[Federal Register Volume 68, Number 227 (Tuesday, November 25, 2003)]
[Notices]
[Pages 66131-66144]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-29107]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from, October 31, through November 13, 2003. The
last biweekly notice was published on November 12, 2003 (68 FR 64133).
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.
[[Page 66132]]
However, should circumstances change during the notice period such that
failure to act in a timely way would result, for example, in derating
or shutdown of the facility, the Commission may issue the license
amendment before the expiration of the 30-day notice period, provided
that its final determination is that the amendment involves no
significant hazards consideration. The final determination will
consider all public and State comments received before action is taken.
Should the Commission take this action, it will publish in the Federal
Register a notice of issuance and provide for opportunity for a hearing
after issuance. The Commission expects that the need to take this
action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By December 26, 2003, 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 Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records will be accessible from the
Agencywide Documents Access and Management System's (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing
or petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of continuing disruptions in delivery of mail to United States
Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the request for
hearing and petition for leave to intervene should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and because of continuing disruptions in
delivery of mail to United States Government offices, it is requested
that copies be transmitted either by means of facsimile transmission to
301-415-3725 or by e-mail to [email protected]. A copy of the
request for hearing and petition for leave to intervene should also be
sent to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions,
[[Page 66133]]
supplemental petitions and/or requests for a hearing will not be
entertained absent a determination by the Commission, the presiding
officer or the Atomic Safety and Licensing Board that the petition and/
or request should be granted based upon a balancing of factors
specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to
[email protected].
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and
50-318, Calvert Cliffs.
Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland.
Date of amendments request: October 14, 2003.
Description of amendments request: The proposed amendment would
change the frequency of surveillance testing for some engineered safety
features (ESF) components.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
Integrated testing of the ESF trains takes place while the unit
is shut down. The equipment being tested is normally used to respond
to an accident when the Unit is in Modes 1, 2, or 3. Changing the
test Frequency to a longer period does not affect the scope of the
testing or the methods used during the testing. Therefore, there is
no increase in the probability of an accident previously evaluated
caused by the testing itself.
The components tested during the integrated ESF test are
components needed to mitigate the consequences of an accident.
Increasing the length of time between integrated tests increases the
likelihood of undetected equipment failure. This creates a change in
plant risk. This change in risk is analyzed and quantified using
probabilistic risk assessment techniques. The risk analysis provides
results that show the proposed increase in ESF component
surveillance testing Frequency meets the guidance of Regulatory
Guide 1.174, ``An Approach for Using Probabilistic Risk Assessment
in Risk-Informed Decisions on Plant-Specific Changes to the
Licensing Basis.'' The increase in risk is within the guidelines of
the regulatory guidance. There is no significant change in the
probability that the equipment will suffer an undetected failure in
the increased time between Surveillance tests. Therefore, there is
no significant increase in the consequences o[f] an accident
previously evaluated.
An additional change is proposed to delete a Surveillance
Requirement because the signal tested in the Surveillance
Requirement is no longer installed in the plant. This deletion has
no impact on plant operations or the response of the plant in an
accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Would not create the possibility of a new or different kind
of accident from any accident previously evaluated.
The proposed change would extend the Surveillance Frequency of
the integrated ESF test. This change does not affect the scope of
the testing or the methods used during the testing. Plant equipment
will continue to operate as designed. Only the testing frequency is
changed. Because there are no changes in the scope or method of
testing and this proposed change does not affect the operation of
the equipment in other circumstances, no new accident initiators
have been introduced.
An additional change is proposed to delete a Surveillance
Requirement because the signal tested in the Surveillance
Requirement is no longer installed in the plant. This deletion has
no impact on plant operations or the response of the plant and
therefore would not create the possibility of a new or different
kind of accident from any previously evaluated.
Therefore, this proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Would not involve a significant reduction in [a] margin of
safety.
Surveillance testing is performed to evaluate the operability of
equipment used to perform safety functions at the Unit. The
components tested during the integrated ESF test are components
needed to mitigate the consequences of an accident. Increasing the
length of time between integrated tests increases the likelihood of
undetected equipment failure. This creates a change in plant risk.
This change in risk is analyzed and quantified using probabilistic
risk assessment techniques. The risk analysis provides results that
show the proposed increase in ESF component surveillance testing
Frequency meets the guidance of Regulatory Guide 1.174. The increase
in risk is within the guidelines of the regulatory guidance. There
is no significant change in the probability that the equipment will
suffer an undetected failure in the increased time between
Surveillance tests. Since the function of Surveillance testing is to
evaluate the operability of equipment, and the increased time
between Surveillance tests has been evaluated and found to be
acceptable under regulatory guidance, the proposed change would not
involve a significant reduction in [a] margin of safety.
An additional change is proposed to delete a Surveillance
Requirement because the signal tested in the Surveillance
Requirement is no longer installed in the plant. This deletion has
no impact on plant operations or the response of the plant in an
accident and does not impact the margin of safety.
Therefore, this proposed change does not significantly reduce [a]
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Attorney for licensee: James M. Petro, Jr., Esquire, Counsel,
Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor,
Baltimore, MD 21202.
NRC Section Chief: Richard J. Laufer.
Consumers Energy Company, Docket No. 50-155, Big Rock Point Nuclear
Plant, Charlevoix County, Michigan.
Date of amendment requests: August 6, 2003.
Description of amendment requests: The Big Rock Point Plant is in
the 6th year of decommissioning. The reactor was defueled and certified
as permanently shutdown by letter to the Nuclear Regulatory Commission
dated September 22, 1997. As of March 26, 2003, all the spent fuel has
been permanently removed from the plant's spent fuel pool and located
to an Independent Spent Fuel Storage Installation (ISFSI). The spent
fuel has been loaded into an NRC approved and licensed Spent Fuel Dry
Storage System and will be temporarily stored at this installation
until such time that a permanent repository is available. The
requirements associated with the wet storage of the spent fuel as
described in Defueled Technical Specifications are no longer applicable
and are being revised.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
[[Page 66134]]
No. The proposed change is an administrative change to update
the facility's Operating License and Defueled Technical
Specifications to reflect the permanent removal of the spent fuel
from the Spent Fuel Pool. Requirements for safe storage and handling
of irradiated fuel, definitions, design features and administrative
controls that were applicable to the facility when spent fuel was
stored in the spent fuel pool are no longer valid and are being
removed to provide clarity to the licensing basis of the facility in
its current configuration. The accidents previously evaluated in the
Updated Final Hazards Safety Analysis are based on spent nuclear
fuel being stored in the spent fuel pool. Since the spent fuel has
been permanently removed from the spent fuel pool, the accidents
previously analyzed are no longer credible. The spent fuel has been
loaded into an NRC approved and licensed Spent Fuel Dry Storage
System and will be temporarily stored at this installation until
such time that a permanent repository is available. The spent fuel
is now controlled by a different set of approved technical
specifications issued and approved pursuant to 10 CFR part 72.
