence of
any failure that would invalidate the above conclusions. Therefore,
the proposed TS change does not create the possibility of a new or
different kind of accident from any previously evaluated.
b. Allowable Value Changes.
The proposed change in Allowable Values is the result of
application of the Instrument Setpoint Methodology using plant-
specific drift values and does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. This is based upon the fact that the method and manner of
plant operation are unchanged.
The use of the proposed Allowable Values does not impact safe
operation of the plant in that the safety analysis limits are
maintained. The proposed change in Allowable Values involves no
system additions or physical modifications to plant systems. The
Allowable Values are revised to ensure the affected instrumentation
remains capable of mitigating accidents and transients. Plant
equipment will not be operated in a manner different from previous
operation, except that setpoints may be changed. Since operational
methods remain unchanged and the operating parameters were evaluated
to maintain the plant within existing design basis criteria, no
different type of failure or accident is created.
c. Surveillance Testing Interval Reductions to Semiannual.
The proposed TS change involves a change in the surveillance
testing interval due to the review of the surveillance test history
of the subject devices. Also, the semiannual tests reflect current
HNP calibration practices. The proposed TS change does not introduce
any failure mechanism of a different type than those previously
evaluated, since the proposed change makes no physical changes to
the plant. No new or different equipment is being installed. No
installed equipment is being operated in a different manner.
Furthermore, an historical review of surveillance test results
and associated maintenance records indicate there is no evidence of
any failure that would invalidate the above conclusions. Therefore,
the proposed TS change does not create the possibility of a new or
different kind of accident from any previously evaluated.
d. Change of CHANNEL CALIBRATION to CHANNEL FUNCTIONAL TEST for
Float Switches.
The proposed TS change does not impact the actual testing
requirements for the subject devices. The proposed TS change does
not introduce any failure mechanism of a different type than those
previously evaluated, since the proposed change makes no physical
changes to the plant. No new or different equipment is being
installed. No installed equipment is being operated in a different
manner. As a result, no new failure mode is being introduced. In
addition, the SRs themselves, and the manner in which surveillance
tests are performed, remain unchanged.
Furthermore, an historical review of surveillance test results
and associated maintenance records indicates there is no evidence of
any failure that would invalidate the above conclusions. Therefore,
the proposed TS change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The proposed amendment will not involve a significant
reduction in a margin of safety
a. Surveillance Testing Interval Extensions.
Although the proposed TS change results in changes in the
interval between surveillance tests, the impact, if any, on system
availability is minimal, based upon other, more frequent testing
that is performed, the existence of redundant systems and equipment,
or overall system reliability. Evaluations show there is no evidence
of any time-dependent failure that would impact the system
availability.
The proposed change does not significantly impact the condition
or performance of structures, systems, and components relied upon
for accident mitigation. The proposed change does not significantly
impact any safety analysis assumptions or results. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety.
b. Allowable Value Changes.
The proposed change does not involve a reduction in a margin of
safety. The proposed change was developed using a methodology to
ensure safety analysis limits are not exceeded. As such, this
proposed change does not involve a significant reduction in a margin
of safety.
c. Surveillance Testing Interval Reductions to Semiannual.
The proposed TS change results in a shorter interval between
surveillance tests to ensure the assumptions of the safety analysis
are maintained. The impact, if any, on system availability is
minimal, as a result of the more frequent testing that is performed.
The proposed change does not significantly impact the condition or
performance of structures, systems, and components relied upon for
accident mitigation. The proposed change does not significantly
impact any safety analysis assumptions or results. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety.
d. Change of CHANNEL CALIBRATION to CHANNEL FUNCTIONAL TEST for
Float Switches.
[[Page 59514]]
The proposed TS change does not impact the actual testing
requirements for the subject devices. The impact, if any, on system
availability due to this change is minimal, based upon the existence
of redundant systems and equipment and overall system reliability.
An historical review of surveillance test results and associated
maintenance records indicates there is no evidence of any failure
that would invalidate the above conclusions. The proposed change
does not significantly impact the condition or performance of
structures, systems, and components relied upon for accident
mitigation. The proposed change does not significantly impact any
safety analysis assumptions or results. 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: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: Richard J. Laufer, Acting.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: September 20, 2001.
