[Federal Register Volume 66, Number 201 (Wednesday, October 17, 2001)]
[Notices]
[Pages 52794-52813]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-25957]
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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.
[[Page 52795]]
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from September 24, 2001 through October 5, 2001.
The last biweekly notice was published on October 3, 2001 (66 FR
50463).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the 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 November 16, 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
[[Page 52796]]
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 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, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: August 13, 2001.
Description of amendment request: The proposed amendment would
change the requirement to withdraw the first set of reactor vessel
surveillance specimens by deferring withdrawal for one additional
operating cycle.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The withdrawal in Fall 2003 refueling outage vice the March 2002
refueling outage and the deferral of the withdrawal of the vessel
surveillance specimens are not initiators of or precursors to any of
the accident scenarios presented in the [Updated Safety Analysis
Report] USAR. 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 a
catastrophic failure of any plant structure, system, or component.
The vessel surveillance specimens are used as the basis for the
pressure-temperature (P/T) curves. However, despite the deferral for
one cycle of withdrawal of the vessel surveillance specimens, the P/
T curves will continue to conservatively be established in
accordance with Regulatory Guide (RG) 1.99, ``Radiation
Embrittlement of Reactor Vessel Materials,'' Revision 2, as
described in the USAR. Therefore, this change does not involve an
increase in the probability of any accident previously evaluated.
The proposed change to the withdrawal schedule for the vessel
surveillance specimens postpones the collection of one of two sets
of data needed to confirm the basis of the P/T curves with no change
to the currently allowed P/T curves. The P/T curves that are in the
[Technical Specifications] TS will continue to be based on RG 1.99.
The deferral of the removal of the first set of specimens will not
affect the confirmation of the bases for the P/T curves because the
withdrawal schedule for the second set of specimens is not being
changed with this request. Because the basis for the P/T curves is
maintained, this proposed change does not impact or increase the
assumed radionuclide source term and will not result in an
unacceptable reduction in reactor vessel toughness. Therefore, this
change does not involve an increase in the consequences of any
accident previously evaluated.
In summary, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed deferral for one cycle of the removal of the vessel
surveillance specimens 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 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 change
does not create the possibility of a new or different kind of
accident from any previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
The deferral for one cycle of the withdrawal of the vessel
surveillance specimens does not affect the P/T curves, and therefore
does not affect the margin to safety for brittle fracture. Because
two sets of specimens are needed to confirm the basis for the P/T
temperatures and because the schedule for the withdrawal of the
second set of specimens is not changing, the P/T curves continue in
the interim to conform to RG 1.99. The proposed change does not
challenge the integrity of the fuel cladding, reactor coolant
pressure boundary that includes the reactor vessel, or the primary
containment.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The Nuclear Regulatory Commission (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: Robert Helfrich, Mid-West Regional Operating
Group, Exelon Generation Company, LLC, 1400 Opus Place, Suite 900,
Downers Grove, IL 60515
NRC Section Chief: Anthony J. Mendiola
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: September 17, 2001.
Description of amendment request: The proposed amendment would
modify the Technical Specification (TS) surveillance requirement for
the containment spray nozzles by changing the test frequency from
``once per 10 years'' to ``following activities that could result in
nozzle blockage.''
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:
[[Page 52797]]
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change revises the testing requirements for the
containment spray nozzles to only require verification that each
spray nozzle is unobstructed following activities that could result
in nozzle blockage. The only event for which the containment spray
system is considered an initiator is the maximum containment
negative pressure event. This event involves inadvertent actuation
of containment spray following a break in the reactor water cleanup
system inside containment described in Updated Safety Analysis
Report (USAR) Section 6.2.1.1.4.2. This change does not increase the
likelihood for an inadvertent actuation of the containment spray
system.
The proposed change does not have a detrimental impact on the
integrity of any plant structure, system, or component that
initiates an analyzed event. No active or passive failure mechanisms
that could lead to an accident are affected. The proposed change
will not alter the operation of, or otherwise increase the failure
probability of any plant equipment that initiates an analyzed
accident. As a result, the probability of any accident previously
evaluated, is not significantly increased.
The consequences of a previously evaluated accident are not
significantly increased. The proposed change revises the current
Surveillance Frequency from 10 years to following activities that
could result in spray nozzle blockage. Since activities that could
introduce foreign material into the system (such as inadvertent
actuation of the containment spray system or loss of foreign
material control) are the most likely cause for obstruction, testing
or inspection following such activities would verify the nozzle(s)
being unobstructed, and the system capable of performing its safety
function. No other evolutions require the system boundary to be
breached, so introduction of debris during times when maintenance
activities are not in progress are precluded. Introduction of
foreign materials into the system from the exterior is highly
unlikely due to the location of the spray headers, the passive
nature of the nozzles, and the fact that the containment spray
headers are maintained dry which does not lend itself to active
degradation mechanisms such as corrosion. The proposed testing
requirements are considered sufficient to provide a high degree of
confidence that containment spray flow will be available when
required. Therefore, the proposed change does not significantly
increase the consequences of an accident previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change to the test frequency for the containment
spray system nozzles does not involve the use or installation of new
equipment. Installed equipment is not operated in a new or different
manner. No new or different system interactions are created, and no
new processes are introduced. The current foreign material exclusion
practices have been reviewed and judged sufficient to provide high
confidence that debris will not be introduced during times when the
system boundary is breached.
Therefore, this proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Does the change involve a significant reduction in a margin of
safety?
The revision to the containment spray nozzle testing frequency
does not introduce any new setpoints at which protective or
mitigative actions are initiated. No current setpoints are altered
by this change. The design and functioning of the containment spray
system is unchanged. Since the system is not susceptible to
corrosion induced obstruction nor is the introduction of foreign
material from the exterior likely, the proposed testing frequency is
sufficient to provide high confidence that the containment spray
system will be available to provide the flow necessary to ensure
that the effects of drywell bypass leakage and low energy line
breaks are mitigated. Therefore, the capacity of the system will
remain unchanged. As a result, this change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Robert Helfrich, Mid-West Regional Operating
Group, Exelon Generation Company, LLC, 1400 Opus Place, Suite 900,
Downers Grove, IL 60515
NRC Section Chief: Anthony J. Mendiola
Carolina Power & Light Company, et al., Docket No. 50-325, Brunswick
Steam Electric Plant, Unit 1, Brunswick County, North Carolina
Date of amendment request: September 18, 2001.
Description of amendment request: The proposed amendments would
change the Technical Specifications (TS) to revise the Minimum Critical
Power Ratio (MCPR) Safety Limit values contained in TS 2.1.1.2, and
revise the MCPR Safety Limit values from 1.10 to 1.12 for two
recirculation loop operation and from 1.11 to 1.14 for single
recirculation loop operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed license amendments do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed license amendment will establish MCPR Safety Limit
values of 1.12 for two recirculation loop operation and 1.14 for
single recirculation loop operation. The revised MCPR Safety Limit
values have been determined using NRC-approved methods and
procedures. These procedures incorporate cycle-specific parameters
and reduced power distribution uncertainties in the determination of
the MCPR Safety Limit values. These proposed MCPR Safety Limit
values do not affect the operability of any plant systems nor do
these revised values compromise any fuel performance limits.
Therefore, the proposed change to the MCPR Safety Limit values does
not result in an increase in the probability of a previously
evaluated accident.
The consequences of a previously evaluated accident are
dependent on the initial conditions assumed for the analysis, the
behavior of the fuel during the accident, the availability and
successful functioning of the equipment assumed to operate in
response to the accident, and the setpoints at which these actions
are initiated. The MCPR Safety Limit values are determined to ensure
that 99.9 percent of the fuel rods will not experience boiling
transition during any plant operation if the limit is not exceeded.
Operational MCPR limits will be applied that ensure the MCPR Safety
Limit is not exceeded during all modes of operation and anticipated
operational occurrences. The MCPR Safety Limit does not impact the
source term or pathways assumed in accidents previously evaluated.
No analysis assumptions are violated, and there are no adverse
effects on the factors contributing to offsite and onsite dose. The
proposed change to the MCPR Safety Limit values does not affect the
performance of any equipment used to mitigate the consequences of a
previously evaluated accident. Also, the proposed change does not
affect setpoints that initiate protective or mitigative actions.
Based on the determination of the MCPR Safety Limit values using
conservative NRC-approved methods and the operability of plant
systems designed to mitigate the consequences of accidents not being
changed, the proposed changes to the MCPR Safety Limit values does
not significantly increase the consequences of a previously
evaluated accident.
2. The proposed license amendments will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors of
that accident. New accident precursors may be created by
modifications of the plant configuration, including changes in
allowable modes of operation. This proposed license amendment does
not involve any facility modifications, and plant equipment will not
be operated in a different manner. Also, no new initiating events or
transients result from the MCPR Safety Limit changes. As a result,
no new failure modes are being introduced. Therefore, the proposed
changes to the MCPR Safety Limit values will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
[[Page 52798]]
3. The proposed license amendments do not involve a significant
reduction in a margin of safety.
The margin of safety is established through the design of the
plant structures, systems, and components; through the parameters
within which the plant is operated; through the establishment of
setpoints for actuation of equipment relied upon to respond to an
event; and through margins contained within the safety analyses. The
proposed change to the MCPR Safety Limit values does not adversely
impact the performance of plant structures, systems, components, and
setpoints relied upon to respond to mitigate an accident. The MCPR
Safety Limit values have been calculated using NRC-approved methods
and procedures. The MCPR Safety Limit values are determined to
ensure that 99.9 percent of the fuel rods will not experience
boiling transition during any plant operation if the limits are not
exceeded, thereby ensuring that fuel cladding integrity is
maintained. Based on the assurance that the fuel design criteria are
being met, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602
NRC Section Chief: Richard P. Correia.
