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Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations
Note: EPA no longer updates this information, but it may be useful as a reference or resource.
[Federal Register: May 31, 2000 (Volume 65, Number 105)]
[Notices]
[Page 34743-34755]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr31my00-96]
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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 May 6, 2000, through May 19, 2000. The last
biweekly notice was published on May 17, 2000 (65 FR 31354).
Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By June 30, 2000, 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 Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC, and electronically from
the ADAMS Public Library component on the NRC Web site, http://
www.nrc.gov (the Electronic Reading Room). 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
[[Page 34744]]
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: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and electronically from the ADAMS Public
Library component on the NRC Web site, http://www.nrc.gov (the
Electronic Reading Room).
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: April 18, 2000.
Description of amendment request: The amendments would revise
Technical Specifications (TS) 3.7.10 and 3.7.12 for Catawba Units 1 and
2. The proposed changes address degraded pressure boundaries on the
Auxiliary Building Filtered Ventilation Exhaust System and the Control
Room Area Ventilation System. The proposed changes in TS 3.7.10 and
3.7.12 would add Notes which allow the affected ventilation system
boundaries to be opened intermittently under administrative controls.
Also, it would add a new condition in TS 3.7.10 and 3.7.12. This new
condition will require that the boundaries for these two systems be
returned to an operable status within 24 hours, when both trains of
these systems are inoperable due to an inoperable boundary.
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:
In accordance with the criteria set forth in 10 CFR 50.91 and
50.92, Duke Energy Corporation has evaluated this license amendment
request and determined it does not represent a significant hazards
consideration. The following is provided in support of this
conclusion.
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The Control Room Area Ventilation System (CRAVS), Control
Room pressure boundary, the Auxiliary Building Filtered Ventilation
Exhaust System (ABFVES), or the Emergency Core Cooling System (ECCS)
pump rooms area pressure boundary are not assumed to be initiators
of any analyzed accident. Therefore, the proposed changes contained
in this LAR [license amendment request] have no significant impact
on the probability of occurrence of any previously analyzed
accident.
The proposed new condition for the CRAVS and ABFVES Technical
Specifications (TS) would permit a 24-hour period to take action to
restore an inoperable pressure boundary to OPERABLE status. The
consequences of implementing the 24 hour Completion Time are
reasonable based upon: (1) The low probability of a design basis
accident occurring during this time period, (2) additional actions
that are available to the operator to minimize doses (e.g., self
contained breathing apparatus and alternate control room air
intakes), and (3) the availability of an operable CRAVS/ABFVES train
to provide a filtered environment (albeit with potential unfiltered
leakage).
For cases where any of the affected control room or ECCS pump
room area/pump rooms pressure boundaries are opened intermittently
under administrative controls, appropriate compensatory measures
would be required by the proposed TS to ensure the pressure boundary
can be rapidly restored. Based on the compensatory measures
available to the plant operators and the administrative controls
required to rapidly restore an opened pressure boundary, the
accident consequences do not cause an increase in dose above the
applicable General Design Criteria, Standard Review Plan, or 10 CFR
100 limits. The plant operators will continue to maintain the
ability to mitigate a design basis event.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. No changes are being made to actual plant hardware which
will result in any new accident causal mechanisms. Also, no changes
are being made to the way in which the plant is being operated.
Therefore, no new accident causal mechanisms will be generated.
3. Does this change involve a significant reduction in a margin
of safety?
No. Margin of safety is related to the ability of the fission
product barriers to perform their design functions during and
following accident conditions. These barriers include the fuel
cladding, the reactor coolant system, and the containment system.
The performance of these barriers will not be significantly degraded
by the proposed changes. The proposed changes would allow affected
pressure boundaries to be degraded for a limited period of time (24
hours).
[[Page 34745]]
However, the probability of a design basis event occurring during
this time is low and additional actions (e.g., breathing apparatus)
would also be taken to minimize dose to the plant operators. When
the boundaries are open on an intermittent basis, as permitted by
the changes proposed in this LAR, administrative controls would be
in place to ensure that the integrity of the pressure boundaries
could be rapidly restored. Therefore, it is expected that the plant,
and the operators, would maintain the ability to mitigate design
basis events and none of the fission product barriers would be
affected by this change. Therefore, the proposed change is not
considered to 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: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: Richard L. Emch, Jr.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: April 5, 2000.
Description of amendment request: The proposed amendment implements
technical specification (TS) changes associated with thermal-hydraulic
stability monitoring. New TS 3.3.1.3, ``Oscillation Power Range Monitor
(OPRM) Instrumentation,'' will provide the minimum operability
requirements for the OPRM channels, the Required Actions when they
become inoperable, and appropriate surveillance requirements. The OPRMs
will provide automatic ``detect and suppress'' actions to replace the
administrative controls currently in effect through operator training
and manual actions. The amendment would remove monitoring guidance from
TS 3.4.1, ``Recirculation Loops Operating,'' that will no longer be
necessary due to the activation of the automatic OPRM instrumentation.
Finally, the amendment would update TS 5.6.5, ``Core Operating Limits
Report (COLR),'' to require the applicable setpoints for the OPRMs to
be included in the COLR.
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 specifies limiting conditions for operation,
required actions and surveillance requirements for the Oscillation
Power Range Monitor (OPRM) system, and allows operation in regions
of the power to flow map currently restricted by the requirements of
Interim Corrective Actions (ICAs) and certain limiting conditions of
operation of Technical Specification (TS) 3.4.1. The restrictions of
the ICAs and TS 3.4.1 were imposed to ensure adequate capability to
detect and suppress conditions consistent with the onset of thermal-
hydraulic (T-H) oscillations that may develop into a T-H instability
event. A T-H instability event has the potential to challenge the
Minimum Critical Power (MCPR) safety limit. The OPRM system can
automatically detect and suppress conditions necessary for T-H
instability. With the activation of the OPRM System, the
restrictions of the ICAs and TS 3.4.1 will no longer be required.
The probability of a T-H instability event is impacted by power
to flow conditions during operation inside specific regions of the
power to flow map, in combination with power shape and inlet
enthalpy conditions, such that only under such conditions can the
occurrence of an instability event be postulated to occur. Operation
in these regions may increase the probability that operation with
conditions necessary for a T-H instability can occur. However, when
the OPRM is OPERABLE with operating limits as specified in the Core
Operating Limits Report (COLR), the OPRM can automatically detect
the onset of significant local power oscillations and generate a
trip signal. Actuation of a Reactor Protection System (RPS) trip
will suppress conditions necessary for T-H instability and decrease
the probability of a T-H instability event. In the event the trip
capability of one or more of the OPRM channels is not maintained,
the proposed change includes Required Actions which limit the period
of time before the affected OPRM channel (or RPS system) must be
placed in the tripped condition. If these actions would result in a
trip function such as a scram, or if the OPRM trip capability is not
maintained, an alternate method to detect and suppress thermal
hydraulic oscillations is required, i.e., the same ICAs as are in
place today. In either case the duration of the period of time
allowed by the Required Actions is limited, and the probability of a
T-H instability event during this limited time is not significantly
increased.