Therefore, the proposed administrative change does not affect the
consequences of any accident described and evaluated in the Updated
Final Hazards Summary Report, and the accidents and transients
associated with spent fuel stored in the facility's spent fuel pool
are no longer applicable.
Therefore, the proposed administrative change to the Operating
License and Defueled Technical Specifications does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Will the proposed change create the possibility of a new or
different type of accident from any accident previously evaluated?
No. The spent fuel has been loaded into an NRC approved and
licensed Spent Fuel Dry Storage System and will be temporarily
stored at this installation until such time that a permanent
repository is available. In accordance with 10 CFR part 72,
``Licensing Requirements for the Independent Storage of Spent
Nuclear Fuel and High-Level Radioactive Waste,'' credible accidents
have been evaluated as part of the licensing and approval process
for the Dry Fuel Storage System. The requirement to evaluate
credible accidents has not changed.
Therefore this proposed administrative change does not create
the possibility of a new or different kind of accident previously
evaluated.
3. Will the proposed change involve a significant reduction in a
margin of safety?
The proposed activity is an administrative change to the
Operating License and Defueled Technical Specifications to reflect
the permanent removal of the spent fuel from the spent fuel pool and
does not involve any significant reduction in any margin of safety
that is usually associated with the design and performance of
systems, structures and components. Requirements for safe storage
and handling of irradiated fuel, definitions, design features and
administrative controls that were applicable to the facility when
spent fuel was stored in the spent fuel pool are no longer
applicable and are being removed to provide clarity to the licensing
basis of the facility in its current configuration.
Therefore, the proposed administrative 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: David A. Mikelonis, Esquire, Consumers
Energy Company, One Energy Plaza, Jackson, MI 49201-2276.
NRC Section Chief: Claudia Craig.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan.
Date of amendment request: October 10, 2003.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) 3.7.3, ``Control Room Emergency
Filtration (CREF) System,'' Surveillance Requirement (SR) 3.7.3.6, to
permit a one-time extension of SR 3.7.3.6 until startup from the next
refueling outage (RF-10) to preclude a mid-cycle shutdown solely for
the performance of this SR. SR 3.7.3.6 requires verifying that
unfiltered inleakage from CREF system duct work outside the control
room envelope that is at negative pressure during accident conditions
is within limits. This SR is required to be performed every 36 months,
and can be performed only when the CREF system is not required to be
Operable (i.e., in MODES 4 or 5, with no operations with a potential
for draining the reactor vessel and with no fuel movement of recently
irradiated fuel in progress).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change allows a one-time extension of SR 3.7.3.6
until startup from the next refueling outage (approximately 10 to 12
months beyond its critical completion date). The Control Room
Emergency Filtration (CREF) system provides a configuration for
mitigating radiological consequences of accidents; however, it is
not considered an initiator of any previously analyzed accident.
Therefore, the proposed change cannot increase the probability of
any previously evaluated accident.
The CREF system provides a radiologically controlled environment
from which the plant can be safely operated following a radiological
accident. The current TS surveillance (SR 3.7.3.6) measures
inleakage from four sections of CREF system duct work outside the
Control Room Envelope (CRE) that are at negative pressure during
accident conditions. Based on the results of previous surveillance
testing, and the continued performance of SR 3.7.3.3 and 3.7.3.5 on
their normal schedule, the delay in performing SR 3.7.3.6 by
approximately 10 to 12 months will provide essentially the same
degree of assurance that CRE integrity is being maintained as
before. It is expected that CRE integrity will remain essentially
unchanged from what it is today. Therefore, the proposed change does
not significantly increase the radiological consequences of any
previously analyzed accident.
Based on the above, the proposed change does not significantly
increase the probability or consequences of any accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change to allow a one-time extension of SR 3.7.3.6
until startup from the next refueling outage (approximately 10 to 12
months beyond its critical completion date) does not alter the
design or function of the system involved, nor does it introduce any
new modes of plant or CREF system operation. Therefore, the proposed
change does not create the potential for a new or different kind of
accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The proposed change to allow a one-time extension of SR 3.7.3.6
until startup from the next refueling outage (approximately 10 to 12
months beyond its critical completion date) will not affect the
radiological release from a design basis accident. Based on the
results of previous surveillance testing and the continued
performance of SR 3.7.3.3 and 3.7.3.5 on their normal schedule, the
delay in performing SR 3.7.3.6 by approximately 10 to 12 months will
provide essentially the same degree of assurance that CRE integrity
is being maintained as existed before; and, the postulated dose to
the control room occupants as a result of an accident will remain
approximately the same. Therefore, the proposed changes will not
result in 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: Peter Marquardt, Legal Department, 688 WCB,
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
[[Page 66135]]
NRC Section Chief: L. Raghavan.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi.
Date of amendment request: October 24, 2003.
Description of amendment request: The proposed amendment would
revise Technical Specification 3.1.8, ``Scram Discharge Volume (SDV)
Vent and Drain Valves,'' to allow a vent or drain line with one
inoperable valve to be isolated instead of requiring the valve to be
restored to Operable status within 7 days.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on February 24, 2003 (68 FR 8637), on possible
amendments to revise the action for one or more SDV vent or drain lines
with an inoperable valve, including a model safety evaluation and model
no significant hazards consideration (NSHC) determination, using the
consolidated line-item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on April 15,
2003 (68 FR 18294). The licensee affirmed the applicability of the
model NSHC determination in its application dated October 24, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
A change is proposed to allow the affected SDV vent and drain
line to be isolated when there are one or more SDV vent or drain
lines with one valve inoperable instead o[f] requiring the valve to
be restored to operable status within 7 days. With one SDV vent or
drain valve inoperable in one or more lines, the isolation function
would be maintained since the redundant valve in the affected line
would perform its safety function of isolating the SDV. Following
the completion of the required action, the isolation function is
fulfilled since the associated line is isolated. The ability to vent
and drain the SDVs is maintained and controlled through
administrative controls. This requirement assures the reactor
protection system is not adversely affected by the inoperable
valves. With the safety functions of the valves being maintained,
the probability or consequences of an accident previously evaluated
are not significantly increased.
Criterion 2--The proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. Thus,
this change does not create the possibility of a new or different
kind of accident from any previously evaluated.
Criterion 3--The proposed change does not involve a significant
reduction in the margin of safety.
The proposed change ensures that the safety functions of the SDV
vent and drain valves are fulfilled. The isolation function is
maintained by redundant valves and by the required action to isolate
the affected line. The ability to vent and drain the SDVs is
maintained through administrative controls. In addition, the reactor
protection system will prevent filling of an SDV to the point that
it has insufficient volume to accept a full scram. Maintaining the
safety functions related to isolation of the SDV and insertion of
control rods ensures that the proposed change does not involve a
significant reduction in the margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont.