Dominion Nuclear Connecticut, Inc., Docket Nos. 50-245, 50-336, and 50-
423, Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3, New London
County, Connecticut
Date of amendment request: August 8, 2001.
Description of amendment request: The proposed amendment would
incorporate two changes into each operating license: (1) Revise the
physical protection (security) related license condition to indicate
that the physical security program plans listed, may, rather than do
contain, safeguards information, and (2) change the name of the
``Millstone Nuclear Power Station'' to the ``Millstone Power Station.''
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 first proposed clarification modifies the physical
protection (security) related license condition within the
respective operating license (OL) to indicate the physical security
program plans listed, may, rather than do contain, safeguards
information. The second proposed change to reflect the change in
name of the facility from the ``Millstone Nuclear Power Station'' to
the ``Millstone Power Station'' is editorial. Neither of these
changes alter any regulatory requirements or have an impact on the
acceptance criteria for any design basis accident described in the
respective Unit 2 or 3 Updated Final Safety Analysis Report (UFSAR)
or the Unit I Defueled Safety Analysis Report (DSAR).
These changes have no impact on plant equipment operation. Since
the changes are solely an administrative or editorial change to the
OL, they cannot affect the likelihood or consequences of accidents.
Therefore, these changes will not 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 changes have no impact on plant operation. Since
the proposed changes are solely an administrative or editorial
change to the OL, they do not affect plant operation in any way.
The changes do not alter the plant configuration (no new or
different type of equipment will be installed) or require any new or
unusual operator actions. The changes do not alter the way any
structure, system, or component functions and do not alter the
manner in which the plant is operated. The changes do not introduce
any new failure modes. Therefore, the proposed changes will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Since the proposed changes are solely a clarification or an
editorial change to the OL, they do not affect plant operation in
any way. The proposed changes do not impact any acceptance criteria
for the design basis accidents described in the respective Unit No.
2 or No. 3 UFSAR or the Unit No. 1 DSAR and do not impact the
consequences of accidents previously evaluated. Therefore, the
proposed changes will not result in a reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Dominion Nuclear Connecticut, Inc., Rope Ferry Road,
Waterford, Connecticut 06835.
NRC Section Chief: Stephen Dembek
Dominion Nuclear Connecticut Inc., et al., Docket Nos. 50-336 and 50-
423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London
County, Connecticut
Date of amendment request: August 9, 2001.
Description of amendment request: The proposed amendments modify
the Millstone Nuclear Power Station, Unit Nos. 2 (MP2) and 3 (MP3)
Technical Specifications (TSs) to avoid confusion between the
qualification standards of the facility staff, who are qualified to
American National Standards Institute (ANSI) N18.1-1971/Regulatory
Guide (RG) 1.8 Revision 0, and the operators who will be qualified to
the education and experience guidelines outlined by National Academy
for Nuclear Training ACAD 00-003 ``Guidelines for Initial Training and
Qualification of Licensed Operators.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed administrative clarification modifies the Unit Nos.
2 and 3 TS to avoid confusion between the qualification standards of
the facility staff, who are qualified to American National Standards
Institute (ANSI), ``Selection and Training of Nuclear Power Plant
Personnel,'' ANSI N18.1-1971/Regulatory Guide 1.8, Revision 0,
``Qualification and Training of Personnel for Nuclear Power
Plants,'' and the operators who will be qualified to the education
and experience guidelines outlined by National Academy for Nuclear
Training (NANT 2000 Guidelines), ACAD 00-003, ``Guidelines for
Initial Training and Qualification of Licensed Operators.'' The
training of the operators themselves is not affected, this change
only modifies the education and experience requirements they must
meet to qualify for the operator training program. The reactor
operator and senior reactor operator applicant (or upgrade) still
must learn and are tested on the same material, demonstrate their
proficiency on the facility simulator and meet other requirements.
Consequently, this change has no impact on the capability of
licensed operators, it only modifies and provides alternative
qualifications for entry into the program.
This change will not alter any regulatory requirements or have
an impact on the acceptance criteria for any design basis accident
described in the respective Unit Nos. 2 or 3 Updated Final Safety
Analysis Report (UFSAR).
The change has no impact on plant equipment operation. Since the
change is solely an administrative change to the Technical
Specifications, it cannot affect the
[[Page 52799]]
likelihood or consequences of accidents. Therefore, this change will
not increase the probability or consequences of an accident
previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change has no impact on plant operation. Since the
proposed change is solely an administrative change to the Technical
Specifications, it does not affect plant operation in any way.
The change does not alter the plant configuration (no new or
different type of equipment will be installed) or require any new or
unusual operator actions. The change does not alter the way any
structure, system, or component functions and does not alter the
manner in which the plant is operated. The change does not introduce
any new failure modes. Therefore, the proposed change will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
Since the proposed change is solely an administrative change to
the Technical Specifications, it does not affect plant operation in
any way.
The proposed change does not impact any acceptance criteria for
the design basis accidents described in the respective Unit Nos. 2
or 3 UFSAR and does not impact the consequences of accidents
previously evaluated. Therefore, the proposed change will not result
in a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
NRC Section Chief: James W. Clifford.
Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: June 28, 2001.
Description of amendment request: The proposed amendment modifies
the Millstone Nuclear Power Station, Unit No. 3 (MP3) Technical
Specifications to remove the surveillance requirement associated with
post maintenance testing of the containment isolation valves.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed Technical Specification change to remove the
surveillance requirement to perform post maintenance testing of the
containment isolation valves will not cause an accident to occur and
will not result in any change in the operation of the associated
accident mitigation equipment. The containment isolation valves are
not accident initiators. The proposed change will not revise the
operability requirements (e.g., valve stroke time) for the
containment isolation valves. Proper operation of the containment
isolation valves will still be verified, as appropriate, following
maintenance activities. As a result, the design basis accidents will
remain the same postulated events described in the Millstone Unit
No. 3 Final Safety Analysis Report, and the consequences of the
design basis accidents will remain the same. Therefore, the proposed
change will not increase the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change to the Technical Specifications does not
impact any system or component that could cause an accident. The
proposed change will not alter the plant configuration (no new or
different type of equipment will be installed) or require any
unusual operator actions. The proposed change will not alter the way
any structure, system, or component functions, and will not
significantly alter the manner in which the plant is operated. The
response of the plant and the operators following an accident will
not be different. In addition, the proposed change does not
introduce any new failure modes. Therefore, the proposed change will
not create the possibility of a new or different kind of accident
from any accident previously analyzed.
3. Involve a significant reduction in a margin of safety.
The proposed Technical Specification change to remove the
surveillance requirement to perform post maintenance testing of the
containment isolation valves will not cause an accident to occur and
will not result in any change in the operation of the associated
accident mitigation equipment. The operability requirements for the
containment isolation valves have not been changed, and proper
operation of the containment isolation valves will still be
verified, as appropriate, following maintenance activities. The
containment isolation valves will continue to be able to mitigate
the design basis accidents as assumed in the safety analysis.
Therefore, the proposed change will not result in a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
NRC Section Chief: James W. Clifford.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: September 7, 2001.
Description of amendment request: The proposed amendment would
change the Technical Specifications Regarding the Post Accident
Monitoring Instrumentation (Table 3.3.3-1 of Section 3.3.3. ``Post
Accident Monitoring Instrumentation''). Specifically, the proposed
amendment would reword the number of required channels stated for the
core exit thermocouples (CETs) to be the same as the Standard Technical
Specifications; delete notes that describe the redundant channels for
the Reactor Coolant System (RCS) Hot Leg Temperature, the RCS Cold Leg
Temperature and Main Steam Line Radiation and modify the note
pertaining to the redundant channel for Steam Generator Level (Wide
Range) to clarify what Condition Statements apply when the instrument
channel and/or the Auxiliary Feedwater Flow instrument channel is
inoperable. Other existing notes in the Table are proposed to be
renumbered to accommodate the above changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Consistent with the criteria of 10 CFR 50.92, the enclosed
[proposed] application is judged to involve no significant hazards
based on the following information:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response: The proposed amendment involves rewording or
clarification of technical specification requirements to properly
reflect the design of post accident monitoring instrumentation at
Indian Point 3. The proposed rewording of the required channels for
core exit thermocouples adopts the wording from the Standard
Technical Specifications, which is applicable to the Indian Point 3
design. The proposed deletion of Notes (a), (b), and (g) removes
design information that is not needed for the specification to limit
plant operation in response to inoperable instrument channels. The
proposed rewording of Note (f) clarifies the existing requirement by
making a more explicit statement about the applicable conditions for
the affected functions.
[[Page 52800]]
Renumbering other Table notes is an editorial change to keep the
notes in sequential order.
The proposed amendment does not involve any changes to plant
equipment, setpoints, or the way in which the plant is operated.
These changes do not affect accident initiators or accident
mitigating systems. Therefore, the proposed amendment does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response: The proposed amendment involves rewording or
clarification of technical specification requirements to properly
reflect the design of post accident monitoring instrumentation at
Indian Point 3. The proposed amendment does not involve any changes
to plant equipment, setpoints, or the way in which the plant is
operated. These changes do not affect accident initiators or
accident mitigating systems. Therefore the proposed amendment does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
(3) Does the proposed license amendment involve a significant
reduction in a margin of safety?