Several changes to TS 3.4.1 are made which are more consistent
with, or conservative with, respect to the reviewed and approved
Standard Technical Specifications for Boiling Water Reactors. These
generic changes are considered applicable to the Perry Nuclear Power
Plant. They simply provide guidance on the operator actions to be
taken and the associated time limits when the Specification is
entered, and do not impact the probability of occurrence of an
accident. For the above reasons, the proposed change does not result
in a significant increase in the probability of an accident
previously evaluated.
An unmitigated T-H instability event is postulated to cause a
violation of the MCPR safety limit. The proposed change ensures
mitigation of T-H instability events prior to challenging the MCPR
safety limit if initiated from anticipated conditions, by detection
of the onset of oscillations and actuation of an RPS trip signal.
The OPRM also provides the capability of an RPS trip being generated
for T-H instability events initiated from unanticipated but
postulated conditions. These mitigating capabilities of the OPRM
system will become available as a result of the proposed change and
have the potential to reduce the consequences of anticipated and
postulated T-H instability events. The OPRM installation has been
evaluated to not adversely impact other installed equipment such as
the Average Power Range Monitors (APRMs) or the RPS in a manner that
could prevent response to various postulated events, so those events
will not have increased consequences due to the OPRMs. Therefore,
the proposed change does not significantly increase the consequences
of an accident previously evaluated.
Therefore, the proposed change, which specifies limiting
conditions for operation, required actions and surveillance
requirements for the OPRM system, and allows operation in certain
regions of the power to flow map, does not significantly increase
either the probability or consequences of an accident previously
evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change specifies limiting conditions for operation,
required actions and surveillance requirements of the OPRM system,
and allows operation in regions of the power to flow map currently
restricted by the requirements of ICAs and TS 3.4.1. The OPRM system
uses input signals shared with APRM and rod block functions to
monitor core conditions and generate an RPS trip when required.
Quality requirements for software design, testing, implementation
and module self-testing of the OPRM system provide assurance that
new equipment malfunctions due to software errors are not created.
The design of the OPRM system also ensures that neither operation
nor malfunction of the OPRM system will adversely impact the
operation of other systems and no accident or equipment malfunction
of these other systems could cause the OPRM system to malfunction or
cause a different kind of accident. Therefore, operation with the
OPRM system does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
Operation in regions currently restricted by the requirements of
ICAs and TS 3.4.1 is within the nominal operating domain and ranges
of plant systems and components, and within the range for which
postulated accidents have been evaluated. Therefore operation within
these regions does not
[[Page 34746]]
create the possibility of a new or different kind of accident from
any accident previously evaluated. The changes to TS 3.4.1 to be
more consistent, or conservative, with respect to the reviewed and
approved Standard Technical Specifications, simply provide guidance
on the operator actions to be taken and the associated time limits
when the Specification is entered, and also do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Therefore, the proposed change, which specifies limiting
conditions for operation, required actions and surveillance
requirements of the OPRM system, and allows operation in certain
regions of the power to flow map, does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change will not involve a significant reduction
in the margin of safety.
The proposed change specifies limiting conditions for operation,
required actions and surveillance requirements of the OPRM system
and allows operation in regions of the power to flow map currently
restricted by the requirements of ICAs and TS 3.4.1.
The OPRM system monitors small groups of LPRM [local power range
monitor] signals for indication of local variations of core power
consistent with T-H oscillations, and generates an RPS trip when
conditions consistent with the onset of oscillations are detected.
An unmitigated T-H instability event has the potential to result in
a challenge to the MCPR safety limit. The OPRM system provides the
capability to automatically detect and suppress conditions which
might result in a T-H instability event, and thereby maintains the
margin of safety by providing automatic protection for the MCPR
safety limit while reducing the burden on the control room
operators. Therefore, operation with the OPRM system does not
involve a significant reduction in a margin of safety. In the event
an OPRM channel becomes inoperable, the proposed change includes
actions which limit the period of time before the affected OPRM
channel (or RPS system) must be placed in the tripp[ed] condition.
If these actions would result in a trip function such as a scram (or
if the OPRM trip capability is not maintained), the alternate method
to detect and suppress thermal hydraulic oscillations (the current
ICAs) is required to be put in place. The duration of the period of
time allowed by the Required Actions is limited, and the probability
of a significant T-H instability event during this limited time is
not significantly increased.
Operation in regions currently restricted by the requirements of
ICAs and Technical Specification [TS] 3.4.1 is within the nominal
operating domain and ranges of plant systems and components, and
within the range assumed for initial conditions considered in the
analysis of anticipated operational occurrences and postulated
accidents. Therefore, operation in these regions does not involve a
significant reduction in the margin of safety. The changes to TS
3.4.1 to be more consistent, or conservative, with respect to the
reviewed and approved Standard Technical Specifications, simply
provide guidance on the operator actions to be taken and the
associated time limits when the Specification is entered, and also
do not significantly reduce the margin of safety.
Therefore, the proposed change, which specifies limiting
conditions for operation, required actions and surveillance
requirements of the OPRM system, and allows operation in certain
regions of the power to flow map, 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 E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
Florida Power and Light Company, et al., Docket No. 50-335, St. Lucie
Plant, Unit No. 1, St. Lucie County, Florida
Date of amendment request: April 23, 2000.
Description of amendment request: The proposed license amendment
(PLA) is associated with the required timing for containment hydrogen
recombiner post operation insulation resistance testing. This PLA
revises Unit 1 Technical Specification 3/4.6.4.2, Electric Hydrogen
Recombiners--W, to clarify the requirement for the post-operation
insulation resistance test of Surveillance Requirement 4.6.4.2.b.4.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendment does not involve an increase in the
probability or consequences of any accident previously evaluated.