Description of amendment request: The proposed amendments would
change specified surveillances from 92 days to 184 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed Technical Specifications (TS) change involves an
increase in the surveillance testing intervals for various
Surveillance Requirements (SRs) from 92 days to 184 days. The
proposed TS changes do not physically impact the plant, nor do they
impact any design or functional requirements of the associated
systems. That is, the proposed TS change does not degrade the
performance of, or increase the challenges to, any safety systems
assumed to function in the safety analysis. The proposed TS changes
neither impact the TS SRs themselves nor the way in which the
surveillances are performed. In addition, the proposed TS change
does not introduce any accident initiators, since no accidents
previously evaluated relate to the frequency of surveillance
testing. Also, evaluation of the proposed TS change demonstrates
that the availability of equipment and systems required to prevent
or mitigate the radiological consequences of an accident are not
significantly affected because of other, more frequent testing that
is performed, the availability of redundant systems and equipment,
or the high reliability of the equipment. Since the impact on the
systems is minimal, it is concluded that the overall impact on the
plant safety analysis is negligible.
A sensitivity analysis was performed to determine the effect of
the increased surveillance intervals on the HNP [Hatch Nuclear
Plant] Probabilistic Risk Assessment (PRA). This sensitivity
analysis shows a negligible increase in core damage frequency (CDF)
and essentially no change in large early release frequency (LERF)
due to the proposed change.
Furthermore, an historical review of surveillance test results
and associated maintenance record indicates there is no evidence of
any failure that would invalidate the above conclusions. Therefore,
the proposed TS change does not significantly increase the
probability or consequences of an accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS change involves a change in the various SR
intervals from 92 days to 184 days. The proposed TS change does not
introduce any failure mechanisms of a different type than those
previously evaluated, since no physical changes to the plant are
being made. Also, no new or different equipment is being installed,
and no installed equipment is being operated in a different manner.
As a result, no new failure modes are introduced. In addition, the
surveillance test requirements themselves, and the way surveillance
tests are performed, remain unchanged.
Furthermore, an historical review of surveillance test results
and associated maintenance records indicates there is no evidence of
any failure that would invalidate the above conclusions. Therefore,
the proposed TS change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The proposed amendment will not involve a significant
reduction in a margin of safety.
Although the proposed TS change results in changes to the
interval between surveillance tests, the impact, if any, on system
availability is minimal, based upon other, more frequent testing
that is performed, the existence of redundant systems and equipment,
or overall system reliability. Evaluations show there is no evidence
of time-dependent failures that would impact the availability of the
systems. The proposed change does not significantly impact the
condition or performance of structures, systems, and components
relied upon for accident mitigation.
A sensitivity analysis was performed to determine the effect of
the increased surveillance intervals on the HNP PRA. This
sensitivity analysis shows a negligible increase in CDF and
essentially no change in LERF due to the proposed change.
Furthermore, an historical review of surveillance test results
and associated maintenance records indicates there was no evidence
of any failure that would invalidate the above conclusions.
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: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: Richard J. Laufer, Acting.
Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: October 16, 2001.
Brief description of amendment request: The proposed amendment
would add a condition to the Operating License to extend certain
Technical Specification surveillance requirement (SR) intervals, one
time. The SR intervals would be extended up to 65 days, but no later
than April 30, 2003, to permit them to be performed during
[[Page 59515]]
the next refueling outage, which has been rescheduled because the plant
is currently in a forced extended outage.
Date of publication of individual notice in Federal Register:
November 13, 2001 (66 FR 56865).
Expiration date of individual notice: December 13, 2001.
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/NRC/ADAMS/index.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by email to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of application for amendment: December 29, 2000, as
supplemented March 22 and July 27, 2001.
Brief description of amendment: The amendment increases the allowed
outage time from 3 to 14 days for a single inoperable Division 1 or 2
diesel generator.
Date of issuance: November 8, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 141.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 24, 2001 (66 FR
7668). The supplemental letters contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 8, 2001.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: March 2, 2001, as supplemented
July 18, 2001.
Brief description of amendment: The amendment extends the
surveillance test interval of the slave relays of the Engineered Safety
Features Actuation System from 90 days to 8 months.
Date of issuance: November 5, 2001.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 198.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 11, 2001 (66 FR
36337).
The July 18, 2001, 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 amendment is contained
in a Safety Evaluation dated November 5, 2001.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: May 2, 2001, as supplemented by
letter dated August 23, 2001.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to not require the moderator temperature
coefficient (MTC) determination in TS 4.1.1.4.2c if the results of the
MTC determination required in TSs 4.1.1.4.2a and 4.1.1.4.2b are within
a certain tolerance of the corresponding design values.
Date of issuance: November 16, 2001.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 236.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 12, 2001 (66 FR
31706).