Response: The proposed amendment involves rewording or
clarification of technical specification requirements to properly
reflect the design and licensing basis of post accident monitoring
instrumentation at Indian Point 3. The proposed rewording of the
required channels for core exit thermocouples adopts the wording
from the Standard Technical Specifications, which is applicable to
Indian Point 3. This will ensure that appropriate condition
statements are entered in the event that core exit thermocouples
become inoperable. Notes (a), (b), and (g) provide design
information that is not needed in the specification for plant
operators to enter appropriate condition statements when inoperable
instrument channels in the affected functions are identified. The
rewording of Note (f) more clearly states the existing requirement
and makes no change to the required actions or completion times for
the associated inoperable instrument channels.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Generating Station, 600 Rocky Hill Road, Plymouth, MA
02360.
NRC Section Chief: Lakshminaras Raghavan, Acting.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: August 13, 2001.
Description of amendment request: The proposed amendments would
revise technical specifications (TS) to support a planned upgrade to
the reactor water level instrumentation. Currently, many low-level
actuation functions use Yarway level indicating switches. This includes
emergency core cooling system (ECCS), reactor core isolation cooling
(RCIC) and feedwater systems. The Yarways will be replaced with more
reliable analog level transmitters and additional electronic trip
units. The upgrade will provide sensing devices for reactor vessel
water level signals and indications that are more reliable with less
drift and will require less frequent surveillance requirements. The
proposed changes align the TS surveillance requirements with the
instrumentation upgrades. This includes changes to calibration
frequencies, functional testing and allowable values.
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?
During the upcoming refueling outages at Quad Cities Nuclear
Power Station (Unit 1 and Unit 2), a design change will be
implemented that upgrades the existing reactor vessel level trip
instrumentation used in various applications at Quad Cities Nuclear
Power Station, including the Emergency Core Cooling System (ECCS),
Reactor Core Isolation Cooling System (RCIC) and Feedwater systems.
Technical Specification (TS) requirements that govern
operability or routine testing of plant instruments are not assumed
to be initiators of any analyzed event because these instruments are
intended to prevent, detect, or mitigate accidents. Therefore, these
changes will not involve an increase in the probability of
occurrence of an accident previously evaluated. Additionally, these
changes will not increase the consequences of an accident previously
evaluated because the proposed change does not adversely impact
structures, systems, or components (SSCs). The planned instrument
upgrade is a more reliable design than existing equipment. The
proposed TS change maintains existing requirements that ensure
components are operable when necessary for the prevention or
mitigation of accidents or transients. Revised allowable values for
the associated functions have been established in accordance with
EGC's setpoint methodology, which is consistent with industry
standards. The setpoint methodology establishes TS allowable values
that assure systems structures and components (including initiation
and trip functions) respond in a manner consistent with the plant
safety analysis. Furthermore, there will be no change in the types
or significant increase in the amounts of any effluents released
offsite. For these reasons, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed changes support a planned instrumentation upgrade.
The change provides revised Surveillance Requirements to ensure
operability. The change does not adversely impact the manner in
which the instrument will operate under normal and abnormal
operating conditions. These changes reflect the improved performance
of the instrumentation upgrade and provide an equivalent level of
safety. The changes in methods governing normal plant operation are
consistent with the current safety analysis assumptions. Therefore,
these changes will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The proposed change supports a planned instrumentation upgrade.
The proposed change does not affect the probability of failure or
availability of the affected instrumentation. The change to an
analog trip system to monitor reactor vessel level provides for
increased reliability. The change has no impact on the underlying
design functions. The proposed TS surveillance requirements are
consistent with current TS requirements for functions that employ
analog trip unit devices. The proposed allowable values have been
established in accordance with EGC's setpoint methodology, which
considers instrument design and performance characteristics. The
methodology establishes TS allowable values with sufficient margin
to assure that the plant safety analysis assumptions (e.g., certain
initiation and trip functions) are maintained. As such, the trip and
actuation functions continue to ensure design basis requirements are
maintained. Therefore, it is concluded that the proposed changes
will not result in a reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Vice President,
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way,
Kennett Square, PA 19348
[[Page 52801]]
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334,
Beaver Valley Power Station, Unit No. 1 (BVPS-1), Beaver County,
Pennsylvania
Date of amendment request: June 29, 2001.
Description of amendment request: The proposed license amendment
would change the technical specifications (TSs) to reflect revised
reactor coolant system (RCS) heatup and cooldown pressure and
temperature (P/T) limit curves that will be valid through 22 effective
full power years (EFPYs). The overpressure protection system (OPPS)
power-operated relief valve (PORV) setpoints and the OPPS enabling
temperature would also be revised. The proposed BVPS-1 P/T limits
incorporate the results from the testing of the Capsule Y described in
WCAP-15571, ``Analysis of Capsule Y from Beaver Valley Unit 1 Reactor
Vessel Radiation Surveillance Program,'' Revision 0, November 2000.
These changes have been prepared using the Nuclear Regulatory
Commission-approved methodology described in WCAP-14040-NP-A,
``Methodology Used to Develop Cold Overpressure Mitigating System
Setpoints and RCS Heatup and Cooldown Limit Curves,'' Revision 2,
January 1996, with two exceptions. These exceptions include the use of
(1) the American Society of Mechanical Engineers (ASME) Code Case N-
640, ``Alternate Reference Fracture Toughness for Development of P-T
Curves for Section XI, Division 1,'' March 1999, and (2) the ASME
Boiler and Pressure Vessel Code, Section XI, ``Rule for Inservice
Inspection of Nuclear Power Plant Components,'' Appendix G, ``Fracture
Toughness Criteria for Protection Against Failure,'' December 1995
(through 1996 Addendum). The TS Bases and Figure Index will also be
changed to reflect the revisions discussed above.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed changes do not result in physical changes being
made to structures, systems, or components (SSCs), or to event
initiators or precursors. Changing the heatup and cooldown curves,
power operated relief valve (PORV) setpoint and overpressure
protection system (OPPS) enable temperature to reflect 22 effective
full power years (EFPY) will not affect the ability of the OPPS to
control the reactor coolant system (RCS) at low temperatures such
that the integrity of the reactor coolant pressure boundary (RCPB)
is not compromised by violating the pressure/temperature (P/T)
limits. These changes were determined in accordance with the
methodologies set forth in the regulations to provide an adequate
margin of safety to ensure the reactor vessel will withstand the
effects of normal cyclic loads due to temperature and pressure
changes as well as the loads associated with postulated faulted
events.
Also, the proposed changes do not impact the design of plant
systems such that previously analyzed SSCs would now be more likely
to fail. The initiating conditions and assumptions for accidents
described in the Updated Final Safety Analysis Report (UFSAR) remain
as previously analyzed. Thus, the proposed changes do not involve a
significant increase in the probability of an accident previously
evaluated.
The proposed changes do not alter any assumptions previously
made in the radiological consequence evaluations nor affect
mitigation of the radiological consequences of an accident described
in the UFSAR. As such, the consequences of accidents previously
evaluated in the UFSAR will not be increased and no additional
radiological source terms are generated. Therefore, there will be no
reduction in the capability of those SSCs in limiting the
radiological consequences of previously evaluated accidents and
reasonable assurance that there is no undue risk to the health and
safety of the public will continue to be provided. Thus, the
proposed changes do not involve a significant increase in the
consequences of an accident previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. The proposed changes do not involve physical changes to
analyzed SSCs or changes to the modes of plant operation defined in
the technical specification. The proposed changes do not involve the
addition or modification of plant equipment (no new or different
type of equipment will be installed) nor do they alter the design or
operation of any plant systems. No new accident scenarios, accident
or transient initiators or precursors, failure mechanisms, or
limiting single failures are introduced as a result of the proposed
changes.
The proposed changes do not cause the malfunction of safety-
related equipment assumed to be operable in accident analyses. No
new or different mode of failure has been created and no new or
different equipment performance requirements are imposed for
accident mitigation. As such, the proposed changes have no effect on
previously evaluated accidents.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
No. The proposed changes have been determined through supporting
analyses to be in accordance with the methodologies set forth in the
regulations. Compliance with NRC approved methodologies provide for
an adequate margin of safety and ensure the reactor vessel will
withstand the effects of normal cyclic loads due to temperature and
pressure changes as well as the loads associated with postulated
faulted events as described in the UFSAR.
The new heatup and cooldown curves define the limits for
ensuring prevention of nonductile failure for the BVPS Unit No. 1
reactor vessel and do not significantly reduce the margin of safety
for the plant.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: L. Raghavan (Acting).
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: September 27, 2001.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to (1) change the diesel fuel
supply volume required for diesel generator (DG) operability, (2)
clarify existing wording, (3) add a TS limiting condition for operation
(LCO) and a TS surveillance requirement (SR) regarding DG air
receivers, (4) delete a current TS SR concerning DG starting air
compressors, and (5) restructure and renumber the TS LCOs and SRs for
applicability and administrative purposes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of a[n] accident
previously evaluated.
The proposed Technical Specification changes do not introduce
new equipment or new equipment operating modes, nor do the proposed
changes alter existing system
[[Page 52802]]
relationships. The proposed amendment does not introduce new failure
modes.