This PLA provides a clarification of the Technical Specification
surveillance requirements for verifying hydrogen recombiner
operability and reliability. This PLA has no affect on the testing
requirements, test frequency, or acceptance criteria for recombiner
operability. This change allows vendor recommended guidance and in-
house methodology to be established when conducting recombiner
heater resistance testing. This will enable consistency in testing
and will allow trending for determination of the material condition
of the recombiner heaters. The PLA change provides clarification and
preserves the intent of the basis to monitor the material condition
of the recombiner heaters. Additionally, this change provides
consistency and is identical with the Unit 2 Technical Specification
surveillance. As such, this change is considered administrative in
nature.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment will not create the possibility of a new
or different kind of accident from any accident previously
evaluated. This PLA is considered administrative in nature and will
not alter the way in which the hydrogen recombiner is operated or
tested. This PLA allows vendor recommended guidance to be
established in order to perform consistent testing and to allow
meaningful trending of the results to verify recombiner operability.
This PLA has no affect on the testing requirements, test frequency,
or acceptance criteria for recombiner operability. This PLA does not
result in any plant configuration changes or new failure modes.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendment does not involve a reduction in the
margin of safety. This administrative PLA clarifies the surveillance
requirement of the subject Technical Specification by allowing the
establishment of vendor recommendations and in-house testing
methodology to provide consistent testing conditions and allow
meaningful trending of results. This PLA has no affect on the
testing requirements, test frequency, or acceptance criteria for
recombiner operability. As such, the assumptions and conclusions of
the accident analyses in the UFSAR [Updated Final Safety Analysis
Report] remain valid and the associated safety limits will continue
to be met.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Richard P. Correia.
Florida Power and Light Company, Docket No. 50-251 and 50-252, Turkey
Point Units 3 and 4 in Miami-Dade County
Date of amendment request: April 27, 2000.
Description of amendment request: Florida Power and Light Company
(FPL)
[[Page 34747]]
requests to amend the Turkey Point Unit 3 Facility Operating License
DPR-31 Fire Protection license condition 3.G, and to amend the Turkey
Point Unit 4 Facility Operating License DPR-41 Fire Protection license
condition 3.F. The proposed revisions to the Facility Operating
Licenses are required to incorporate references to NRC Safety
Evaluations issued in support of 10 CFR 50 Appendix R granted
exemptions. In addition, the proposed amendments requests to modify
Appendix A of the Facility Operating Licenses DPR-31 and DPR-41 of the
Turkey Point Units 3 and 4 Technical Specifications (TS), Section
4.7.6.g. Due to an oversight, the submittal for the request of License
Amendments Nos. 201 and 195 for Section 6.0 ``Administrative
Controls,'' L-99-056, dated March 8, 1999, discussed revision to TS
Section 4.7.6.g on TS Page 3/4 7-21, but inadvertently did not attach
the revised marked up Page 3/4 7-21.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments do not involve an increase in the
probability or consequences of an accident previously evaluated
because the proposed changes are administrative in nature adding
references to exemptions granted by the NRC and to reflect
relocation of record retention requirements from the TS to the
UFSAR. These amendments will not involve a significant increase in
the probability or consequences of an accident previously evaluated
because they do not affect assumptions contained in plant safety
analyses, the physical design and/or operation of the plant, nor do
they affect Technical Specifications that preserve safety analysis
assumptions. Therefore, the proposed changes do not affect the
probability or consequences of accidents previously analyzed.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed changes to the Facility Operating Licenses and the
Technical Specifications are administrative in nature and can not
create the possibility of a new or different kind of accident from
any previously evaluated since the proposed amendments will not
change the physical plant or the modes of plant operation defined in
the facility operating license. No new failure mode is introduced
due to the administrative changes since the proposed changes do not
involve the addition or modification of equipment nor do they alter
the design or operation of affected plant systems, structures, or
components.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The operating limits and functional capabilities of the affected
systems, structures, and components are unchanged by the proposed
amendments. The proposed changes to the Facility Operating License
Conditions and the TS are administrative in nature and do not reduce
any of the margins of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Richard P. Correia.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: April 28, 2000.
Description of amendment request: The Seabrook Station Technical
Specifications (TSs) are proposed to be revised to implement the
Relaxed Axial Offset Control (RAOC) operating strategy in support of
the use of upgraded Westinghouse fuel with Intermediate Flow Mixers.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
The proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes to TS 2.1.1, 3.2.1, 4.2.1.1, 4.2.2.2, 4.2.2.3,
4.2.2.4, 6.8.1.6.b, and changes to the aforementioned TS Bases, are in
support of North Atlantic's long-term operating strategy to refuel and
operate, commencing with Cycle 8, with Biweekly Notice Coordinator
upgraded Westinghouse fuel with Intermediate Flow Mixers (VANTAGE+(w/
IFMs)). Evaluations/analyses of accidents which are potentially
affected by the parameters and assumptions associated with the fuel
upgrade and RAOC strategy have shown that all design standards and
applicable safety criteria will continue to be met. The consideration
of these changes does not result in a situation where the design,
material, and construction standards that were applicable prior to the
change are altered. Therefore, the proposed changes occurring with the
fuel upgrade will not result in any additional challenges to plant
equipment that could increase the probability of any previously
evaluated accident.
The proposed changes associated with the fuel upgrade and RAOC
strategy do not affect plant systems such that their function in the
control of radiological consequences is adversely affected. The actual
plant configuration, performance of systems, and initiating event
mechanisms are not being changed as a result of the proposed changes.
The design standards and applicable safety criteria limits will
continue to be met and therefore fission barrier integrity is not
challenged. The proposed changes associated with fuel upgrade and RAOC
strategy have been shown not to adversely affect the response of the
plant to postulated accident scenarios. The proposed changes will
therefore not affect the mitigation of the radiological consequences of
any accident described in the Updated Final Safety Analysis Report
(UFSAR).
The proposed changes to TS Table 2.2-1, TS 3.2.2, TS 3.2.3, and the
title on page 3/4 2-6 are editorial changes to correct either
typographical errors, simplification of statements, clarification of
specific parameters associated with temperature pressure measurements,
making some notations consistent with improved Standard Technical
Specifications -- Westinghouse Plants, NUREG-1431, Rev. 1, and
relocating additional cycle-specific values for temperature, pressure
and time constants to the [Core Operating Limits Report] COLR, or
correcting an erroneous title. These changes do not result in a change
to the design basis of any plant structure, system or component or
parameters currently specified in the COLR, therefore, operation of the
facility within the prescribed limits of TS remains unchanged.
The proposed change to TS 3.2.1, ACTION a.2, to delete the need to
reduce the power range neutron flux high trip setpoints subsequent to
reducing rated thermal power (RTP) to less than 50% whenever axial flux
difference (AFD) is outside of the applicable limits specified in the
COLR, does not significantly increase the
[[Page 34748]]
probability or consequences of an accident previously evaluated.