Date of amendment request: July 31, 2003, as supplemented on
October 10, 2003.
Description of amendment request: This amendment request
incorporates a revision to the licensing basis of the Vermont Yankee
Nuclear Power Station (VYNPS) that supports a full scope application on
an Alternative Source Term (AST) methodology.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR) Section 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration which is
presented below:
1. Will the proposed changes involve a significant increase in
the probability or consequences of an accident previously evaluated?
Adoption of the AST and those plant systems affected by
implementation of the AST do not initiate DBAs [design basis
accidents]. The proposed change does not affect the design or manner
in which the facility is operated; rather, once the occurrence of an
accident has been postulated, the new accident source term is an
input to analyses that evaluate the radiological consequences.
Therefore, the proposed change does not involve an increase in the
probability of an accident previously evaluated.
The structures, systems and components (SSCs) affected by the
proposed change act as mitigators to the consequences of accidents.
Based on the revised analyses, the proposed changes do revise
certain performance requirements; however, the proposed changes do
not involve a revision to the parameters or conditions that could
contribute to the initiation of a design basis accident discussed in
Chapter 14 of the Updated Final Safety Analysis Report.
Because of the changed methodology, it is difficult to draw a
quantitative comparison of before and after accident consequences
due to the use of different dose calculations, conversion factors,
source term, and other assumptions. However qualitatively, it can be
shown that there is no significant increase in offsite doses,
although there may be small variations in potential doses for
postulated accidents. Plant-specific radiological analyses have been
performed using the AST methodology. Based on the results of these
analyses, it has been demonstrated that the dose consequences of the
limiting events considered in the analyses meet the regulatory
guidance provided for use with the AST, and the offsite doses are
well within acceptable limits. This guidance is presented in 10 CFR
50.67, Regulatory Guide 1.183, and Standard Review Plan (SRP)
Section 15.0.1.
Therefore, the proposed amendment does not result in a
significant increase in the consequences or increase the probability
of any previously evaluated accident.
2. Will the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
Implementation of AST and the proposed changes does not alter or
involve any design basis accident initiators. These changes do not
affect the design function or mode of operations of SSCs in the
facility prior to a postulated accident. Since SSCs are operated
essentially no differently after the AST implementation, no new
failure modes are created by this proposed change.
Therefore, the proposed license amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Will the proposed changes involve a significant reduction in
a margin of safety?
The changes proposed are associated with a revision to the
licensing basis for the VYNPS. Approval of the licensing basis
change from the original source term to the alternative source term
is requested by this application for a license amendment. The
results of the accident analyses revised in support of the proposed
change are subject to the acceptance criteria in 10 CFR 50.67. The
analyzed events have been carefully selected, and the analyses
supporting these changes have been performed using approved
methodologies to ensure that analyzed events are bounding and safety
margin has not been reduced. The dose consequences of these limiting
events are within the acceptance criteria presented in 10 CFR 50.67,
Regulatory Guide 1.183, and SRP 15.0.1. Thus, by meeting the
applicable regulatory
[[Page 66136]]
limits for AST, there is no significant reduction in a margin of
safety.
Therefore, because the proposed changes continue to result in
dose consequences within the applicable regulatory limits, the
changes are considered to not result in 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. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334
and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and
2), Beaver County, Pennsylvania.
Date of amendment request: October 17, 2003.
Description of amendment request: The proposed amendments revise
the action requirements of Technical Specification (TS) 3/4 6.3,
``Containment Isolation Valves [CIVs],'' to more clearly define action
requirements for inoperable CIVs. The proposed changes to the action
requirements also include: (1) Provisions for allowing the intermittent
unisolation of penetration flow paths which have been isolated per
action requirements under administrative control; (2) use of check
valves as an isolation device; and (3) an increase in the allowed
outage time to 72 hours for CIVs associated with closed systems inside
containment. The proposed amendments also revise the TS surveillance
requirements (SRs) for CIVs by replacing existing SRs with new SRs
similar to those in NUREG-1431, Revision 2, ``Standard Technical
Specifications for Westinghouse Plants.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not involve any changes to plant
equipment, system design functions or a change in the methods
governing normal plant operation. Therefore, the probability of a
malfunction of a structure, system or component to perform its
design function will not be increased.
The proposed change modifies existing action requirements for
inoperable containment isolation valves. Action requirements and
their associated allowed outage times are not initiating conditions
for any accident previously evaluated and the accident analyses do
not assume that repaired equipment is out of service prior to the
analyzed event. In addition, changes that are consistent with the
ISTS [improved Standard Technical Specifications] have been
previously evaluated and found not to adversely affect the safe
operation of Westinghouse plants or the initiation of any accident
previously evaluated. Based on the conclusions of the plant specific
evaluation associated with the changes and the evaluation performed
in developing the ISTS, the proposed revised action requirements do
not result in operating conditions that will significantly increase
the probability of initiating an analyzed event. The revised action
requirements provide appropriate remedial actions to be taken in
response to the degraded condition considering the operability
status of the redundant systems of required features, and the
capability of remaining features while minimizing the risk
associated with continued operation. As a result, the consequences
of any accident previously evaluated are not significantly
increased.
The proposed change also modifies and deletes some surveillance
requirements. Surveillances are not initiators to any accident
previously evaluated. Consequently, the probability of an accident
previously evaluated is not significantly increased. The equipment
specified in the Limiting Condition for Operation is still required
to be operable and capable of performing the accident mitigation
functions assumed in the accident analysis. This equipment will
continue to be tested in a manner and at a frequency to give
confidence that the equipment can perform its assumed safety
function. The proposed changes are generally made to conform to the
ISTS and have been evaluated to not be detrimental to plant safety.
As a result, the proposed surveillance requirement changes do not
significantly affect the consequences of any 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 proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve any changes to plant
equipment, system design functions or a change in the methods
governing normal plant operation. The [technical] specification for
containment isolation valves provide[s] controls for maintaining the
containment pressure boundary. The revised action requirements and
revised surveillance requirements are sufficient to ensure the
containment isolation valves are capable of performing their
accident mitigation functions. No new accident initiators are
introduced by these changes. Therefore, the proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The revised action requirements do not involve a significant
reduction in the margin of safety. The proposed actions for
inoperable containment isolation valves minimize the risk of
continued operation under the specified conditions, considering the
operability status of the redundant containment isolation barriers,
a reasonable time for repairs or replacement of the isolation
feature, and the low probability of a design basis accident
occurring during the repair period.
The revised surveillance requirements do not involve a
significant reduction in the margin of safety. The proposed
surveillance requirements provide the required verifications for
ensuring containment isolation valves operability. Containment
isolation valve testing will continue to be performed in a manner
and at a frequency necessary to give confidence that the equipment
can perform its assumed safety function.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-
Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio.
Date of amendment request: December 17, 2001, as supplemented by
letter dated June 4, 2002.