The proposed revision to the Monticello TS[s] renumbers and
relocates TS[s] as appropriate to provide a more understandable TS,
deletes an existing TS SR which does not satisfy the requirements of
10 CFR 50.36 for inclusion in the TS[s], adds a new TS LCO and SR
for DG air start receivers which more appropriately complies with
the requirements of 10 CFR 50.36, and revises the minimum number of
gallons of diesel fuel required in the Diesel Oil Storage Tank for
the DG to be declared operable.
Therefore, the proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
The proposed changes do not introduce a new mode of plant
operation, or involve a physical modification to the plant. The
proposed Technical Specification changes do not introduce new
equipment, nor do the proposed changes alter existing system
relationships. The proposed amendment does not introduce new failure
modes.
Therefore, the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously analyzed.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
The proposed changes maintain the current TS requirements for
safe operation of the Monticello plant. The proposed changes do not
involve a physical modification to the plant, or a new mode of
operation. The proposed changes do not alter the scope of equipment
currently required to be operable nor do the proposed changes affect
equipment safety functions. The proposed Technical Specification
changes do not introduce new equipment, nor do the proposed changes
alter existing system relationships. The proposed amendment does not
introduce new failure modes.
Therefore, these proposed changes will not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: William D. Reckley
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: August 17, 2001.
Description of amendment request: The proposed Technical
Specifications (TS) change will: (1) modify Salem TS surveillance
requirement 4.6.2.3, and (2) revise the associated TS Bases.
Specifically, the proposed change will modify the current acceptance
criterion for the service water flow rate through the Containment Fan
Coil Units from 2,550 gallons per minute (gpm) to
2,300 gpm.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is
presented below:
1. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The containment ventilation system, including the containment fan
coil units is not an accident initiator.
The proposed TS change to modify the Salem TS surveillance
requirement 4.6.2.3 to the service water-cooling water flow through the
fan coil units is bounded by the present licensing and design bases
analyses. The new proposed flow rate, in conjunction with its
associated heat exchanger thermal fouling factor, will continue to
maintain the assumed minimum containment heat removal capability to be
within the Salem Updated Final Safety Analysis Report (UFSAR) Chapter
15 analyses. Therefore, the proposed change will not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
2. Does not create the possibility of a new or different kind of
accident from any accident previously analyzed.
The proposed TS change to modify the Salem TS surveillance
requirement 4.6.2.3 to the service water-cooling water flow though the
fan coil units is bounded by the present licensing and design bases
analyses. The manner and frequency at which the surveillance test is
conducted remains unchanged. The physical facility remains unchanged.
Therefore, the new proposed flow rate does not create the
possibility of a new or different kind of accident from any accident
previously analyzed.
3. Does not involve a significant reduction in a margin of safety.
The proposed TS change to modify the Salem TS surveillance
requirement 4.6.2.3 to the service water-cooling water flow though the
fan coil units is bounded by the present licensing and design bases
analyses. The new proposed flow rate, in conjunction with its
associated heat exchanger thermal fouling factor, will continue to
maintain the assumed minimum containment heat removal capability to be
within the Salem UFSAR Chapter 15 analyses. Consequently, the existing
margins of safety with respect to the current design-basis assumptions
of pressure of 47 psig and a saturation temperature of 271 deg.F in
containment during a design-basis accident is maintained.
Therefore, the proposed changes do 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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket No. 50-321, Edwin I. Hatch Nuclear
Plant, Unit 1, Appling County, Georgia
Date of amendment request: August 31, 2001.
Description of amendment request: The proposed amendment would
allow a one-time deferral of the Type A Containment Integrated Leak
Rate Test based on the risk-informed guidance in Regulatory Guide
1.174.
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 Technical Specification change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
The proposed revision to Technical Specification 5.5.12
(``Primary Containment Leakage Rate Testing Program'') involves a
one-time extension to the current interval for Type A containment
testing. The current test interval of ten (10) years would be
extended on a one-time basis to no longer than fifteen (15) years
from the last Type A test. The proposed Technical Specification
change does not involve a physical change to the plant or a change
in the manner which the
[[Page 52803]]
plant is operated or controlled. The reactor containment is designed
to provide an essentially leak tight barrier against the
uncontrolled release of radioactivity to the environment for
postulated accidents. As such the reactor containment itself and the
testing requirements invoked to periodically demonstrate the
integrity of the reactor containment exist to ensure the plant's
ability to mitigate the consequences of an accident, and do not
involve the prevention or identification of any precursors of an
accident. Therefore, the proposed Technical Specification change
does not involve a significant increase in the probability of an
accident previously evaluated.
The proposed change involves only the extension of the interval
between Type A containment leakage tests. Type B and C containment
leakage tests will continue to be performed at the frequency
currently required by plant Technical Specifications. Industry
experience has shown, as documented in NUREG-1493, that Type B and C
containment leakage tests have identified a very large percentage of
containment leakage paths and that the percentage of containment
leakage paths that are detected only by Type A testing is very
small. HNP Unit 1 ILRT test history supports this conclusion. NUREG-
1493 concluded, in part, that reducing the frequency of Type A
containment leak tests to once per twenty (20) years leads to an
imperceptible increase in risk. The integrity of the reactor
containment is subject to two types of failure mechanisms which can
be categorized as (1) activity based and (2) time based. Activity
based failure mechanisms are defined as degradation due to system
and/or component modifications or maintenance. Local leak rate test
requirements and administrative controls such as design change
control and procedural requirements for system restoration ensure
that containment integrity is not degraded by plant modifications or
maintenance activities. The design and construction requirements of
the reactor containment itself combined with the containment
inspections performed in accordance with ASME Section XI, the
Maintenance Rule and the containment coatings program serve to
provide a high degree of assurance that the containment will not
degrade in a manner that is detectable only by Type A testing.
Therefore, the proposed Technical Specification change does not
involve a significant increase in the 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 revision to the Technical Specifications involves a
one-time extension to the current interval for Type A containment
testing. The reactor containment and the testing requirements
invoked to periodically demonstrate the integrity of the reactor
containment exist to ensure the plant's ability to mitigate the
consequences of an accident and do not involve the prevention or
identification of any precursors of an accident. The proposed
Technical Specification change does not involve a physical change to
the plant or the manner in which the plant is operated or
controlled. Therefore, the proposed Technical Specification 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 revision to Technical Specifications involves a
one-time extension to the current interval for Type A containment
testing. The proposed Technical Specification change does not
involve a physical change to the plant or a change in the manner in
which the plant is operated or controlled. The specific requirements
and conditions of the Primary Containment Leakage Rate Testing
Program, as defined in Technical Specifications, exist to ensure
that the degree of reactor containment structural integrity and
leak-tightness that is considered in the plant safety analysis is
maintained. The overall containment leakage rate limit specified by
Technical Specifications is maintained. The proposed change involves
only the extension of the interval between Type A containment
leakage tests. Type B and C containment leakage tests will continue
to be performed at the frequency currently required by plant
Technical Specifications.
HNP Unit 1 and industry experience strongly supports the
conclusion that Type B and C testing detects a large percentage of
containment leakage paths and that the percentage of containment
leakage paths that are detected only by Type A testing is small. The
containment inspections performed in accordance with ASME Section
XI, the Maintenance Rule and the Coatings Program serve to provide a
high degree of assurance that the containment will not degrade in a
manner that is detectable only by Type A testing. Additionally, the
on-line containment monitoring capability that is inherent to
inerted BWR containments allows for detection of gross containment
leakage that may develop during power operation. The combination of
these factors ensures that the margin of safety that is inherent in
plant safety analysis is maintained. Therefore, the proposed
Technical Specification 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 L. Emch, Jr.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: August 31, 2001.
Description of amendment request: The proposed amendments would
extend the completion times for the required actions associated with
restoring an inoperable emergency diesel generator.
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?
The proposed changes extend the Technical Specifications
required Completion Times for restoration of an inoperable emergency
diesel generator (DG) to a maximum of 14 days. Additionally, the
proposed extension of the Completion Time to 14 days results in a
corresponding extension of the time period associated with discovery
of failure to meet Limiting Condition for Operation 3.8.1 to 17
days. (This provides a maximum time limit for overlapping
inoperabilities of DGs and offsite sources.)
For both Plant Hatch units A and C DGs, to utilize the 72 hours
to 14 day period of the proposed extended Completion Time,
compensatory action is required to ensure two DGs per unit remain
available. This action consists of dedicating the 1B DG to that unit
with the inoperable DG. This means that the 1B DG will be inhibited
from an automatic swap to the opposite unit when that unit (the non-
maintenance unit) experiences and undervoltage condition on its F
4160 volt bus, regardless of the presence or absence of a loss of
coolant accident (LOCA) signal. Inhibiting the automatic transfer
makes the 1B DG inoperable (with a Completion Time of 14 days) for
the non-maintenance unit.
Completion Times are not an initial condition or assumption of
any analyzed event. DGs are not initiators of any analyzed event. No
event mitigation assumes more than two DGs per unit. The
consequences of an accident are independent of the time the DGs are
out of service provided adequate DG availability is assured.
Compensatory actions are proposed in this amendment request that
ensure adequate DG availability for both Plant Hatch units.
Therefore, the assumptions regarding DG available are maintained.
To fully evaluate the effect of the proposed DG Completion Time
extension, Probabilistic Safety Assessment methods and a
deterministic analysis were utilized. The
[[Page 52804]]
results of the analyses show no significant increase in Core Damage
Frequency (CDF) and Large Early Release Frequency (LERF).
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an event previously
analyzed.