Therefore, for the reasons stated above, the probability or
consequences of an accident previously evaluated are not significantly
increased for all the proposed TS changes presented herein.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The possibility for a new or different type of accident from any
accident previously evaluated is not created since the proposed changes
associated with the fuel upgrade and RAOC strategy do not result in a
change to the design basis of any plant structure, system or component.
These proposed changes do not cause the initiation of any accident nor
create any new failure mechanisms. Equipment important to safety will
continue to operate as designed. Component integrity is not challenged.
The proposed changes do not result in any event previously deemed
incredible being made credible.
The proposed changes are not expected to result in conditions that
are more adverse and are not expected to result in any increase in the
challenges to safety systems.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes will assure continued compliance within the
acceptance limits previously reviewed and approved by the NRC for use
of upgraded fuel features with RAOC. All of the appropriate acceptance
criteria for the various analyses and evaluations will continue to be
met.
The proposed editorial changes do not change the current limits
specified in Technical Specifications.
Removing the requirement for manually reducing the power range
neutron flux high trip setpoint does not result in a significant
reduction in a margin of safety. There are other levels of trip
protection to terminate a rapid rise in power excursion, such as the
overtemperature delta-temperature (OT-T) trip function and previous
power range neutron flux high trip setpoint. In addition, a rapid rise
in power to greater than 50 percent RTP with AFD outside limits does
not immediately create an unacceptable situation. The increased
potential for a reactor trip caused by the manual manipulation of the
setpoint needlessly exposes the plant to an unnecessary trip with the
potential for an undesirable plant transient which may unnecessarily
challenge safety systems.
Therefore, the proposed aforementioned TS changes do not involve a
signification reduction in a margin of safety.
Based on this review, it appears that the three standards of
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Section Chief: James W. Clifford.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: February 1, 2000, as supplemented by
letter dated April 13, 2000.
Description of amendment request: The proposed amendment proposes
changes to the cable spreading room technical specifications to permit
pressurizing the cable spreading room to a pressure that exceeds the
pressure of the adjacent control room envelope area during testing.
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:
In accordance with 10 CFR 50.92, NNECO has reviewed the proposed
changes and has concluded that they do not involve a significant
hazards consideration (SHC). The basis for this conclusion is that
the three criteria of 10 CFR 50.92(c) are not compromised. The
proposed changes do not involve an SHC because the changes would
not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed Technical Specification and Bases changes to
exclude the requirements of Surveillance Requirements (SRs)
4.7.7.e.2, 4.7.8.c.2, and 4.7.8.c.3 during pressurization testing of
the Cable Spreading Room (CSR) will not increase the probability of
an accident previously evaluated. Operation of the Control Room
Emergency Air Filtration System and the Control Room Envelope
Pressurization System cannot cause an accident to occur.
The proposed Technical Specification and Bases changes to
exclude the requirements of SRs 4.7.7.e.2, 4.7.8.c.2, and 4.7.8.c.3
during pressurization testing of the CSR may adversely impact the
consequences of previously evaluated accidents. During CSR
pressurization testing, the Control Room Emergency Air Filtration
and the Control Room Envelope Pressurization Systems may not be able
to pressurize and maintain the Control Room envelope at a positive
pressure with respect to adjacent areas and the outside atmosphere.
As a result, radioactivity released from a design basis accident may
enter the Control Room envelope. However, since the CSR area will
actually be at a higher pressure than the outside atmosphere (during
CSR pressurization testing), radioactive leakage into the CSR area,
and subsequently into the Control Room envelope, should not occur
after the temporary fan has been stopped. Administrative controls
will be established to immediately stop the temporary fan and
rapidly depressurize the CSR area in the event Control Building
isolation is necessary. Once the CSR area is depressurized, the
Control Room Emergency Air Filtration System and the Control Room
Envelope Pressurization System will be able to function as designed
to mitigate the consequences of the accident. In addition, the
probability of a design basis accident (DBA) occurring while the CSR
is pressurized is low. Therefore, exempting the requirements of SRs
4.7.7.e.2, 4.7.8.c.2, and 4.7.8.c.3 during CSR pressurization
testing will not result in a significant increase in the
consequences of an accident previously evaluated.
The proposed Technical Specification and Bases change to clarify
the mode of operation of the Control Room Emergency Air Filtration
System when the pressurization requirement of SR 4.7.7.e.2 applies,
will have no adverse effect on plant operation, or the availability
or operation of any accident mitigation equipment. The plant
response to the design basis accidents will not change. In addition,
the proposed change can not cause an accident. Therefore, there will
be no significant increase in the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed Technical Specification and Bases changes to
exclude the requirements of SRs 4.7.7.e.2, 4.7.8.c.2, and 4.7.8.c.3
during pressurization testing of the CSR, and to clarify the mode of
operation of the Control Room Emergency Air Filtration System when
the pressurization requirement of SR 4.7.7.e.2 applies, will not
alter the plant configuration (no new or different type of permanent
equipment will be installed) or require any new or unusual operator
actions. Temporary equipment will be utilized to pressurize the CSR,
and administrative controls, using additional personnel beyond the
normal shift complement, will be implemented to restore the CSR to a
configuration that will allow the Control Room Emergency Air
Filtration System and the Control Room Envelope Pressurization
System to function as designed to mitigate the consequences of an
accident. The temporary equipment and administrative controls that
will be implemented to perform the CSR pressurization testing will
not introduce any new failure modes that could result in a new
accident. Also, the response of the plant and the operators
following these accidents is unaffected by the changes. Therefore,
the proposed changes will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
[[Page 34749]]
3. Involve a significant reduction in a margin of safety.
The proposed Technical Specification and Bases changes to
exclude the requirements of SRs 4.7.7.e.2, 4.7.8.c.2, and 4.7.8.c.3
during pressurization testing of the CSR may adversely impact the
ability of the Control Room Emergency Air Filtration System and the
Control Room Envelope Pressurization System to function as designed
to protect the Control Room Operators following a DBA, and to use
other accident mitigation equipment contained within the Control
Room envelope. However, the administrative controls that will be
established to immediately stop the temporary fan and rapidly
depressurize the CSR area if Control Building isolation is necessary
will provide reasonable assurance that the habitability of the
Control Room envelope will be maintained. Therefore, exempting the
requirements of SRs 4.7.7.e.2, 4.7.8.c.2, and 4.7.8.c.3 during CSR
pressurization testing will not result in a significant reduction in
a margin of safety.