Description of amendment request: The proposed amendment would
change Technical Specification (TS) Section 3/4.3.1, ``Reactor Coolant
System Instrumentation,'' to delete an action involving either reducing
core thermal power and the high neutron flux reactor trip setpoint or
monitoring quadrant power tilt when a reactor protection system (RPS)
channel is inoperable. Additionally, changes to the content and format
of TS Tables 3.3-1 and 4.3-1 are proposed to enhance specification
clarity.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided their analysis of
[[Page 66137]]
the issue of no significant hazards consideration. The staff has
reviewed the licensee's analysis against the standards of 10 CFR
50.92(c). The NRC staff's review is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change does not result in an increase in the
probability of an accident previously evaluated because no change is
being made to any accident initiator. The proposed change does not
result in an increase in the consequences of an accident previously
evaluated because TS 3/4.2.4, ``Quadrant Power Tilt,'' continues to
ensure the radial power distribution of the core is within the
limits assumed in the accident analyses. In addition, compensatory
actions will continue to be required should a single channel of RPS
High Flux or Flux-'Flux-Flow become inoperable. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes affect the TS requirements for the RPS
instrumentation. The proposed changes do not change the RPS design
function or result in the RPS being operated outside its design
operating range. There are no new or different equipment failure
modes introduced by the proposed changes. The proposed changes do
not introduce any new or different accident initiators. Therefore,
the proposed changes do not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed changes affect the TS requirements for the RPS
instrumentation. The capability of the RPS to perform its required
functions is not adversely affected by the proposed changes. The
proposed changes do not alter any initial conditions contributing to
accident severity or consequences. There will be no changes to the
plants' systems, structures, or components, nor in the manner in
which they will be operated as a result of the proposed changes.
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: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine.
Date of amendment request: September 11, 2003.
Description of amendment request: Revise the dose model for the
containment activated concrete, rebar (hereafter referred to as
activated concrete) and liner, by incorporating more realistic
radionuclide release rates and to change the associated derived
concentration guideline limit (DCGL) for activated concrete.
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 amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The requested license amendment does not authorize any plant
activities beyond those allowed by 10 CFR Chapter I or beyond those
considered in the DSAR. The bounding accident described in the
Defueled Safety Analysis Report (DSAR) for potential airborne
activity is the postulated resin cask drop accident in the Low Level
Radioactive Waste Storage Building. This accident is expected to
contain more potential airborne activity than can be released from
other decommissioning events. The radionuclide distribution assumed
for the spent resin cask has a greater inventory of transuranic
radionuclides (the major dose contributor) than the distribution of
plant derived radionuclides in the components involved in other
decommissioning accidents. The other accidents considered in the
DSAR include: (1) Explosion of liquid petroleum gas (LPG) leaked
from a front end loader or forklift; (2) Explosion of oxyacetylene
during segmenting of the reactor vessel shell; (3) Release of
radioactivity from the RCS decontamination ion exchange resins; (4)
Gross leak during in-situ decontamination; (5) Segmentation of RCS
piping with unremoved contamination; (6) Fire involving contaminated
clothing or combustible waste; (7) Loss of local airborne
contamination control during blasting or jackhammer operations; (8)
Temporary Loss of Services; (9) Dropping of Contaminated Concrete
Rubble; (10) Natural phenomena; and (11) Transportation accidents.
The probabilities and consequences for these accidents are estimated
in the basis documentation for DSAR Section 7. No systems,
structures, or components that could initiate or be required to
mitigate the consequences of an accident are affected by the
proposed change in any way not previously evaluated in the DSAR.
Since Maine Yankee does not exceed the salient parameters associated
with the plant referenced in the basis documentation in any material
respects, it is concluded that these probabilities and consequences
are not increased. Therefore, the proposed change to the Maine
Yankee license does not involve any increase in the probability or
consequences of any accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The requested license amendment does not authorize any plant
activities that could precipitate or result in any accidents beyond
those considered in the DSAR. The accidents previously evaluated in
the DSAR are described above. These accidents are described in the
basis documentation for DSAR Section 7. The proposed change does not
affect plant systems, structures, or components in any way not
previously evaluated in the DSAR. Since Maine Yankee does not exceed
the salient parameters associated with the plant referenced in the
basis documentation in any material respects, it is concluded that
these accidents appropriately bound the kinds of accidents possible
during decommissioning. Therefore, the proposed change to the Maine
Yankee license would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The margin of safety defined in Maine Yankee's license basis for
the consequences of decommissioning accidents has been established
as the margin between the bounding decommissioning accident and the
dose limits associated with the need for emergency plan offsite
protection, namely the Environmental Protection Agency Protective
Action Guidelines EPA-PAGs. As described above, the bounding
decommissioning accident is the postulated resin cask drop accident
in the Low Level Radioactive Waste Storage Building. Since the
bounding decommissioning accident is expected to contain more
potential airborne activity than can be released from other
decommissioning events and since the radionuclide distribution
assumed for the spent resin cask has more transuranics (the major
dose contributor) than the distribution in the components involved
in other decommissioning accidents, the margin of safety associated
with the consequences of decommissioning accidents cannot be
reduced. The margin of safety defined in the statements of
consideration for the final rule on the Radiological Criteria for
License Termination is described as the margin between the 100 mrem/
yr public dose limit established in 10 CFR 20.1301 for licensed
operation and the 25 mrem/yr dose limit to the average member of the
critical group at a site considered acceptable for unrestricted use.
This margin of safety accounts for the potential effect of multiple
sources of radiation exposure to the critical group. Since the
license termination plan (LTP) was designed to comply with the
radiological criteria for license termination for unrestricted use,
the margin of safety cannot be reduced. Therefore, the proposed
changes to the Maine Yankee license would not involve a significant
reduction in any margin of safety.
[[Page 66138]]
Conclusion
Based on the above, Maine Yankee concludes that the proposed
amendment presents no significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendment involves no significant hazards consideration.
Attorney for licensee: Joe Fay, Esquire, Maine Yankee Atomic Power
Company, 321 Old Ferry Road, Wiscasset, Maine 04578.
NRC Section Chief: Claudia M. Craig.
Nuclear Management Company, LLC (NMC), Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant (PBNP), Units 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of amendment request: September 26, 2003.
Description of amendment request: The proposed amendments would
modify TS 5.6.5.b to add a reference to a Nuclear Regulatory Commission
(NRC) letter that would approve the use of a new master curve
methodology for Unit 2. The NRC staff is currently reviewing an
associated exemption request by NMC to use this new methodology. The
requested exemption would allow the use of the master curve methodology
described in Babcock & Wilcox Report BAW-2308, Revision 1, ``Initial
RTNDT [reference nil-ductility temperature] of Linde 80 Weld
Materials,'' for determining the adjusted RTNDT of the Unit
2 reactor vessel limiting circumferential weld metal. This method is
used for the pressurized thermal shock screening evaluation. The
proposed amendments would also make editorial changes to TS 5.6.5.b.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Operation of PBNP in accordance with the proposed amendments
does not result in a significant increase in the probability or
consequences of any accident previously evaluated.