2. Do the proposed changes create the possibility of a new or
different type of accident from any previously evaluated?
The proposed changes do involve a change to the plant
configuration when either unit's A or C DG is utilizing the extended
Completion Time (i.e., inoperable in excess of 72 hours). That
configuration change ensures that both units have two dedicated DGs.
Furthermore, affixing the 1B DG to one unit will cause it (1B DG) to
be inoperable with respect to the Technical Specifications. Ensuring
two DGs available to each unit for event mitigation in no way
creates the possibility of a new or different type of accident.
No other change in the design, configuration, or method of
operation of the plant is introduced by the proposed change. The
changes do not alter any assumptions made in the safety analyses. No
new failure modes are introduced.
Therefore, the proposed changes do not create the possibility of
a new or different type of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
Since all assumptions of the plant event analyses are
maintained, there is no effect on the margin of safety in any safety
analyses. If there is any margin of safety ascribed to DG
availability and plant risk, it has been determined that such a
margin of safety is not significantly reduced, as the proposed
changes have been evaluated both deterministically and using a risk-
informed approach. These evaluations concluded the following with
respect to the proposed changes:
Applicable regulatory requirements will continue to be met,
adequate defense-in-depth will be maintained, sufficient safety
margins will be maintained, and any increases in CDF and LERF are
small and consistent with the NRC Safety Goal Policy Statement
(Federal Register, Vol. 51, p. 30028 (51 FR 30028), August 4, 1986,
as interpreted by NRC Regulatory Guides 1.174 and 1.177).
Furthermore, increases in risk posed by potential combinations of
equipment out of service during the proposed DG extended Completion
Time will be managed by the site configuration risk management
procedure, consistent with 10 CFR 50.65, ``Requirements for
Monitoring the Effectiveness of Maintenance at Nuclear Power
Plants,'' paragraph (a)(4).
The availability of offsite power together with the availability
of the other DGs and the use of on-line risk assessment tools
provide adequate compensation for the potential small incremental
increase in plant risk of the extended DG Completion Time. In
addition, the increased availability of the DGs during refueling
outages offsets the small increase in plant risk during operation.
The proposed extended DG Completion Times, in conjunction with the
availability of the other DGs continues to provide adequate
assurance of the capability to provide power to the engineered
safety features buses. Therefore, implementation of the proposed
changes will not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: Richard J. Laufer, Acting.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: September 7, 2001.
Description of amendment request: The proposed amendment would
revise Technical Specification Section 3.6.11, ``Ice Bed,''
Surveillance Requirement (SR) 3.6.11.2, SR 3.6.11.3, and the associated
Bases, to lower the minimum required average ice basket weight from
1236 pounds to 1110 pounds, and the corresponding total weight of the
stored ice in the ice condenser from 2,403,800 pounds to 2,158,000
pounds.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The primary purpose of the ice bed is to provide a large heat
sink to limit peak containment pressure in the event of a release of
energy from a design basis loss-of-coolant (LOCA) or high energy
line break (HELB) in containment. The LOCA requires the greatest
amount of ice compared to other accident scenarios, therefore the
reduction in ice weight is based on the LOCA analysis. The amount of
ice in the bed has no impact on the initiation of an accident, but
rather on the mitigation of the accident.
The containment integrity analysis shows that the proposed
reduced ice weight is sufficient to maintain the peak containment
pressure below the containment design pressure, and that the
containment heat removal systems function to rapidly reduce the
containment pressure and temperature in the event of a LOCA.
Therefore, containment integrity is maintained and the consequences
of an accident previously evaluated in the Updated Final Safety
Analysis Report (UFSAR) are not significantly increased.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The ice condenser serves to limit the peak pressure inside
containment following a LOCA. TVA has evaluated the revised
containment pressure analysis and determined that sufficient ice
would be present to maintain the peak containment pressure below the
containment design pressure. Therefore, the reduced ice weight does
not create the possibility of an accident that is different than any
already evaluated in the WBN UFSAR. No new accident scenarios,
failure mechanisms, or limiting single failures are introduced as a
result of this proposed change.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The containment integrity analysis for reduced ice weight
results in a peak containment pressure that is slightly lower than
that in the previous analysis of record. This reduction in peak
pressure, along with the ice weight reduction, is due to the removal
of analytical conservatism combined with a better segmental
representation of the mass and energy release transient from the
computer models.
The revised technical specifications ice weight surveillance
limits are based on the ice weight assumed in the containment
integrity analysis, with margin included for sublimation that is
based on actual sublimation data from the first three refueling
cycles at WBN. The analysis further demonstrates that the existing
relationship between ice bed melt-out and containment spray
switchover has been conservatively maintained. With the reduced ice
inventory, melt-out of the ice bed following a worst case large
break LOCA has been determined to occur after the switchover of
containment spray to the recirculation mode. Thus, the reduced ice
bed mass does not result in a reduction in the margin for operator
action to effect the switchover.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: August 24, 2001.
Brief description of amendments: The proposed change would revise
Comanche Peak Steam Electric Station,
[[Page 52805]]
Units 1 and 2, Technical Specification (TS) 3.3.2, entitled ``ESFAS
[Engineered Safety Features Actuation System] Instrumentation,'' and TS
3.3.6, entitled ``Containment Ventilation Isolation Instrumentation,''
to change the surveillance frequency for Westinghouse Electric Company-
type AR relays, used as Solid State Protection System slave relays or
auxiliary relays, from quarterly to refueling outage frequency.
Surveillance Requirements 3.3.2.6 and 3.3.6.5 would be revised to
change the frequency from ``92 days'' to ``92 days OR 18 months for
Westinghouse type AR relays.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the Technical Specifications does not
result in a condition where the design, material, or construction
standards that were applicable prior to the change are altered. The
same ESFAS instrumentation is being used and the same ESFAS system
reliability is expected. The proposed change will not modify any
system interface or function and could not increase the likelihood
of an accident since these events are independent of this change.
The proposed activity will not change, degrade or prevent the
performance of any accident mitigation systems or alter any
assumptions previously made in evaluating the radiological
consequences of an accident as described in the safety analysis
report.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not alter the performance of the ESFAS
mitigation systems assumed in the plant safety analysis. Changing
the interval for periodically verifying ESFAS slave relays (assuring
equipment operability) will not create any new accident initiators
or scenarios.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not affect the total ESFAS system
response assumed in the safety analysis. The periodic slave relay
functional verification is relaxed because of the demonstrated high
reliability of the relay and its insensitivity to any short term
wear or aging effects.
Therefore the proposed change does not involve a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: September 27, 2001 (WO 01-0038)
Description of amendment request: The proposed amendment would
revise Section 5.3.1.1 of the Technical Specifications to replace the
current qualifications in ANSI/ANS 3.1-1981 for licensed operators and
senior operators with the National Academy for Nuclear Training,
``Guidelines for Initial Training and Qualification of Licensed
Operators,'' dated January 2000.
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 TS change is an administrative change to clarify
the current requirements for licensed operator qualifications and
licensed operator training program. [The change conforms] to the
current requirements of 10 CFR 55.
Although licensed operator qualifications and training may have
an indirect impact on accidents previously evaluated, the NRC
considered this impact during the rulemaking process, and by
promulgation of the revised 10 CFR 55 rule, concluded that this
impact remains acceptable as long as the licensed operator training
program is certified to be accredited and is based on a systems
approach to training. WCNOC's [Wolf Creek Nuclear Operating
Corporation's] licensed operator training program is accredited by
INPO [Institute for Nuclear Power Operations] and is based on a
system[']s approach to training. The proposed TS change takes credit
for the INPO accreditation of the licensed operator training
program. The TS requirements for all other unit staff qualifications
remain unchanged.
Therefore, the proposed change does not involve a signification
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed TS change is an administrative change to clarify
the current requirements for licensed operator qualifications and
licensed operator training program and to conform to the revised 10
CFR 55.
As noted above, although licensed operator qualifications and
training may have an indirect impact on the possibility of a new or
different kind of accident from any accident previously evaluated,
the NRC considered this impact during the rulemaking process, and by
promulgation of the revised [10 CFR 55] rule, concluded that this
impact remains acceptable as long as the licensed operator training
program is certified to be accredited and based on a system[']s
approach to training. As previously noted, WCNOC's licensed operator
training program is accredited by INPO and is based on a system[']s
approach to training. The proposed TS change takes credit for the
INPO accreditation of the licensed operator training program. The TS
requirements for all other unit staff qualifications remain
unchanged.
Additionally, the proposed TS change does not affect plant
design, hardware, system operation, or procedures. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed TS change is an administrative change to clarify
the current requirements applicable to licensed operator
qualifications and licensed operator training program. This change
is consistent with the requirements of 10 CFR 55. The TS
qualification requirements for all other unit staff remain
unchanged.
Licensed operator qualifications and training can have an
indirect impact on a margin of safety. However, the NRC considered
this impact during the rulemaking process, and by promulgation of
the revised 10 CFR 55 [rule], determined that this impact remains
acceptable when licensees maintain a licensed operator training
program that is accredited and based on a system[']s approach to
training. As noted previously, WCNOC's licensed operator training
program is accredited by INPO and is based on a system[']s approach
to training.