The August 23, 2001, supplemental letter provided clarifying
information that was within the scope of the original Federal Register
notice and did not change the staff's initial no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 16, 2001.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-237, Dresden Nuclear
Power Station, Unit 2, Grundy County, Illinois
Date of application for amendment: June 6, 2001, as supplemented by
letter dated September 17, 2001.
Brief description of amendment: The amendment revises the values of
the Safety Limit for the Minimum Critical Power Ratio in Technical
Specification Section 2.1.1.
Date of issuance: November 2, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 189.
Facility Operating License No. DPR-19: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 5, 2001 (66
FR 46479).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 2, 2001.
No significant hazards consideration comments received: No.
[[Page 59516]]
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: September 29, 2000, as
supplemented by letters dated March 1, July 13, August 9, August 13,
and October 17, 2001
Brief description of amendments:The amendments change the technical
specifications to reflect a change in fuel vendors from Siemens Power
Corporation to General Electric, and a transition to GE14 fuel. As part
of the transition, changes are made to the number of required automatic
depressurization system valves and to the time delay relay settings on
emergency core cooling system pumps. These changes were noticed in the
Federal Register on December 27, 2000 (65 FR 81908), August 22, 2001
(66 FR 44170), and August 23, 2001 (66 FR 44382).
Date of issuance: November 2, 2001
Effective date: As of the date of issuance and shall be implemented
following refueling outage 17.
Amendment Nos.: 188 and 183
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 27, 2000
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 2, 2001.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: May 30, 2001, as supplemented
September 10, 2001.
Brief description of amendments: The amendments change the
Technical Specifications (TS) Surveillance Requirement (SR) 3.6.1.1.3
and adds two new SRs, SR 3.6.1.1.4 and SR 3.6.1.1.5, covering the
testing of Suppression Chamber-Drywell Vacuum Breakers and the Drywell-
to-Suppression Chamber Bypass Leakage Test.
Date of issuance: November 7, 2001
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 149 and 135
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: July 25, 2001 (66 FR
38761).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments are contained
in a Safety Evaluation dated November 7, 2001.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: July 9, 2001
Brief description of amendments: These amendments revised the
current Technical Specifications of Limerick Generating Station, Units
1 and 2, to make them more consistent with changes to Title 10 of the
Code of Federal Regulations, Section 50.59.
Date of issuance: As of date of issuance and shall be implemented
within 60 days.
Effective date: November 1, 2001
Amendment Nos.: 154 and 118
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 22, 2001 (66 FR
44170).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 1, 2001.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
County, Pennsylvania
Date of application for amendments: July 9, 2001
Brief description of amendments: These amendments replaced the term
``unreviewed safety question'' with ``requires NRC approval pursuant to
10 CFR 50.59'' in order to provide consistency with the changes to 10
CFR 50.59, ``Changes, tests, and experiments,'' which became effective
on March 13, 2001.
Date of issuance: November 6, 2001
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendments Nos.: 242 and 246.
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications and the License.
Date of initial notice in Federal Register: August 22, 2001 (66 FR
44170).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 6, 2001.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: April 1, 2001, as supplemented
July 20, 2001.
Brief description of amendment: This amendment introduces new
Technical Specification 6.17, ``Technical Specification (TS) Bases
Control Program'' to provide consistency with the changes to 10 CFR
50.59 as published in the Federal Register (Volume 64, Number 191)
dated October 4, 1999.
Date of issuance: November 15, 2001.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 249.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 30, 2001 (66 FR
29356).
The supplemental letter contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 15, 2001.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: August 22, 2001.
Brief description of amendments: Revised Technical Specifications
Section 6.0, ``Administrative Controls,'' to change the title of the
corporate executive responsible for plant nuclear safety from
``President-Nuclear Division'' to ``Chief Nuclear Officer.''
Date of Issuance: November 13, 2001.
Effective Date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 178 and 121.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
[[Page 59517]]
Date of initial notice in Federal Register: October 3, 2001 (66 FR
50469).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 13, 2001.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: June 12, 2000, as supplemented
by letters dated November 7, 2000, June 19 and August 17, 2001.
Brief description of amendments: The amendments would use the
methodology and the alternative source term (AST) in 10 CFR 50.67 and
described in NUREG-1465, ``Accident Source Terms for Light-Water
Nuclear Power Plants,'' and Regulatory Guide 1081, ``Alternative
Radiological Source Terms for Evaluating the Radiological Consequences
of Design-Basis Accidents at Boiling and Pressurized Water Reactors.''