The proposed change references the NRC safety evaluation
[currently under NRC staff review] accepting the new Master Curve
Methodology used in the evaluation of the revised P/T [pressure/
temperature] limits and LTOP [low-temperature overpressure
protection] setpoints. Implementation of revisions to Topical
Reports would still be reviewed in accordance with 10 CFR 50.59 and,
where required, receive NRC review and approval. The proposed change
does not adversely affect accident initiators or precursors nor
alter the design assumptions, conditions, or configuration of the
facility or the manner in which the plant is operated and
maintained. The proposed change does not alter or prevent the
ability of structures, systems, and components (SSCs) from
performing their intended function to mitigate the consequences of
an initiating event within the assumed acceptance limits. The
proposed change does not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of an accident previously evaluated.
Further, the proposed change does not increase the types or amounts
of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures. The proposed change is consistent with safety
analysis assumptions and resultant consequences. Therefore, it is
concluded that this change does not increase the probability of
occurrence of an accident previously evaluated.
2. Operation of PBNP in accordance with the proposed amendments
does not result in a new or different kind of accident from any
accident previously evaluated.
The proposed change references the NRC safety evaluation
[currently under NRC staff review] accepting the new Master Curve
Methodology used in the evaluation of the revised P/T limits and
LTOP setpoints. Implementation of revisions to Topical Reports would
still be reviewed in accordance with 10 CFR 50.59 and, where
required, receive NRC review and approval. The change does not
involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operation. In addition, the changes
do not impose any new or different requirements or eliminate any
existing requirements. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Operation of PBNP in accordance with the proposed amendments
does not result in a significant reduction in a margin of safety.
The proposed change references the NRC safety evaluation
[currently under NRC staff review] accepting the new Master Curve
Methodology used in the evaluation of the revised P/T limits and
LTOP setpoints. Implementation of revisions to Topical Reports would
still be reviewed in accordance with 10 CFR 50.59 and, where
required, receive NRC review and approval. The proposed change does
not alter the manner in which safety limits, limiting safety system
settings or limiting conditions for operation are determined. The
setpoints at which protective actions are initiated are not altered
by the proposed changes. Sufficient equipment remains available to
actuate upon demand for the purpose of mitigating an analyzed event.
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: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Section Chief: L. Raghavan.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo
County, California.
Date of amendment requests: September 12, 2003.
Description of amendment requests: The proposed license amendments
would revise Technical Specification (TS) 3.3.1, ``Reactor Trip System
(RTS) Instrumentation,'' and TS 3.3.2, ``Engineered Safety Feature
Actuation System (ESFAS) Instrumentation,'' to change the current steam
generator (SG) narrow range (NR) water level-low low setpoints from
greater than or equal to 7.0 percent allowable value and 7.2 percent
nominal value, to greater than or equal to 14.8 percent allowable value
and 15.0 percent nominal value.
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 protection system performance will remain within the bounds
of the previously performed accident analyses since there are no
hardware changes and the actuation logic changes are conservative.
The design of the steam generator (SG) water level sensing equipment
and the coincidence logic will be unaffected. The only physical
change to the reactor trip system (RTS) and the engineered safety
feature actuation system (ESFAS) instrumentation is the increased
actuation setpoints. These changes have already been implemented in
the plant through the design change process. These changes are in
the conservative direction, i.e., a trip actuation signal will be
generated sooner for an event that challenges the ability of the SGs
to provide a heat sink for the reactor. In all other regards, the
design of the RTS and ESFAS instrumentation will be unaffected.
These protection systems will continue to function in a manner
consistent with the plant design basis.
[[Page 66139]]
The probability and consequences of accidents previously
evaluated in the Final Safety Analysis Report Update (FSARU) are not
adversely affected because changes to the RPS and ESFAS trip
setpoints assure a conservative response of the affected trip
functions, consistent with the safety analyses and licensing basis.
The proposed changes will not affect the probability of any
accident initiators. There will be no degradation in the performance
of, or an increase in the number of challenges imposed on, safety-
related equipment assumed to function during an accident. There will
be no change to normal plant operating parameters or accident
mitigation performance.
The proposed changes will not alter any assumptions or change
any mitigation actions in the radiological consequence evaluations
in the FSARU.
Therefore the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not change any hardware or the design
functions of any structures, systems or components involved, other
than to revise the SG narrow range (NR) water level-low low
setpoints; changes that have already been implemented. The proposed
changes will not affect the normal method of plant operation or
change any operating parameters. No new accidents, accident
initiators, or failure mechanisms are created by the proposed
changes.
Therefore, the proposed changes do not create the possibility of
a new or different accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The SG NR water level-low low setpoints specified in the
Technical Specifications have already been increased in the
conservative direction. The safety analysis limits assumed in the
transient and accident analyses remain unchanged. None of the
acceptance criteria for any accident analysis are changed.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo
County, California.
Date of amendment requests: October 22, 2003.
Description of amendment requests: The proposed license amendments
would revise Surveillance Requirement 3.6.3.7 of Technical
Specification (TS) 3.6.3, ``Containment Isolation Valves,'' by
extending the leakage rate testing frequency of the containment purge
supply and exhaust and vacuum/pressure relief valves, all with
resilient seals, from 184 days to 24 months. The amendments would also
delete the requirement to leakage rate test the containment vacuum/
pressure relief valves within 92 days after opening.
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.
Operability and leakage control effectiveness of the containment
purge supply and exhaust and containment vacuum/pressure relief
isolation valves have no effect on whether an accident occurs.