The NRC has concluded, as stated in NUREG-1262, ``Answers to
Questions at Public Meetings Regarding Implementation of Title 10,
Code of Federal Regulations, Part 55 on Operators' Licenses,'' that
the standards and guidelines applied by INPO in their training
accreditation program are equivalent to those put forth or endorsed
by the NRC. As a result, maintaining an INPO accredited, systems
approach based licensed operator training program is equivalent to
maintaining an NRC approved licensed operator training program which
conform with applicable NRC Regulatory Guides or
[[Page 52806]]
NRC endorsed industry standards. The margin of safety is maintained
by virtue of maintaining an INPO accredited licensed operator
training program.
In addition, the NRC has recently published NRC Regulatory Issue
Summary 2001-01, ``Eligibility of Operator License Applicants,''
dated January 18, 2001, ``to familiarize addresses with the NRC's
current guidelines for the qualification and training of reactor
operator (RO) and senior operator (SO) license applicants.'' This
document again acknowledges that the INPO National Academy for
Nuclear Training (NANT) guidelines for education and experience,
outline acceptable methods for implementing the NRC's regulations in
this area.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: August 14, 2001, as supplemented on
August 21, 2001.
Description of amendment request: The proposed amendment would
extend the allowed outage time for the high pressure coolant injection
(HPCI) and reactor core isolation cooling systems from 7 days to 14
days. Requirements were added to immediately assure the availability of
alternate means of high pressure coolant makeup. Also clarifying
changes were made to Technical Specification (TS) 3.5.E.2 and TS
3.5.G.2 by reformatting the TSs to make nomenclature consistent
regarding HPCI and the automatic depressurization system (ADS) as being
systems not subsystems.
Date of publication of individual notice in Federal Register:
September 16, 2001 (66 FR 48152).
Expiration date of individual notice: October 18, 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. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of application for amendment: January 29, 2001, as
supplemented July 6, 2001.
Brief description of amendment: The amendment removes the note from
TMI-1 Technical Specification 4.5.4.1 which restricts the applicability
of the specified Engineered Safeguards Feature (ESF) Systems leakage
rate limit of 15 gallons per hour to the current operating cycle (Cycle
13). The amendment also approves full scope implementation of an
alternate source term for TMI-1 in accordance with Title 10 of the Code
of Federal Regulations (10 CFR) Section 50.67.
Date of issuance: September 19, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 235.
Facility Operating License No. DPR-50: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 12, 2001 (66 FR
31703).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 19, 2001.
No significant hazards consideration comments received: No.
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of application for amendment: January 23, 2001, as
supplemented August 22 and September 17, 2001.
Brief description of amendment: The amendment revised the TMI-1
Technical Specification requirements for containment integrity
associated with the personnel and emergency air locks during fuel
movement and refueling operations. Partial implementation of an
alternate source term (AST) in accordance with Regulatory Guide 1.183,
``Alternate Source Terms for Evaluating Design Basis Accidents at
Nuclear Power Plants,'' and Title 10 of the Code of Federal
Regulations, Section 50.67,
[[Page 52807]]
which the licensee had also requested in its application, was not
necessary because the Commission approved full implementation of an AST
for TMI-1 in Amendment No. 235 dated September 19, 2001.
Date of issuance: October 2, 2001.
Effective date: As of its date of issuance, contingent upon the
licensee's implementation of regulatory commitments contained in the
licensee's letters dated August 22 and September 17, 2001, and shall be
implemented within 30 days of issuance.
Amendment No.: 236.
Facility Operating License No. DPR-50: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 12, 2001 (66 FR
31702).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 2, 2001.
No significant hazards consideration comments received: No.
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of application for amendment: December 6, 2000, as
supplemented July 13 and September 6, 2001.
Brief description of amendment: The amendment revises the once-
through steam generator (OTSG) surveillance criteria contained in the
TMI-1 Technical Specifications (TSs) to allow OTSG tubes to remain in
service with indications of inside diameter intergranular attack
located below the upper tubesheet secondary face. The changes also
extend the repair criteria from a cycle-to-cycle basis to a permanent
basis.
Date of issuance: October 5, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 237.
Facility Operating License No. DPR-50: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 24, 2001 (66 FR
7669).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 5, 2001.
No significant hazards consideration comments received: No.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: June 15, 2001.
Brief description of amendments: The amendments delete Technical
Specifications Section 5.5.3, ``Post Accident Sampling System,'' for
Palo Verde Nuclear Generating Station, Units Nos. 1, 2 and 3, and
thereby eliminate the requirements to have and maintain the post-
accident sampling system.
Date of issuance: September 28, 2001.
Effective date: September 28, 2001, and shall be implemented within
7 months of the date of issuance.
Amendment Nos.: Unit 1-136, Unit 2-136, Unit 3-136.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 8, 2001 (66 FR
41611).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 28, 2001.
No significant hazards consideration comments received: No.
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 application for amendments: May 1, 2001, as supplemented
August 20, 2001.
Brief description of amendments: The amendments change the
Technical Specifications related to the pressure-temperature limit
curves.
Date of issuance: October 4, 2001.
Effective date: October 4, 2001.
Amendment Nos.: 214 and 241.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: May 30, 2001 (66 FR
29350). The August 20, 2001, supplement contained clarifying
information only, and did not change the initial no significant hazards
consideration determination, or expand the scope of the initial
application.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 4, 2001.
No significant hazards consideration comments received: No.
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 application for amendments: January 17, 2001, as
supplemented March 23 and August 31, 2001.
Brief description of amendments: The amendments change the
Technical Specifications to relax the 24-month surveillance frequency
of excess flow check valves (EFCVs) by limiting the number of tests to
a representative sample every 24 months such that each EFCV will be
tested at least once every 10 years.
Date of issuance: October 4, 2001.
Effective date: October 4, 2001.
Amendment Nos.: 215 and 242.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: February 21, 2001 (66
FR 11052). The March 23 and August 31, 2001, supplements contained
clarifying information only, and did not change the initial no
significant hazards consideration determination or expand the scope of
the initial application.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 4, 2001.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket No. 50-400, Shearon Harris
Nuclear Power Plant (HNP), Wake and Chatham Counties, North Carolina
Date of application for amendment: December 13, 2000, as
supplemented on February 9, and August 3, 2001.
Brief description of amendment: This amendment revises Technical
Specification (TS) 3/4.9.2, ``Refueling Operations--Instrumentation,''
and the associated Bases to permit using one Source Range Nuclear Flux
Monitor and one Wide Range Neutron Flux Monitor during MODE 6
(Refueling) instead of the two Source Range Nuclear Flux Monitors
specified in the current HNP TS.
Date of issuance: September 10, 2001.
Effective date: September 10, 2001.
Amendment No. 105.
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: January 24, 2001 (66 FR
7672).
The February 9, and August 3, 2001, submittals contained clarifying
information only, and did not change the initial no significant hazards
consideration determination or expand the scope of the initial
application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 10, 2001.
No significant hazards consideration comments received: No.
[[Page 52808]]
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: December 14, 2000, as
supplemented August 16, and September 12, 2001.
Brief description of amendment: The amendment revises Technical
Specification 3/4.8.1 related to Emergency Diesel Generators (EDGs),
and specifically revises Surveillance Requirement 4.8.1.1.2.f.7, the
24-hour EDG endurance run test, by removing the restriction to perform
the test during shutdown conditions.
Date of issuance: October 3, 2001.
Effective date: October 3, 2001.
Amendment No.: 106.
Facility Operating License No. NPF-63: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 24, 2001 (66 FR
7673). The August 16, and September 12, 2001 supplements contained
clarifying information that did not change the scope of the December
14, 2000, application nor the proposed initial no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 3, 2001.
No significant hazards consideration comments received: No.
Consumers Energy Company, Docket No. 50-155, Big Rock Point Plant,
Charlevoix, County, Michigan
Date of amendment request: October 26, 2000, as supplemented by
letters dated February 9, February 28, March 14, March 15, March 23,
May 2, July 13, July 17, and August 2, 2001.
Brief description of amendment: The amendment revises the Defueled
Technical Specifications to reflect the removal of the original 75-ton
Reactor Building gantry crane and its replacement with an upgraded
single-failure proof crane.
Date of issuance: September 28, 2001.
Effective date: The license amendment is effective as of its date
of issuance and shall be implemented within 60 days from the date of
issuance.
Amendment No.: 122.
Facility Operating License No. DPR-6: The amendment revised the
Defueled Technical Specifications.
Date of initial notice in Federal Register: May 2, 2001 (66 FR
22025). The supplemental letters dated May 2, July 13, July 17, and
August 2, 2001, provided additional clarifying information, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 28, 2001.
No significant hazards considerations comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: December 29, 2000, as
supplemented May 2 and July 19, 2001.
Brief description of amendment: The amendment revises the Fermi 2
Technical Specifications associated with handling irradiated fuel
assemblies, based on reevaluation of the design-basis fuel handling
accident analysis with an alternative radiological source term.
Date of issuance: September 28, 2001.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 144.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 7, 2001 (66 FR
9381). The application was renoticed on August 27, 2001 (66 FR 45062),
due to supplemental information beyond the scope of the initial notice.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated September 28, 2001.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket No. 50-414, Catawba Nuclear
Station, Unit 2, York County, South Carolina
Date of application for amendment: March 9, 2001, as supplemented
by letters dated July 25, September 10, and September 13, 2001.
Brief description of amendment: The amendment revised the cold leg
elbow tap flow coefficients used in the determination of Reactor
Coolant System flow rate. There are no changes to the associated
Technical Specifications with this amendment.
Date of issuance: October 2, 2001.
Effective date: As of its date of issuance and shall be implemented
before the startup of Cycle 12, and will be in effect only for the
duration of Cycle 12.