The proposed Technical Specification and Bases change to clarify
the mode of operation of the Control Room Emergency Air Filtration
System when the pressurization requirement of SR 4.7.7.e.2 applies
will have no adverse effect on plant operation, or the availability
or operation of any accident mitigation equipment. The plant
response to the design basis accidents will not change. Therefore,
there will be no significant reduction in a margin of safety.
The proposed changes do not alter the design, function, or
operation of the equipment involved. The impact of the proposed
changes has been analyzed, and it has been determined they do not
involve a significant increase in the probability or consequences of
an accident previously evaluated, do not create the possibility of a
new or different kind of accident from any accident previously
evaluated, and do not involve a significant reduction in a margin of
safety. Therefore, NNECO has concluded the proposed changes do not
involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Section Chief: James W. Clifford.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: May 4, 2000.
Description of amendment request: The proposed amendment would add
new sections to the Technical Specifications (TSs) addressing missed
surveillance test requirements and establishing a TS Bases control
program, revise TS Chapter 6 to allow use of generic titles in lieu of
plant-specific titles, allow an alternative when the radiation
protection manager does not meet the qualifications of Regulatory Guide
1.8, relocate sections of TS Chapter 6 pertaining to onsite and offsite
review and special inspections to the Operational Quality Assurance
Plan, and correct typographical errors.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Operation of the Monticello plant in accordance with the
proposed changes does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
None of the proposed changes involves a physical modification to the
plant, a new mode of operation or a change to the Updated Safety
Analysis Report transient analysis. These proposed amendments
conform to the guidance of NUREG-1433, Revision 1, which was
previously issued by the NRC.
The proposed changes do not reduce the level of qualification or
training and the aggregate knowledge of the plant staff remains
intact. In total, these changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
because the proposed changes do not introduce a new mode of plant
operation, surveillance test requirement or involve a physical
modification to the plant. These proposed amendments generally
conform to the guidance of NUREG-1433, Revision 1, which was
previously issued by the NRC.
The proposed changes are administrative in nature. The changes
propose to relocate specifications from the Technical Specifications
to the Operational Quality Assurance Plan through which adequate
control is maintained.
The proposed changes do not alter the design, function or
operation of any plant component and therefore no new accident
scenarios are created. Therefore, the possibility of a new or
different kind of accident from any accident previously evaluated
would not be created by these amendments.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
The proposed changes do not involve a significant reduction in a
margin of safety because the current Technical Specification
requirements for safe operation of the Monticello plant are
maintained. The proposed changes are administrative in nature and do
not involve a physical modification to the plant, a new mode of
operation or a change to the Updated Safety Analysis Report
transient analyses. The proposed changes do not alter the scope of
equipment currently required to be operable or subject to
surveillance testing nor does the proposed change affect any
instrument setpoints or equipment safety functions.
Therefore, a significant reduction in the margin of safety would
not be involved with these proposed changes.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Section Chief: Claudia M. Craig.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: May 3, 2000 (PCN-516).
Description of amendment requests: The amendment application
proposes to revise the San Onofre Nuclear Generating Station, Units 2
and 3, Technical Specification (TS) 3.4.3, ``RCS Pressure and
Temperature (P/T) Limits,'' and the associated Bases. The proposed
change would reduce the minimum boltup temperature from 86 deg.F to 65
deg.F for the reactor coolant system during the period of time when
the reactor vessel head bolts are in tension.
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:
Will operation of the facility in accordance with this proposed
change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
This proposed change is a request to revise Technical
Specification 3.4.3, ``Pressure Temperature Limits.'' The proposed
change reduces the Minimum Boltup Temperature (MBT) from 86 deg.F to
65 deg.F. During operations below 86 deg.F, the plant is in a
shutdown mode, open to the atmosphere, and depressurized.
[[Page 34750]]
This proposed change does not affect the shape of the Pressure
Temperature Limits when Reactor Coolant System (RCS) temperature is
above 86 deg.F. Therefore, the probability or consequences of an
accident previously evaluated will not be increased by operating the
facility in accordance with this proposed change.
Will operation of the facility in accordance with this proposed
change create the possibility of a new or different kind of accident
from any accident previously evaluated?
Response: No.
This proposed change does not change the design or configuration
of the plant. Therefore, this proposed change will not create the
possibility of a new or different kind of accident from any accident
that has been previously evaluated.
(3) Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No.
This proposed change involves reducing the MBT from 86 deg.F to
65 deg.F. This proposed change meets the American Society of
Mechanical Engineers (ASME) Code requirements for establishing the
minimum temperature in the reactor pressure vessel flange region
when the pressure does not exceed 20% of the pre-operational
hydrostatic test pressure. All margins of safety established by the
ASME Code requirements are maintained. The operation of the facility
in accordance with this proposed change will not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1(WBN), Rhea County, Tennessee
Date of amendment request: April 10, 2000 (TS 99-013).
Description of amendment request: The proposed amendment requests
NRC's approval to use the F* alternate repair criterion in the
tubesheet region of the steam generator (SG). The F* criterion
addresses the action required when degradation has been detected in the
top of the mechanically expanded portion of SG tubes within the SG
tubesheet.
The proposed change designates a portion of the tube for which tube
degradation of a defined type does not necessitate remedial action,
except as dictated for compliance with tube leakage limits as set forth
in the WBN Technical Specifications (TS). The proposed amendment would
modify the TS which provide tube inspection requirements and acceptance
criteria to determine the level of degradation for which the tube may
remain in service. The proposed amendment would add definitions
required for the F* alternate plugging criterion and prescribe the
portion of the tube subject to the criterion.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The presence of the tubesheet enhances the tube integrity in the
region of the hardroll by precluding tube deformation beyond its
initial expanded outside diameter. The resistance to both tube
rupture and tube collapse is strengthened by the presence of the
tubesheet in that region. Hardrolling of the tube into the tubesheet
results in an interference fit between the tube and the tubesheet.
Tube rupture cannot occur because the contact between the tube and
tubesheet does not permit sufficient movement of tube material. In a
similar manner, the tubesheet does not permit sufficient movement of
tube material to permit buckling collapse of the tube during
postulated loss-of-coolant-accident (LOCA) loadings.
The type of degradation for which the alternate plugging
criterion, F*, has been developed (cracking with a circumferential
orientation) can theoretically lead to a postulated tube rupture
event, provided that the postulated through-wall circumferential
crack exists near the top of the tubesheet. An evaluation including
analysis and testing has been performed to determine the resistive
strength of roll expanded tubes within the tubesheet. That
evaluation provides the basis for the acceptance criteria for tube
degradation subject to the F* criterion.