Consequently, increasing the interval between surveillances of
isolation valve leak rate does not involve any significant increase
in the probability of an accident previously evaluated. The
consequences of a unisolated reactor containment building at the
time of a fuel-handling accident or loss of coolant accident (LOCA)
are the release of radionuclides to the environment. Offsite
exposures due to containment leakage during a LOCA and fuel-handling
accident have been evaluated in Final Safety Analysis Report Update
(FSARU) sections 15.5.17.3 and 15.5.22, respectively. For a LOCA,
the Diablo Canyon Power Plant (DCPP) analyses assume containment
leakage of 0.1 percent of the containment volume per day for the
first 24 hours and 0.05 percent per day for the rest of the duration
of the accident. Calculated radiological exposures from the LOCA are
listed in FSARU Chapter 15, Table 15.5-75 and are within the 10 CFR
part 100 limits. The good performance history of these valves, along
with the very low total containment leakage rate, are reasonable
bases that there should not be any significant increase in the
consequences of [an] accident previously evaluated. For the fuel-
handling accident inside containment, DCPP analyses do not credit
these valves to provide a containment isolation function. It was
assumed that activity released from the containment refueling pool
is transported to the environment over a short time period through
the open equipment hatch. Calculated radiological exposures from the
fuel-handling accident inside containment are listed in FSARU
Chapter 15, Table 15.5-50 and are also within the 10 CFR part 100
limits. In summary, increasing the interval between leakage rate
surveillances of these isolation valves will not involve any
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve a modification to the
physical configuration of the plant (i.e., no new equipment will be
installed) or change in the methods governing normal plant
operation. The proposed change will not impose any new or different
requirements or introduce a new accident initiator, accident
precursor, or malfunction mechanism. The functions of the
containment purge and containment vacuum/pressure relief systems are
not altered by this change. Therefore, the proposed change does not
create the possibility of a new or different accident from any
accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
This proposed change only increases the interval between
surveillance tests of the containment purge supply and exhaust, and
containment vacuum/pressure relief valves. These valves have a good
performance history and should be able to perform their intended
containment isolation function reliably when called upon. In FSARU
Chapter 15, two offsite exposure scenarios are applicable to the
containment isolation function. These scenarios are LOCA containment
leakage and fuel-handling accident inside containment. For LOCA
containment leakage, the DCPP analyses assume containment leakage of
0.1 percent of the containment volume per day for the first 24 hours
and 0.05 percent per day for the remainder of the accident.
Calculated radiological exposures from a LOCA are listed in FSARU
Chapter 15, Table 15.5-75 and meet the 10 CFR part 100 limits. For
the fuel-handling accident inside containment, the DCPP analyses do
not credit these valves to provide a containment isolation function.
The analyses assume that activity released from the containment
refueling pool is transported to the environment over a short time
period through the open equipment hatch. Calculated radiological
exposures from the fuel-handling accident inside containment are
listed in FSARU Chapter 15, Table 15.5-50 and also meet the 10 CFR
part 100 limits. If in the unlikely event that these valves exceed
their leakage rate limits due to the extension of the surveillance
interval, the consequences will be consistent with the containment
leakage assumed in the accident analyses. Therefore, the extension
of leakage rate test interval will have an insignificant
radiological consequence, and the proposed change will not involve
any 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
[[Page 66140]]
amendment requests involve no significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo
County, California.
Date of amendment requests: October 22, 2003.
Description of amendment requests: The proposed license amendments
would revise Technical Specifications (TS) Section 5.5.9, ``Steam
Generator (SG) Tube Surveillance Program,'' and TS Section 5.6.10,
``Steam Generator (SG) Tube Inspection Report,'' to allow use of leak
limiting Alloy 800 sleeves to repair degraded SG tubes as an
alternative to plugging the SG tubes. The proposed amendments would
also remove an unnecessary reporting requirement contained in TS Table
5.5.9-2, ``Steam Generator (SG) Tube Inspection.''
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 leak limiting Alloy 800 sleeves are designed using the
applicable American Society of Mechanical Engineers (ASME) Boiler
and Pressure Vessel Code and, therefore, meet the design objectives
of the original steam generator (SG) tubing. The applied stresses
and fatigue usage for the sleeves are bounded by the limits
established in the ASME Code. Mechanical testing has shown that the
structural strength of sleeves under normal, upset, emergency, and
faulted conditions provides margin to the acceptance limits. These
acceptance limits bound the most limiting (three times normal
operating pressure differential) burst margin recommended by NRC
Regulatory Guide 1.121, ``Bases for Plugging Degraded PWR Steam
Generator Tubes.'' Burst testing of sleeve-tube assemblies has
confirmed the analytical results and demonstrated that no
unacceptable levels of primary-to-secondary leakage are expected
during any plant condition.
The leak limiting Alloy 800 sleeve depth-based structural limit
is determined using NRC guidance and the pressure stress equation of
ASME Code, Section III with additional margin added to account for
the configuration of long axial cracks. A sleeved tube will be
plugged on detection of an imperfection in the sleeve or in the
pressure boundary portion of the original tube wall in the leak
limiting sleeve/tube assembly.
Evaluation of the repaired SG tube testing and analysis
indicates no detrimental effects on the leak limiting Alloy 800
sleeve or sleeved tube assembly from reactor system flow, primary or
secondary coolant chemistries, thermal conditions or transients, or
pressure conditions as may be experienced at Diablo Canyon Power
Plant (DCPP) Units 1 and 2. Corrosion testing and historical
performance of sleeve-tube assemblies indicates no evidence of
sleeve or tube corrosion considered detrimental under anticipated
service conditions.
The implementation of the proposed change has no significant
effect on either the configuration of the plant or the manner in
which it is operated. The consequences of a hypothetical failure of
the leak limi[ti]ng Alloy 800 sleeve-tube assembly is bounded by the
current SG tube rupture (SGTR) analysis described in the DCPP Final
Safety Analysis Report Update. Due to the slight reduction in the
inside diameter caused by the sleeve wall thickness, primary coolant
release rates through the parent tube would be slightly less than
assumed for the SGTR analysis and therefore, would result in lower
total primary fluid mass release to the secondary system. A main
steam line break or feedwater line break will not cause a SGTR since
the sleeves are analyzed for a maximum accident differential
pressure greater than that predicted in the DCPP safety analysis.
The sleeve-tube assembly leakage during plant operation would be
minimal and is well within the Technical Specification (TS) leakage
limits.
The proposed change to TS 5.5.9 Table 5.5.9-2, ``Steam Generator
(SG) Tube Inspection,'' to delete the requirement to notify the NRC
pursuant to 10 CFR 50.72(b)(2) if the first sample inspection or the
second sample inspection results in a C-3 classification, is an
administrative change only and does not affect plant equipment or
accident analyses.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The leak limiting Alloy 800 sleeves are designed using the
applicable ASME Code as guidance, and therefore meet the objectives
of the original SG tubing. As a result, the functions of the SG will
not be significantly affected by the installation of the proposed
sleeve. The proposed sleeves do not interact with any other plant
systems. Any accident as a result of potential tube or sleeve
degradation in the repaired portion of the tube is bounded by the
existing SGTR accident analysis. The continued integrity of the
installed sleeve-tube assembly is periodically verified by the TS
requirements and a sleeved tube will be plugged on detection of an
imperfection in the sleeve or in the pressure boundary portion of
the original tube wall in the leak limiting sleeve/tube assembly.
Implementation of the proposed change has no significant effect
on either the configuration of the plant, or the manner in which it
is operated. The proposed change to delete the requirement to notify
the NRC pursuant to 10 CFR 50.72(b)(2) from TS 5.5.9 Table 5.5.9-2
is an administrative change only and does not affect plant equipment
or accident analyses.