Implementing the AST of 10 CFR 50.67 results in a new acceptance
criterion for 10 CFR Part 50, Appendix A, General Design Criterion 19,
of 5 rem total effective dose equivalent. The licensee determined that
use of the revised analysis assumptions, methodology, and acceptance
criterion required prior Nuclear Regulatory Commission (NRC) approval.
In addition, the NRC requires in 10 CFR 50.67, a license amendment to
implement the AST as a replacement for the Technical Information
Document 14844 source term.
Date of issuance: November 13, 2001.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 258 and 241.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
approve changes to the updated final safety analysis report.
Date of initial notice in Federal Register: August 23, 2000 (65 FR
51356).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 13, 2001.
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: February 15, 2001.
Brief description of amendment: The amendment consists of deletion
of Operating License Condition 2.D, and revision to the Technical
Specifications (TSs) to remove depiction of railroad tracks in TS
Figure 4.1-1.
Date of issuance: November 16, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 190
Facility Operating License No. DPR-46: Amendment revised the
Operating License and the Technical Specifications.
Date of initial notice in Federal Register: June 27, 2001 (66 FR
34285).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 16, 2001.
No significant hazards consideration comments received: No.
Niagara Mohawk Power Corporation, Docket Nos. 50-220 and 50-410, Nine
Mile Point Nuclear Station, Unit Nos. 1 and 2, Oswego County, New York
Date of application for amendments: February 1, 2001; as
supplemented on March 1, March 16, March 29, April 5, April 27, May 30,
June 7, September 10, September 26, September 28, and November 2, 2001.
Brief description of amendments: The amendments changed the
operating licenses and associated documents to reflect the transfer of
Niagara Mohawk Power Corporation's (NMPC's) ownership interest in Nine
Mile Point Nuclear Station, Unit No. 1, the transfer of the ownership
interests of NMPC, New York State Electric and Gas Corporation,
Rochester Gas and Electric Corporation, and Central Hudson Gas &
Electric Corporation in Nine Mile Point Nuclear Station, Unit No. 2,
and the transfer of NMPC's operating authority for both units, to Nine
Mile Point Nuclear Station, LLC. The amendments and corresponding
license transfers were approved by the U.S. Nuclear Regulatory
Commission by Order dated June 22, 2001, and Supplemental Order dated
October 30, 2001.
Date of issuance: November 7, 2001.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 172 (for Unit 1), 100 (for Unit 2).
Facility Operating License Nos. DPR-63 and NPF-69: Amendments
revised the operating licenses (both units), Technical Specifications
(both units) and Environmental Protection Plan (Unit 2).
Date of initial notice in Federal Register: April 2, 2001 (66 FR
17584).
The staff's related evaluation of the amendments is contained in
two Safety Evaluations dated June 22 and October 30, 2001.
No significant hazards consideration comments received: Not
applicable.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of application for amendments: August 16, 2001.
Brief description of amendments: The amendments revised the
Technical Specifications by deleting Section 5.5.3, ``Post Accident
Sampling,'' and thereby eliminating the requirements to have and
maintain the post-accident sampling program. The amendments also
revised Section 5.5.2, ``Primary Containment Sources Outside
Containment,'' to reflect the elimination of requirements to maintain
the post accident sampling system.
Date of issuance: November 13, 2001.
Effective date: As of the date of issuance and shall be implemented
on or before June 28, 2002.
Amendment Nos.: 123 and 101.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 3, 2001 (66 FR
50472).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 13, 2001.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: August 2, 2001.
Brief description of amendments: The amendments revised the
Technical Specifications by deleting Section 6.8.3.d, ``Post Accident
Sampling,'' and thereby eliminate the requirements to have and maintain
the post-accident sampling program. The amendments also revise Section
6.8.3.a, ``Primary Containment Sources Outside Containment,'' to
reflect the elimination of requirements to maintain the post accident
sampling system.
Date of issuance: November 7, 2001.
[[Page 59518]]
Effective date: As of the date of issuance and shall be implemented
within 6 months of the date of issuance.
Amendment Nos.: Unit 1--133; Unit 2--122.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated November 7, 2001.
No significant hazards consideration comments received: No.
Note: The publication date for this notice will change from
every other Wednesday to every other Tuesday, effective January 8,
2002. The notice will contain the same information and will continue
to be published biweekly.
Dated at Rockville, Maryland, this 20th day of November 2001.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 01-29446 Filed 11-27-01; 8:45 am]
BILLING CODE 7590-01-P