The F* length of roll expansion is sufficient to preclude tube
pullout from tube degradation located below the F* distance,
regardless of the extent of the tube degradation. The existing
technical specification leakage rate requirements and accident
analysis assumptions remain unchanged in the unlikely event that
significant leakage from this region does occur. As noted above,
tube rupture and pullout are not expected for tubes using the F*
alternate plugging criterion. Any leakage out of the tube from
within the tubesheet at any elevation in the tubesheet is fully
bounded by the existing steam generator tube rupture analysis
included in the WBN Unit 1 Final Safety Analysis Report (FSAR). The
proposed alternate plugging criterion (F*) does not adversely impact
any other previously evaluated design basis accident.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Implementation of the proposed F* tubesheet alternate plugging
criterion does not introduce any significant changes to the plant
design basis. Use of the criterion does not provide a mechanism to
result in an accident initiated outside of the region of the
tubesheet expansion. A hypothetical accident as a result of any tube
degradation in the expanded portion of the tube would be bounded by
the existing tube rupture accident analysis. Tube bundle structural
integrity and leaktightness are expected to be maintained.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The use of the tubesheet alternate plugging criterion has been
demonstrated to maintain the integrity of the tube bundle
commensurate with the requirements of Regulatory Guide 1.121,
``Bases for Plugging Degraded PWR Steam Generator Tubes,'' for
indications in the free span of tubes and the primary to secondary
pressure boundary under normal and postulated accident conditions.
Acceptable tube degradation for the F* criterion is any degradation
indication in the tubesheet region, more than the F* distance below
either the bottom of the transition between the roll expansion and
the unexpanded tube, or the top of the tubesheet, whichever is
lower. The safety factors used in the verification of the strength
of the degraded tube are consistent with the safety factors in the
American Society of Mechanical Engineering (ASME) Boiler and
Pressure Vessel Code used in steam generator design. The F* distance
has been verified by testing to be greater than the length of roll
expansion required to preclude both tube pullout and significant
leakage during normal and postulated accident conditions. Resistance
to tube pullout is based upon the primary to secondary pressure
differential as it acts on the surface area of the tube, which
includes the tube wall cross-section, in addition to the inside
diameter-based area of the tube. The leak testing acceptance
criteria are based on the primary to secondary leakage limit in the
technical specifications and the leakage assumptions used in the
FSAR accident analyses.
Implementation of the alternate tubesheet plugging criterion
will decrease the number of tubes which must be taken out of service
with tube plugs or repaired with sleeves. Both plugs and sleeves
reduce the RCS flow margin; thus, implementation of the F* alternate
plugging criterion will maintain the margin of flow that would
otherwise be reduced in the event of increased plugging or sleeving.
Based on the above, it is concluded that the proposed change does
not result in a significant reduction in a loss of margin with
respect to plant safety as defined in the FSAR or the bases of the
WBN technical specifications.
[[Page 34751]]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
Tennessee Valley Authority (TVA), Docket No. 50-390 Watts Bar Nuclear
Plant (WBN), Unit 1, Rhea County, Tennessee
Date of amendment request: April 10, 2000 (TS 99-014).
Description of amendment request: The proposed amendment would
revise the WBN Unit 1 Technical Specification (TS) to incorporate new
requirements associated with steam generator (SG) tube inspection and
repair. The new requirements establish an alternate voltage based SG
tube repair criteria at tube support plate and Flow Distribution Baffle
plate intersections. This change is consistent with NRC Generic Letter
95-05 ``Voltage-Based Repair Criteria for Westinghouse Steam Generator
Tubes Affected By Outside Diameter Stress Corrosion Cracking.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Tube burst criteria are inherently satisfied during normal
operating conditions due to the proximity of the tube support plate.
Test data indicates that tube burst cannot occur within the tube
support plate (TSP), even for tubes which have 100 percent through-
wall electric discharge machining (EDM) notches, 0.75 inches long,
provided that the TSP is adjacent to the notched area. Since tube to
tube support plate proximity precludes tube burst during normal
operating conditions, use of the criteria must retain tube integrity
characteristics which maintain a margin of safety of 1.43 times the
bounding faulted condition [main steam line break (MSLB)]
differential pressure of 2405 psig. As previously stated, the
Regulatory Guide (RG) 1.121 criterion requiring maintenance of a
safety factor of 1.43 times the MSLB pressure differential on tube
burst is satisfied by \3/4\-inch diameter tubing with bobbin coil
indications with signal amplitudes less than
VS L = 6.03 volts, regardless of the indicated
depth measurement. At the flow distribution baffle (FDB), a safety
factor of 3 against the normal operating condition DP is applied;
here a voltage of VS L = 3.81 volts satisfies
the burst capability recommendation.
The upper voltage repair limit (VURL ) will be
determined prior to each outage using the most recently approved NRC
database to determine the tube structural limit (VSL ).
The structural limit is reduced by allowances for nondestructive
examination (NDE) uncertainty (VNDE ) and growth
(VG ) to establish VURL . As an example, the NDE
uncertainty component of 20 percent and a voltage growth allowance
of 30 percent per full power year can be utilized to establish a
VURL of 3.71 volts for TSP indications, and 2.34 volts
for the FDB indications. The 20 percent NDE uncertainty represents a
square-root-sum-of-the-squares (SRSS) combination of probe wear
uncertainty and analyst variability. The flaw growth allowance
should be an average growth rate or 30 percent per effective full
power year, whichever is larger. The 30 percent growth allowance
used to determine VURL is conservative for the current
conditions at WBN Unit 1. The most current NRC approved database,
contained in EPRI [Electric Power Research Institute] NP-7480-L,
Addendum 2, was used to establish the VURL values for the
FDB and TSP intersections. Once approved by NRC, the industry
protocol for updating the database will be followed by TVA, ensuring
that the most current database is utilized for future applications
of the criteria.
Relative to the expected leakage during accident condition
loadings, it has been previously established that a postulated MSLB
outside of containment but upstream of the main steam isolation
valves (MSIV) represents the most limiting radiological condition
relative to the alternate voltage based repair criteria. In support
of implementation of the revised repair limit, it will be determined
whether the distribution of cracking indications at the tube support
plate intersections during future cycles are projected to be such
that primary to secondary leakage would result in site boundary
doses within a fraction of the 10 CFR 100 guidelines or control room
doses within the 10 CFR 50, Appendix A, General Design Criteria
(GDC)-19 limit. A separate calculation has determined this allowable
MSLB leakage limit to be 10 gallons per minute (gpm) in the faulted
loop assuming a reactor coolant system (RCS) dose equivalent iodine
concentration of 1.0 mCi/gm. The establishment of the 10 gpm leak
rate value is controlled by the 0 to 2 hour offsite doses at the
site boundary for the accident initiated iodine spike case, not
control room dose.