Therefore, the proposed change does not create the possibility
of a new or different accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The repair of degraded SG tubes with leak limiting Alloy 800
sleeves restores the structural integrity of the degraded tube under
normal operating and postulated accident conditions and thereby
maintains current core cooling margin as opposed to plugging the
tube and taking it out of service. The design safety factors
utilized for the sleeves are consistent with the safety factors in
the ASME Boiler and Pressure Vessel Code used in the original SG
design. The sleeve and portions of the installed sleeve-tube
assembly that represent the reactor coolant pressure boundary will
be monitored and a sleeved tube will be plugged on detection of an
imperfection in the sleeve or in the pressure boundary portion of
the original tube wall in the leak limiting sleeve/tube assembly.
Use of the previously identified design criteria and design
verification testing assures that the margin to safety is not
significantly different from the original SG tubes.
The proposed change to delete the requirement to notify the NRC
pursuant to 10 CFR 50.72(b)(2) from TS 5.5.9 Table 5.5.9-2 is an
administrative change only, does not affect plant equipment or
accident analyses, does not relax any safety system settings, and
does not relax the bases for any limiting conditions for operations.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas.
Date of amendment request: November 4, 2003.
Description of amendment request: The proposed amendments would
revise the South Texas Project, Units 1 and 2 Technical Specifications
for the Remote Shutdown System to reflect requirements consistent with
those in NUREG-1431, ``Standard Technical Specifications--Westinghouse
Plants.''
[[Page 66141]]
The proposed changes would increase the allowed outage time for
inoperable Remote Shutdown System components to a time that is more
consistent with their safety significance. It would also relocate the
description of the required components to the Bases where it will be
directly controlled by the licensee.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Because the proposed changes do not involve potential accident
initiators, there is no significant increase in the probability of
an accident previously evaluated. There is no proposed change to the
design basis or configuration of the plant and the extension of the
allowed outage time of the Remote Shutdown System functions does not
have a significant effect on safety. Consequently there is no
significant increase in the consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not affect how the plant is operated or
involve any physical changes to the plant. Therefore there is no
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Except for extending the allowed outage time for Remote Shutdown
System function from 7 days to 30 days, the proposed changes are
essentially administrative. The evaluation of the extension of the
allowed outage time demonstrated that there was no 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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: Robert A. Gramm.
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, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by e-mail to [email protected].
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan.
Date of application for amendment: June 24, 2003.
Brief description of amendment: The amendment revises Technical
Specification 3.1.8, ``Scram Discharge Volume (SDV) Vent and Drain
Valves,'' to allow a vent or drain line with one inoperable valve to be
isolated instead of requiring the valve to be restored to Operable
status within 7 days.
Date of issuance: October 30, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 157.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 19, 2003 (68 FR
49815).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 30, 2003.
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 10, 2003.
Brief description of amendments: The amendments revised the
Technical Specifications to remove requirements that are no longer
applicable because the implementation of the automatic feedwater
isolation system modification has been completed on all three Oconee
units.
Date of Issuance: November 5, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 336, 336, & 337.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 19, 2003 (68 FR
49816). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 5, 2003.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station,
Unit 1, Claiborne County, Mississippi.
Date of application for amendment: April 3, 2003.
Brief description of amendment: The changes revise the Updated
Final Safety Analysis Report to change the Reactor Vessel Material
Surveillance Program. The change reflects participation in the Boiling
Water Reactor Vessel and Internals Project Integrated Surveillance
Program.
Date of issuance: November 4, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
[[Page 66142]]
Amendment No: 160.
Facility Operating License No. NPF-29: The amendment revises the
Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: May 13, 2003 (68 FR
25653).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 4, 2003.
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: October 23, 2001, as
supplemented on March 29 and December 17, 2002, and June 12, 2003.
Brief description of amendment: The amendment revised Technical
Specification (TS) 5.5.10, ``Ventilation Filter Testing Program,'' to
adopt the requirements of the American Society for Testing and
Materials Standard D3803-1989, ``Standard Test Method for Nuclear-Grade
Activated Carbon.'' The TS revisions are in response to Nuclear
Regulatory Commission (NRC) Generic Letter (GL) 99-02, ``Laboratory
Testing of Nuclear-Grade Activated Charcoal.'' The amendment revises
the TSs: (1) To provide a control room ventilation system (CRVS) methyl
iodide removal efficiency of greater than or equal to 95.5% and remove
the notation that there is a 1-inch charcoal bed depth; (2) to allow
for the continued use of the existing CRVS through Refueling Outage 13,
in order to design, fabricate, and install a 2-inch charcoal filter
bed; and (3) to add a note in the TS requiring a demonstration of
charcoal efficiency of 93% when changing the charcoal in the existing
CRVS bed prior to any fuel movement in the upcoming Refueling Outage 12
and every 6 months thereafter until the new beds are installed. The NRC
had previously published a notice of consideration on December 12, 2001
(66 FR 64292) regarding a similar proposal from the licensee in
response to GL 99-02. However, in response to a request for additional
information from the NRC dated March 29, 2002, the licensee revised its
application and withdrew the prior request to change the maximum CRVS
differential pressure in TS 5.5.10.d.
Date of issuance: October 30, 2003.
Effective date: As of the date of issuance and shall be implemented
30 days from the date of issuance.
Amendment No.: 219.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 18, 2003 (68 FR
12951).
The March 29 and December 17, 2002, and June 12, 2003, letters
provided clarifying information that did not enlarge the scope of the
amendment request or 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 30, 2003.
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: December 16, 2002, as supplemented by
letters dated July 30, and September 29, 2003.
Brief description of amendment: The amendment adds Combustion
Engineering topical report CEN-372-P-A, May 1990, ``Fuel Rod Maximum
Allowable Gas Pressure,'' to the list of topical reports in Technical
Specification 6.9.1.11.1, used to determine the Waterford Steam
Electric Sation, Unit 3 core operating limits. In addition, the
amendment approves the deletion of applicable dates and revision
numbers for CEN-372-P-A and other topical reports listed in TS
6.9.1.11.1.
Date of issuance: October 31, 2003.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 191.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5673). The July 30, and September 29, 2003, supplemental letters
provided clarifying information that did not change the scope of the
original Federal Register notice or the original no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 31, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374,
LaSalle County Station, Units 1 and 2, LaSalle County, Illinois.
Date of application for amendments: March 31, 2003.
Brief description of amendments: The amendments revise Appendix A,
Technical Specifications (TS), of Facility Operating License Nos. NPF-
11 and NPF-18. Specifically, the changes modify TS 5.7, ``High
Radiation Area,'' by incorporating the wording and requirements from
NUREG-1434, ``Standard Technical Specifications General Electric
Plants, BWR/6,'' Revision 2, dated June 2001. The revision also
includes administrative changes regarding access control and
terminology for high radiation areas.
Date of issuance: October 31, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 161/147.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 27, 2003 (68 FR
28852).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 31, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353.
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania.
Date of application for amendments: December 20, 2002, as
supplemented May 30, 2003.