Amendment No.: 186.
Facility Operating License No. NPF-52: Amendment did not revise the
Technical Specifications.
Date of initial notice in Federal Register: June 27, 2001 (66 FR
34281). The supplements dated July 25, September 10, and September 13,
2001, provided clarifying information that did not change the scope of
the March 9, 2001, application nor the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 2, 2001.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: July 18, 2000, supplemented
August 22 and November 8, 2000, and June 7, July 26, and September 5,
2001.
Brief description of amendments: The amendments revised the
Technical Specifications to incorporate provisions of the Automatic
Feedwater Isolation System.
Date of Issuance: September 26, 2001.
Effective date: As of the date of issuance. It shall be implemented
for each Oconee unit prior to reactor startup following installation of
the system and training of appropriate personnel.
Amendment Nos.: 320, 320, and 320.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 20, 2000 (65
FR 56949). The supplements dated August 22 and November 8, 2000, and
June 7, July 26, and September 5, 2001, provided clarifying information
that did not change the scope of the July 18, 2000, application nor the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 26, 2001.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: January 24, 2001, as supplemented by
letter dated September 24, 2001.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to incorporate the provisions to perform routine
diesel generator (DG) monthly testing by gradually accelerating the DG
to
[[Page 52809]]
operating speed. In addition, a new TS was added to require fast starts
of the DGs on a 184-day frequency.
Date of issuance: September 27, 2001.
Effective date: As of the date of issuance and shall be implemented
120 days from the date of issuance.
Amendment No.: 121.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 7, 2001 (66 FR
13801). The supplemental letter dated September 27, 2001, provided
additional information that did not expand the scope of the NRC staff's
initial proposed no significant hazards consideration determination (66
FR 13801, published March 7, 2001).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 27, 2001.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: May 23, 2001, as supplemented by letters
dated July 23, 2001, and August 23, 2001.
Brief description of amendment: The amendment changes the following
Technical Specifications (TSs): (1) the value of the safety limit
minimum critical power ratio was changed in TS 2.1.1.2, (2) an
editorial clarification to TS 5.6.5.a.5) was added to include the
applicable reactor protection system instrumentation function, and (3)
the list of the approved methodologies in TS 5.6.5.b. and the
associated Bases and References were updated.
Date of issuance: October 3, 2001.
Effective date: As of the date of issuance and shall be implemented
30 days from the date of issuance.
Amendment No.: 122.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 27, 2001 (66 FR
34281). The supplemental letters dated July 23, 2001, and August 23,
2001, provided additional information that did not expand the scope of
the application or change the staff's initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 3, 2001.
No significant hazards consideration comments received: No.
Entergy Nuclear Generation Company, Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: May 31, 2001.
Brief description of amendment: The amendment revised Technical
Specification 5.5.6, ``Technical Specification (TS) Bases Control
Program,'' to provide consistency with the changes to 10 CFR 50.59
which were published in the Federal Register (64 FR 53582) on October
4, 1999.
Date of issuance: October 2, 2001.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 192.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 25, 2001 (66 FR
38761). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 2, 2001.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: September 7, 2000, as
supplemented December 29, 2000.
Brief description of amendment: The changes revise Technical
Specification Section 3.7.B.4 to allow a one-time replacement of
Station 125V DC batteries 31 and 32 while at power. The one-time change
is necessary to support an on-line replacement of the existing
batteries with new batteries. In addition, a change is made on a one-
time basis to conduct testing the battery while the plant is not
shutdown. Also included is an administrative change involving the
deletion of an expired one-time limiting condition for operation
statement related to an Emergency Diesel Generator Fuel Oil Storage
Tank repair effort.
Date of issuance: September 19, 2001.
Effective date: September 19, 2001.
Amendment No.: 208.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 15, 2000 (66
FR 15922).
The December 29, 2000, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 19, 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: July 31, 2001.
Brief description of amendment: The amendment revised and
transferred the inservice testing portion of Technical Specification
(TS) 4.0.5 to TS 6.5.8, and eliminated the inservice inspection portion
of TS 4.0.5. In addition, other sections of the TSs that reference TS
4.0.5 were revised to be consistent with the revisions discussed above.
Date of issuance: September 24, 2001.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 233.
Facility Operating License No. NPF-6: Amendment revised the TSs.
Date of initial notice in Federal Register: August 22, 2001 (66 FR
44167).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 24, 2001.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: May 22, 2001.
Brief description of amendment: The amendment changes Technical
Specifications (TS) 3/4.7.1.2, Emergency Feedwater System, and expands
and clarifies the current TS.
Date of issuance: October 4, 2001.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 173.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications and Surveillance Requirements.
Date of initial notice in Federal Register: June 27, 2001 (66 FR
34283).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 4, 2001.
No significant hazards consideration comments received: No.
[[Page 52810]]
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station (LGS), Units 1 and 2, Montgomery County,
Pennsylvania
Date of application for amendments: April 23, 2001.
Brief description of amendments: The amendments deleted the loose
parts monitoring system from the LGS Units 1 and 2 Technical
Specifications and Bases. The amendments were based on the conclusions
of the Boiling Water Reactor Owners' Group TopicalReport NEDC-32975P,
``Regulatory Relaxation for BWR Loose Parts Monitoring System,'' which
was approved by the Nuclear Regulatory Commission's Safety Evaluation
dated January 25, 2001.
Date of issuance: As of date of issuance and shall be implemented
within 30 days.
Effective date: September 19, 2001.
Amendment Nos.: 153 and 117.
Facility Operating License Nos. NPF-39 and NPF-85: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 8, 2001 (66 FR
41619).
The Commission's related evaluation of the amendments is contained
in a safety evaluation dated September 19, 2001.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, PSEG Nuclear LLC, and Atlantic City
Electric Company, Docket No. 50-278, Peach Bottom Atomic Power Station,
Unit 3, York County, Pennsylvania
Date of application for amendment: May 30, 2001 (two letters), as
supplemented July 24 (two letters), and August 13, 2001.
Brief description of amendment: This amendment revises Technical
Specification 5.5.12 to allow a one-time change in the containment
integrated leak rate test interval from the current 10 years to a test
interval of 15 years.
Date of issuance: October 4, 2001.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 244.
Facility Operating License No. DPR-56: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 11, 2001 (66 FR
36341). The July 24 (two letters), and August 13, 2001, letters
provided clarifying information that did not change the initial
proposed no significant hazards consideration determination or expand
the application beyond the scope of the original Federal Register
notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 4, 2001.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334,
Beaver Valley Power Station, Unit No. 1, (BVPS-1) Beaver County,
Pennsylvania
Date of application for amendment: March 28, 2001, as supplemented
by letters dated May 18, June 15, and July 18, 2001.
Brief description of amendment: The amendment approves changes to
the BVPS-1 Technical Specification boron concentration limits for the
refueling water storage tank, accumulators, boron injection tank (BIT),
and the reactor coolant system/refueling canal during Mode 6. In
conjunction with the reduction in the maximum boron concentration in
the BIT, the temperature controls on the BIT are eliminated.
Date of issuance: September 24, 2001.
Effective date: Immediately and to be implemented within 60 days.
Amendment No: 242.
Facility Operating License No. DPR-66: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 25, 2001 (66 FR
38763). The May 18, June 15, and July 18, 2001, letters provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination or expand the scope of
the initial Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 24, 2001.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334,
Beaver Valley Power Station, Unit No. 1, Beaver County, Pennsylvania
Date of application for amendment: March 28, 2001.
Brief description of amendment: These amendments approved
reductions in the reactor coolant system and secondary coolant system
specific activity limits specified in TS 3/4.4.8, ``Reactor Coolant
System Specific Activity,'' and TS 3/4.7.1.4, ``Plant Systems
Activity.'' These TS changes support revised safety analyses of the
design-basis main steam line break dose consequence analysis, which
assumes higher primary-to-secondary accident induced leakage in
accordance with the methodology described in Generic Letter 95-05,
``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes
by Outside Diameter Stress Corrosion Cracking.'' These amendments also
authorized Updated Final Safety Analysis Report (UFSAR) changes.
Date of issuance: September 28, 2001.
Effective date: As of date of issuance and shall be implemented
within 60 days.
Amendment No: 244.
Facility Operating License No. DPR-66: Amendment revised the
Technical Specifications and authorized changes to the UFSAR.
Date of initial notice in Federal Register: May 30, 2001 (66 FR
29354).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 28, 2001.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412,
Beaver Valley Power Station, Unit No. 2 (BVPS-2), Beaver County,
Pennsylvania
Date of application for amendment: January 18, 2001, as
supplemented by letters dated February 20, April 12, May 7, May 18,
June 9 (3 letters), June 26, June 29, August 21, and September 5, 2001.
Brief description of amendment: A portion of this amendment
approves revisions to BVPS-2 TS 3/4.4.9, ``Pressure/Temperature
Limits,'' heatup and cooldown curves.
Date of issuance: September 24, 2001.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 122.
Facility Operating License No. NPF-73: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 27, 2001 (66 FR
39211). The February 20, April 12, May 7, May 18, June 9 (3 letters),
June 26, June 29, August 21, and September 5, 2001, letters provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination or expand the amendment
beyond the scope of the original notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 24, 2001.
No significant hazards consideration comments received: No.
[[Page 52811]]
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of application for amendments: July 18, 2001, as supplemented
August 30 and September 6, 2001.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3.9.4 and its associated Bases to allow the
containment equipment door to be open during core alterations or
movement of non-recently irradiated fuel within the containment,
provided that the capability for closure is maintained.