The methods for calculating the radiological dose consequences
for this MSLB are consistent with FSAR Chapter 15 and therefore, the
WBN licensing basis. TVA bases the calculated thyroid dose
consequences on conversion factors from the International Commission
on Radiation Protection (ICRP) Publication 2.
In summary, the calculated radiological consequences of the
exclusion area boundary and the low population zone are larger than
previously reported for the postulated steamline break event due to
the increased leakage and more conservative iodine spiking factors.
However, the calculated radiological consequences remain in
compliance with the guidelines in the Standard Review Plan, Chapter
15, 10 CFR 50, Appendix A, GDC-19 and 10 CFR 100 reported for the
postulated steamline break event. Therefore, it is concluded that
the proposed changes do not result in a significant increase in the
radiological consequences of an accident previously analyzed.
Consistent with the guidance of Section 2.c of Generic Letter
(GL) 95-05, the WBN Unit 1 MSLB leak rate analysis performed prior
to returning the SGs to service may be performed based on the
projected next end-of-cycle (EOC) voltage distribution or the actual
measured distribution at a given outage. The method to be used for
the first outage when outside diameter stress corrosion cracking
(ODSCC) indication growth rates are available will be based on the
indications found during that outage. As noted in GL 95-05, it may
not always be practical to complete EOC calculations prior to
returning the SGs to service. Under these circumstances, it is
acceptable to use the actual measured bobbin voltage distribution
instead of the projected EOC voltage distribution to determine
whether the reporting criteria is being satisfied.
Therefore, as implementation of the 1.0 volt voltage-based
repair criteria at WBN Unit 1 does not adversely affect steam
generator tube integrity and implementation is shown to result in
acceptable radiological dose consequences, the proposed TS change
does not result in a significant increase in the probability or
consequences of an accident previously evaluated within the WBN
Final Safety Analysis Report (FSAR).
B. The proposed amendment does not create the possibility of a
new or different kind of accident from previously analyzed.
Implementation of the proposed steam generator tube alternate
voltage based repair criteria (1.0 volts) does not introduce any
significant changes to the plant design basis. Neither a single or
multiple tube rupture event would be expected in a steam generator
in which the repair limit has been applied (during all plant
conditions).
The bobbin probe voltage-based tube repair criteria of 1.0 volt
is supplemented by: enhanced eddy current inspection guidelines to
provide consistency in voltage normalization, a 100 percent eddy
current inspection sample size at the tube support plate elevations,
and rotating pancake coil (RPC) inspection requirements for the
larger indications left in service to characterize the principal
degradation as ODSCC.
TVA will implement a maximum normal operating condition primary
to secondary leakage rate limit of 600 gallons per day (gpd) total
primary to secondary leakage and 150 gpd primary to secondary
leakage per steam generator to help preclude the potential for
excessive leakage during all plant conditions. The 150 gpd leakage
limit is more restrictive than the current TS operating leakage
limit (of 500 gpd) and is intended to provide additional margin to
accommodate a stress corrosion crack which might grow at a greater
than expected rate or unexpectedly extend outside the thickness of
the tube support
[[Page 34752]]
plate. Leakage trending capability consistent with EPRI Report TR-
104788, ``Primary-to-Secondary Leak Guidelines'' has been
implemented at WBN Unit 1.
As steam generator tube integrity upon implementation of the 1.0
volt repair limit continues to be maintained through in-service
inspection and primary to secondary leakage monitoring, the
possibility of a new or different kind of accident from any accident
previously evaluated is not created.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The use of the voltage-based bobbin probe tube support plate
elevation repair criteria at WBN Unit 1 maintains steam generator
tube integrity commensurate with the criteria of Regulatory Guide
(RG) 1.121. RG 1.121 describes a method acceptable to the NRC staff
for meeting GDCs 14, 15, 31, and 32 by reducing the probability or
the consequences of steam generator tube rupture. This is
accomplished by determining the limiting conditions of degradation
of steam generator tubing, as established by in-service inspection,
for which tubes with unacceptable cracking should be removed from
service. Upon implementation of the proposed criteria, even under
the worst case conditions, the occurrence of ODSCC at the tube
support plate elevations is not expected to lead to a steam
generator tube rupture event during normal or faulted plant
conditions. The EOC distribution of crack indications at the tube
support plate elevations is confirmed to result in acceptable
primary to secondary leakage during all plant conditions and that
radiological consequences are not adversely impacted.
As a preventative measure, a total of 214 tubes are excluded
from the application of the ODSCC criteria because of the combined
effects of loss-of-coolant-accident (LOCA) + safe shutdown
earthquake (SSE) on the steam generator component (as required by
GDC 2). It was determined that tube deformation or through-wall
cracks could occur in these tubes.
As noted previously, implementation of the tube support plate
intersection voltage-based repair criteria will decrease the number
of tubes which must be repaired. The installation of steam generator
tube plugs reduces the RCS flow margin. Thus, implementation of the
1.0 volt repair limit will maintain the margin of flow that would
otherwise be reduced in the event of increased tube plugging.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
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, the Gelman Building, 2120 L
Street, NW., Washington, DC, and electronically from the ADAMS Public
Library component on the NRC Web site, http://www.nrc.gov (the
Electronic Reading Room).
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: November 23, 1999.
Brief description of amendments: The amendments changed Technical
Specification 5.5.7.c.1, ``Ventilation Filter Testing.'' The testing
criteria have been changed to be consistent with the NRC request in
Generic Letter 99-02, ``Laboratory Testing of Nuclear-Grade Activated
Charcoal.''
Date of issuance: May 16, 2000.
Effective date: May 16, 2000.
Amendment Nos.: 209 and 237.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: December 29, 1999 (64
FR 73086) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 16, 2000.
No significant hazards consideration comments received: No.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois
Date of application for amendments: December 22, 1999, as
supplemented on March 1, 2000.
Brief description of amendments: The amendments relocate Reactor
Coolant System (RCS) related cycle-specific parameter limits from the
technical specifications to, and thus expanding, the Core Operating
Limits Reports (COLRs) for Byron, Units 1 and 2, and Braidwood, Units 1
and 2.
Date of issuance: May 15, 2000.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 113 and 106.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 23, 2000 (65
FR 9003). The March 1, 2000, submittal provided additional clarifying
information that did not change the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 15, 2000.