Brief description of amendments: The amendments removed the current
facility reactor material specimen surveillance schedule from the
Technical Specifications for Limerick Generating Station, Units 1 and 2
(LGS-1 and 2). The licensee also revised the Updated Final Safety
Analysis Report (UFSAR) for LGS-1 and 2 to reflect implementation of
the Boiling Water Reactor Vessel and Internals Project reactor pressure
vessel integrated surveillance program as the basis for demonstrating
the compliance with the requirements of Appendix H, ``Reactor Vessel
Material Surveillance Program Requirements,'' to title 10 of the Code
of Federal Regulations, Part 50.
Date of issuance: November 4, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 167 and 130.
Facility Operating License Nos. NPF-39 and NPF-85: The amendments
revised the Technical Specifications and authorized changes to the
UFSAR for LGS-1 and 2.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5669). The supplement dated May 30, 2003, provided additional
information that
[[Page 66143]]
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluationdated November 4, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, and PSEG Nuclear LLC,
Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station,
Units 2 and 3, (PBAPS-2 and 3) York County and Lancaster County,
Pennsylvania.
Date of application for amendments: December 20, 2002, as
supplemented May 30, 2003.
Brief description of amendments: The amendments revised the Updated
Final Safety Analysis Report (UFSAR) for Peach Bottom Atomic Power
Station, Units 2 and 3, by allowing implementation of the Boiling Water
Reactor Vessel and Internals Project reactor pressure vessel integrated
surveillance program as the basis for demonstrating the compliance with
the requirements of Appendix H, ``Reactor Vessel Material Surveillance
Program Requirements,'' to Title 10 of the Code of Federal Regulations,
Part 50.
Date of issuance: November 4, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 249 and 253.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The
amendments authorized changes to the UFSAR for PBAPS-2 and 3.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5669). The supplement dated May 30, 2003, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated November 4, 2003.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit 2, Berrien County, Michigan.
Date of application for amendment: March 27, 2003, as supplemented
August 15, 2003.
Brief description of amendment: The amendment lowers the trip
setpoint and allowable value contained in Technical Specification (TS)
Table 3.3-4 for the pressurizer pressure low safety injection signal.
The amendment also lowers the value for the P-11 setpoint in TS Table
3.3-3. These changes increase the margin between the low pressurizer
pressure safety injection actuation setpoint and the minimum
pressurizer pressure that occurs immediately following a reactor trip.
Date of issuance: November 12, 2003.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 263.
Facility Operating License No. DPR-74: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 27, 2003 (68 FR
28853).
The supplemental letter contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 12, 2003.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska.
Date of amendment request: December 31, 2002, as supplemented by
letter dated July 24, 2003.
Brief description of amendment: The amendment revises the Updated
Safety Analysis Report (USAR) reflecting a change of the reactor vessel
material surveillance program to incorporate the Boiling Water Reactor
Vessel and Internals Project Integrated Surveillance Program into the
licensing basis.
Date of issuance: October 31, 2003.
Effective date: As of the date of issuance. The amendment shall be
implemented within 30 days of issuance and the USAR changes shall be
implemented in the next periodic update to the USAR in accordance with
10 CFR 50.71(e).
Amendment No.: 201.
Facility Operating License No. DPR-46: Amendment revised the USAR.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5678).
The July 24, 2003, supplemental letter provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register on February 4, 2003 (68 FR 5678).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 31, 2003.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska.
Date of amendment request: January 27, 2003, as supplemented by
letter dated August 1, 2003.
Brief description of amendment: The amendment authorizes revisions
to the Updated Safety Analysis Report (USAR) to incorporate the NRC
approval of the GOTHIC 7.0 computer program for performing containment
analyses.
Date of issuance: November 5, 2003.
Effective date: November 5, 2003, and shall be implemented within
30 days of the date of issuance. The implementation of the amendment
includes the incorporation into the USAR the changes discussed above,
as described in the licensee's application dated January 27, 2003, and
supplement dated August 1, 2003, and evaluated in the staff's Safety
Evaluation attached to the amendment.
Amendment No.: 222.
Renewed Facility Operating License No. DPR-40: The amendment
revised the USAR.
Date of initial notice in Federal Register: March 18, 2003 (68 FR
12956).
The August 1, 2003, supplemental letter 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 November 5, 2003.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska.
Date of amendment request: January 27, 2003, as supplemented by
letter dated October 14, 2003.
Brief description of amendment: The amendment deletes Technical
Specification (TS) 2.3(2)i and the corresponding Bases that allows the
performance of the surveillance test of Table 3-2, Item 20
(Recirculation Actuation Logic Channel Functional Test) under
administrative controls, while components in excess of those allowed by
Conditions a, b, d, and e of TS 2.3(2) are inoperable, provided they
are returned to operable status within one hour. This allowance was
granted in Amendment No. 206 issued April 19, 2002, and only applied
until the end of Cycle 21.
[[Page 66144]]
Date of issuance: November 10, 2003.
Effective date: November 10, 2003, and shall be implemented within
60 days from the date of issuance.
Amendment No.: 223.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: March 18, 2003 (68 FR
12955).
The October 14, 2003, supplemental letter provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 10, 2003.
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: May 6, 2003, as supplemented by
letters dated August 12 and September 18, 2003.
Brief description of amendments: These amendments deleted Technical
Specification (TS) 3.3.1.3, ``Oscillation Power Range Monitor (OPRM)
Instrumentation,'' and revised TS 3.4.1, ``Recirculation Loops
Operating,'' to formally extend the currently implemented requirements,
which define appropriately conservative restrictions to plant operation
and operator response to thermal hydraulic instability events. In
addition, the amendments revise TS 3.4.1 to refer to the power flow map
in the core operating limits report and include a reference in TS
5.6.5.
Date of issuance: October 29, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 215 and 190.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 24, 2003 (68 FR
37582).
The supplemental letters dated August 12 and September 18, 2003,
provided clarifying information that did not change the scope of the
amendment as described in the initial notice of the proposed action
published in the Federal Register notice (68 FR 37582, June 24, 2003),
or the U.S. Nuclear Regulatory Commission staff's proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 29, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama.
Date of application for amendments: July 25, 2003.
Description of amendment request: The amendments revised Technical
Specification 3.1.8, ``Scram Discharge Volume (SDV) Vent and Drain
Valves,'' to allow a vent or drain line with one inoperable valve to be
isolated instead of requiring the valve to be restored to operable
status within 7 days.
Date of issuance: November 3, 2003.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment Nos.: 248, 285, and 243.
Facility Operating License Nos. DPR-33, DPR-52, and DPR-68.
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 18, 2003 (68
FR 54753).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 17th day of November, 2003.
For the Nuclear Regulatory Commission.
Eric Leeds,
Deputy Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 03-29107 Filed 11-24-03; 8:45 am]
BILLING CODE 7590-01-P