Date of issuance: September 27, 2001.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos: 216 and 210.
Facility Operating License Nos. DPR-31 and DPR-41: Amendments
revised the TS.
Date of initial notice in Federal Register: August 8, 2001 (66 FR
41622). The August 30 and September 6, 2001, submittals provided
clarifying information that did not change the scope of the July 18,
2001, application and the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 27, 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: September 26, 2000, as
supplemented February 1, 2001, June 29, 2001, and August 10, 2001.
Brief description of amendments: The amendments would approve
changes to revise the current licensing basis, as stated in the updated
final safety analysis report, to require operator action to mitigate
the effects of a loss of seal injection cooling to the reactor coolant
pumps.
Date of issuance: September 28, 2001.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 255 and 238.
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: October 18, 2000 (65 FR
62386) The February 1, June 29, and August 10, 2001, supplemental
letters did not change the scope of the proposed action and did not
change the NRC's preliminary no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 28, 2001.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: April 30, 2001, as supplemented
June 27 and August 3, 2001.
Brief description of amendment: The amendment conforms the license
to reflect the transfer of Facility Operating License No. DPR-43 for
the Kewaunee Nuclear Power Plant (KNPP) to the extent held by Madison
Gas & Electric Company (MG&E) to Wisconsin Public Service Corporation
(WPSC), as approved by Order of the Commission dated September 20,
2001.
Date of issuance: September 27, 2001.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 159.
Facility Operating License No. DPR-43: Amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: July 27, 2001 (66 FR
39214). The August 3, 2001, supplement was within the scope of the
initial application as originally noticed. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
September 20, 2001.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendments: August 8, 2000.
Brief description of amendments: The amendments revised the
requirements for the containment isolation valves in the hydrogen/
oxygen analyzer containment penetrations. The related safety evaluation
also provided approval of an associated request to use closed system
boundary valves that do not completely meet the guidance described in
the Standard Review Plan, Section 6.2.4, ``Containment Isolation
System.''
Date of issuance: September 28, 2001.
Effective date: As of date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 195 & 170.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 12, 2001 (66 FR
31713). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 28, 2001.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: October 12, 2000, as
supplemented April 9, 2001.
Brief description of amendment: This amendment changes the
Technical Specifications (TSs) associated with the drywell vacuum
breakers and the suppression pool vacuum breakers to provide
consistency between the Hope Creek TSs and the improved standard TSs
(NUREG-1433).
Date of issuance: October 3, 2001.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 133.
Facility Operating License No. NPF-57: This amendment revised the
TSs.
Date of initial notice in Federal Register: November 29, 2000 (65
FR 71137). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 3, 2001.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: May 17, 2001, as supplemented on
August 6, August 17, and September 12, 2001
Brief description of amendment: The amendment revises the Technical
Specifications to permit an increase in the allowable leak rate for the
Main Steam Isolation Valves (MSIVs) and to delete the MSIV Sealing
System. These changes are based on the use of an alternate source term
and the guidance provided in Regulatory Guide 1.183, ``Alternate
Radiological Source Terms for Evaluating Design Basis Accidents at
Nuclear Power Reactors.''
Date of issuance: October 3, 2001.
Effective date: As of the date of issuance, and shall be
implemented during Refueling Outage 10, currently scheduled to commence
in October 2001.
Amendment No.: 134.
Facility Operating License No. NPF-57: This amendment revised the
TSs.
[[Page 52812]]
Date of initial notice in Federal Register: June 27, 2001 (66 FR
34288). The letters dated August 6, August 17, and September 12, 2001,
provided clarifying information that did not change the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 3, 2001.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: September 26, 2000, as
supplemented on October 6, 2000, and May 21, 2001.
Brief description of amendments: The amendments modify the Salem
Technical Specifications by increasing the as-found setpoint tolerance
for the Pressurizer Safety Valves from 1% to
3%; increasing the as-found setpoint tolerance for the Main
Steam Safety Valves (MSSV) from 1% to 3%;
changing the required actions for inoperable MSSVs; and removing
specifications and references related to plant operation with three
Reactor Coolant System loops.
Date of issuance: September 19, 2001
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment Nos.: 244 and 225
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 15, 2000 (65
FR 69065), as superseded on August 8, 2001 (66 FR 41624).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 19, 2001.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: May 31, 2000, as supplemented
on August 2, 2001.
Brief description of amendments: The amendments modify the Salem
Technical Specifications (TSs) Surveillance Requirements for: (1) The
Control Room Envelope Air Conditioning System (CREACS), (2) the
Auxiliary Building Ventilation System (ABVS), and (3) the Fuel Handling
Building Ventilation System (FHVS). Salem TSs will now require the use
of American Society for Testing and Materials (ASTM) D3803-1989,
``Standard Test Method for Nuclear-Grade Activated Carbon,'' as the
test protocol to evaluate charcoal samples from the ABVS, CREACS, and
FHVS.
Date of issuance: September 19, 2001.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment Nos.: 245 and 226.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 26, 2000 (65 FR
46014). The August 2, 2001, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 19, 2001.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: June 29, 2001 as supplemented
by letter dated August 20, 2001.
Brief description of amendments: Revise Technical Specifications
(TSs) 3.7.10, ``Emergency Chilled Water (ECW)'' and 3.7.11, ``Control
Room Emergency Air Cleanup System (CREACUS)'' and the associated TSs
Bases. The proposed change would revise the Allowed Outage Time for a
single inoperable train of both the ECW and CREACUS from 7 days to 14
days.
Date of Issuance: October 4, 2001.
Effective date: October 4, 2001, to be implemented within 30 days
of issuance
Amendment Nos.: Unit 2-181; Unit 3-172
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 22, 2001 (66 FR
44175). The August 20, 2001 supplemental letter provided additional
clarifying information, did not expand the scope of the proposed
amendment as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 4, 2001.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: June 29, 2000, as supplemented August
31, 2001.
Brief description of amendments: The amendments revise design bases
in the Final Safety Analysis Report. The change adds a description of
the methodology Southern Nuclear Operating Company uses to determine
what systems and components need to be protected from tornado missiles.
Date of issuance: September 26, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 150 and 142.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: August 9, 2000 (65 FR
48758).
The supplement dated August 31, 2001, provided clarifying
information that did not change the scope of the June 29, 2000,
application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 26, 2001.
No significant hazards consideration comments received: No.
TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: October 4, 2000, as supplemented by
letters dated April 30, June 18, and July 18, 2001.
Brief description of amendments: The amendments revise the
Technical Specifications to increase the spent fuel storage capacity
from 2,026 to 3,373 fuel assemblies in the spent fuel pool.
Date of issuance: October 2, 2001.
Effective date: As of the date of issuance and shall be implemented
no later than January 31, 2002.
Amendment Nos.: 87/87.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 4, 2000 (65 FR
75737).
The April 30, June 18, and July 18, 2001, supplemental letters
provided clarifying information that was within the scope of the
original Federal Register notice and did not change the staff's initial
no significant hazards consideration determination.
[[Page 52813]]
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 2, 2001.
No significant hazards consideration comments received: No.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: April 17, 2001.
Brief description of amendment: The amendment removes unnecessary
details for certain secondary post-accident monitoring instrumentation
from Technical Specification Table 3.2.6.
Date of Issuance: October 2, 2001.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 204.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 16, 2001 (66 FR
27178).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated October 2, 2001.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: March 23, 2001 (CO 01-0013).
Brief description of amendment: The amendment deletes (1) certain
license conditions from Facility Operating License No. NPF-42, and (2)
reporting requirements in Table 5.5.9-2, ``Steam Generator Tube
Inspection,'' in Section 5.5.9, ``Steam Generator (SG) Tube
Surveillance Program,'' of the technical specifications. License
Conditions 2.C.(4), and 2.C.(6) through 2.C.(14), Section 2.F, and
Attachments 2 and 3 to Facility Operating License No. NPF-42 are
deleted, and the list of the attachments and appendices to Facility
Operating License No. NPF-42 is revised to reflect the deletion of the
attachments. The reporting requirements deleted in Table 5.5.9-2
duplicate requirements in 10 CFR 50.72.
Date of Issuance: September 24, 2001.
Effective date: September 24, 2001, and shall be implemented within
60 days from the date of issuance.
Amendment No.: 141.
Facility Operating License No. NPF-42: The amendment revised the
operating license and the Technical Specifications.
Date of initial notice in Federal Register: May 2, 2001 (66 FR
22035).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 24, 2001
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: September 15, 2000, and supplements
dated October 3, 2000, and September 13, 2001.
Brief description of amendment: The amendment revises footnotes (b)
and (c) of Table 1.1-1, ``Modes,'' and adds a program plan to Section
5.5, ``Programs and Manuals,'' of the Wolf Creek Generating Station
Technical Specifications. The amendment will allow the plant to operate
at full power with one closure bolt less than fully tensioned for one
operating cycle.
Date of issuance: September 27, 2001.
Effective date: September 27, 2001, to be implemented within 60
days from the date of issuance.
Amendment No.: 142.
Facility Operating License No. NPF-42: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 4, 2000 (65 FR
59227).
The supplements dated October 3, 2000, and September 13, 2001,
provided additional clarifying information, did not expand the scope of
the application as originally noticed, and did not change the staff's
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 27, 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 10th of October 2001.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 01-25957 Filed 10-16-01; 8:45 am]
BILLING CODE 7590-01-P