No significant hazards consideration comments received: No.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: July 14, 1999, as supplemented
on January 21, February 15, February 23, March 10, March 24, March 31
(two letters), April 7 and April 14, 2000.
Brief description of amendments: The amendments increase the
maximum reactor core power level to 3489 megawatts thermal; an increase
of 5 percent of rated core thermal power, for
[[Page 34753]]
LaSalle County Station, Units 1 and 2. In addition, the proposed
amendments correct a non-conservative value in the upper limit for
drywell and suppression chamber internal pressure that was discovered
during the course of the power uprate review.
Date of issuance: May 9, 2000.
Effective date: For Unit 1, immediately, to be implemented within
60 days; for Unit 2, immediately, to be implemented prior to start up
of cycle 9.
Amendment Nos.: 140 and 125.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the license and Technical Specifications.
Date of initial notice in Federal Register: August 25, 1999 (64 FR
46427). The letters dated January 21, February 15, February 23, March
10, March 24, two letters on March 31, April 7, and April 14, 2000,
contain supplemental, clarifying information that did not change the
staff's initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 9, 2000.
No significant hazards consideration comments received: No.
Commonwealth Edison Company, Docket No. 50-374, LaSalle County Station,
Unit 2, LaSalle County, Illinois
Date of application for amendment: February 28, 2000, as
supplemented on April 28, 2000
Brief description of amendment: The amendment increases the
Technical Specification safety limit for the Minimum Critical Power
Ratio from 1.08 for two loop operation and 1.09 for single loop
operation to 1.11 and 1.12, respectively.
Date of issuance: May 17, 2000.
Effective date: Immediately, to be implemented within 60 days.
Amendment No.: 126.
Facility Operating License No. NPF-18: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 22, 2000 (65 FR
15377). The April 28, 2000, submittal provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 17, 2000.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: August 4, 1999, as supplemented
by letter dated April 19, 2000.
Brief description of amendments: The amendments revise the
Technical Specifications (TS) Limiting Conditions for Operation for
Reactor Coolant System (RCS) Subcooling Margin Monitor in TS Table
3.3.3-1 and revise the functions associated with surveillance
requirements for RCS Loops-Test Exceptions in TS 3.4.17. By letter
dated April 19, 2000, the licensee withdrew the proposal to relocate
the Auxiliary Feedwater Loss of Offsite Power function from TS 3.3.2-1
to TS 3.3.2-1. The other changes requested by August 4,1999,
application were addressed under separate correspondence.
Date of issuance: May 19, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 186 and 179.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 8, 1999 (64
FR 48861)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 19, 2000.
No significant hazards consideration comments received: No
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: August 18, 1999, as supplemented by
letter dated April 20, 2000.
Brief description of amendment: The requested change would revise
Technical Specification 3.5.3, ``Safety Feature Actuation System
Setpoints,'' and its associated Bases to allow for an increase to the
low reactor coolant system pressure setpoint. This setpoint change was
requested to account for additional instrument uncertainties associated
with cable insulation resistance effects and to allow for the plugging
of up to 1200 tubes in each steam generator.
Date of issuance: May 10, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 207.
Facility Operating License No. DPR-51: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 26, 2000 (65 FR
4270). The April 20, 2000, letter provided clarifying information that
did not change the scope of the August 18, 1999, application and the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 10, 2000.
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: July 15, 1999, as supplemented by
letters dated March 29, 2000, April 13, 2000, April 25, 2000, and May
9, 2000.
Brief description of amendment: The amendment changed the Technical
Specifications to institute a Technical Specification Bases Control
Program and to provide for record retention as specified in the Quality
Assurance Program Manual.
Date of issuance: May 9, 2000.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 161
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 26, 2000 (65 FR
4274). The supplements dated March 29, 2000, April 13, 2000, April 25,
2000, and May 9, 2000, did not change the scope of the initial proposed
no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 9, 2000. 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: July 19, 1999.
Brief description of amendment: The proposed change modifies
Technical Specification 4.5.2.f.2 by increasing the performance
requirement for the low pressure safety injection pumps.
Date of issuance: May 10, 2000.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No. 162.
[[Page 34754]]
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 26, 2000 (65 FR
4277).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 10, 2000.
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: July 29, 1999.
Brief description of amendment: The amendment changed the Technical
Specifications (TS) to extend the allowable outage time to seven days
for one containment spray system (CSS) train inoperable. A new ACTION
has been added to provide a shutdown requirement for the inoperability
of two CSSs. The associated changes to TS Bases are included. However,
the licensee requested MODE 4 end state for TS 3.6.2.1 is being
deferred.
Date of issuance: May 15, 2000.
Effective date: As of the date of issuance and shall be implemented
90 days from the date of issuance.
Amendment No.: 163.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 9, 2000 (65 FR
6406). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 15, 2000.
No significant hazards consideration comments received: No.
PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating
Station (LGS), Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: June 22, 1999, as supplemented
January 3, 2000.
Brief description of amendments: The amendments remove the
recirculation system motor generator set stop surveillance requirement
from the LGS Units 1 and 2 Technical Specifications.
Date of issuance: May 8, 2000.
Effective date: Both units--As of date of issuance, to be
implemented within 30 days.
Amendment Nos.: 142 and 104.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 8, 1999 (64
FR 48864). The January 3, 2000, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination or expand the scope of the original Federal
Register Notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 8, 2000.
No significant hazards consideration comments received: No
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing
(Exigent Public Announcement or Emergency Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and
electronically from the ADAMS Public Library component on the NRC Web
site, http://www.nrc.gov (the Electronic Reading Room).
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By June 14, 2000, 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
[[Page 34755]]
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 Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC and electronically from the ADAMS Public Library
component on the NRC Web site, http://www.nrc.gov (the Electronic
Reading Room). 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. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: May 8, 2000.
Description of amendment request: The amendment revises Technical
Specification Surveillance Requirement 3.8.1.9 to increase the limit
for the peak transient voltage measured following a full-load rejection
by the emergency diesel generator that is being tested.
Date of issuance: May 9, 2000.
Effective date: As of its date of issuance and shall be implemented
within 2 days.
Amendment No.: 140.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: No. The Commission's related
evaluation of the amendment, finding of emergency circumstances, and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated May 9, 2000.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan, 48226.
NRC Section Chief: Claudia M. Craig.
Dated at Rockville, Maryland, this 24th day of May 2000.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 00-13518 Filed 5-30-00; 8:45 am]
BILLING CODE 7590-01-P