[Federal Register Volume 66, Number 45 (Wednesday, March 7, 2001)]
[Notices]
[Pages 13797-13813]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-5216]
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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 February 12, 2001, through February 23,
2001. The
[[Page 13798]]
last biweekly notice was published on February 21, 2001 (66 FR 11050).
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 Administrative
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, located at One White Flint North, 11555 Rockville Pike
(first floor), Rockville, Maryland 20852. The filing of requests for a
hearing and petitions for leave to intervene is discussed below.
By April 6, 2001, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, located at One
White Flint North, 11555 Rockville Pike (first floor), Rockville,
Maryland 20852. Publicly available records will be accessible 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.
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
[[Page 13799]]
the Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Branch, or may be delivered to the Commission's Public Document Room,
located at One White Flint North, 11555 Rockville Pike (first floor),
Rockville, Maryland 20852, by the above date. A copy of the petition
should also be sent to the Office of the General Counsel, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and to the attorney
for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852.
Publicly available records will be accessible and electronically from
the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room).
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: December 20, 2000.
Description of amendments request: The amendments would revise the
Technical Specifications to incorporate changes required to support
operation with replacement steam generators. The proposed changes will
(1) accommodate geometric differences between the original and
replacement steam generators, (2) increase the reactor coolant flow
rate from the current value which was recently established to
accommodate more tube plugging, and (3) delete tube sleeving options
approved for the original steam generators.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability or
consequences of an accident previously evaluated
A. Technical Specification Table 3.3.1-1, Item 7
Technical Specification Table 3.3.1-1, ``Reactor Protective System
Instrumentation,'' Item 7 sets the allowable value for ``Steam
Generator Level-Low'' function to greater than or equal to 10 inches
below the top of the feed ring. To accommodate the geometric difference
in the location of the top of the feed ring with respect to the
pedestal between the original steam generators (OSG) (510.8 inches) and
the replacement steam generators (RSG) (484.8 inches), the proposed
amendment would change the allowable value for ``Steam Generator Level-
Low'' function to greater than or equal to 50 inches below normal water
level. Since normal water levels for RSG and OSG with respect to the
pedestal are identical and the current steam generator level-low
reactor trip setpoint ``10 inches below top of feed ring''
is `` 50 inches below normal water level'' for both the RSG
and OSG, the functionality of the steam generator level-low reactor
trip setpoint will be unchanged. Furthermore, use of normal water level
as the point of reference instead of top of the feed ring is more
practical and appropriate since it is the frame of reference for steam
generator water level indication used in the Control Room by the
operators.
The design basis accident affected by the proposed change is the
Loss of Feedwater Flow event. The Steam Generator Level-Low Reactor
Trip Setpoint, in combination with the Auxiliary Feedwater Actuation
System, ensures that adequate secondary side water inventory exists in
both RSGs to remove decay heat following a Loss of Feedwater Flow
event. To ensure that the acceptance criteria for the Loss of Feedwater
Flow event are met with the RSGs, there must be at least as much mass
in RSG at the Safety Analysis water level as in the OSG. The OSG Safety
Analysis water level is 116.4 inches below normal water level. Using
the same method to predict steam generator inventory, at this water
level, OSG has 64,049 Ibm water mass and RSG has 64,115 Ibm water mass.
Therefore, the RSG has more post-reactor trip secondary side inventory
than the OSG which ensures the Loss of Feedwater event acceptance
criteria are not challenged.
Therefore, the proposed revision to change the reference setpoint
for steam generator low level reactor trip function will not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
B. LCO [limiting condition for operation] 3.4.1 and Surveillance
Requirement 3.4.1.3
The proposed amendment would revise Technical Specification LCO
3.4.1 and Surveillance Requirement 3.4.1.3 to increase reactor coolant
minimum required total flow rate back to the originally established
value of 370,000 gpm [gallons per minute] from the current value of
340,000 gpm, which was recently established to accommodate more tube
plugging in the OSG. The flow resistance of the RSG is equivalent to
that of the OSG with zero plugged tubes. Therefore, the required
minimum RCS [reactor coolant system] total flow rate can be increased
to the value previously established for the original steam generators
with zero plugged tubes, 370,000 gpm.
Increasing the required minimum RCS total flow rate has no adverse
impact on the safety analysis. Crediting more RCS flow in the safety
analysis allows for greater flexibility in core design and operation.
The increase in RCS flow associated with the RSG is within the bounds
previously analyzed for the OSG. The hydraulic forces experienced
around the RCS loop, including the core uplift force, are acceptable.
The change is more restrictive in nature in that more RCS flow will be
required to meet Surveillance Requirement 3.4.1.3 and more RCS flow
ensures enhanced core heat removal. The overall core thermal margin in
the safety analysis will remain essentially the same.
Therefore, the proposed revision to increase reactor coolant
minimum required total flow rate will not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
C. Technical Specification Administrative Control 5.5.9
The proposed revision deletes three sleeving options from
Administrative Technical Specification 5.5.9. The sleeving options are:
Westinghouse Laser Welded sleeves, Asea Brown Boveri, Inc. (ABB)-
Combustion Engineering Leak Tight sleeves, and the ABB-Combustion
Engineering Alloy 800 Leak Limiting sleeves. One of the differences
between the OSG and the RSG design is the use of thermally-treated
Alloy 690 tube material instead of high temperature mill-annealed Alloy
600 used for the OSG. The three sleeving tube repair options described
in Calvert Cliffs Nuclear Power Plant (CCNPP) Technical Specification
[[Page 13800]]
Administrative Control 5.5.9, are designed specifically for the OSGs'
mill-annealed Alloy 600 tubes.
The three sleeving options were acquired by CCNPP for economic
reasons to maintain OSG thermal output by minimizing the number of
tubes plugged. Therefore, deletion of these repair options from
Administrative Control 5.5.9 has no safety significance.
Therefore, the proposed revision will not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Would not create the possibility of a new or different [kind] of
accident from any accident previously evaluated
A. Technical Specification Table 3.3.1-1, Item 7
The RSGs are equivalent in function to the OSGs. Changing Technical
Specification Table 3.3.1-1, Item 7 is required to provide a correct
and practical reference point from which to measure the Reactor Trip
Steam Generator Level-Low Setpoint. As described above in Item 1, the
normal water levels for RSG and OSG with respect to the pedestal are
identical and the current steam generator level-low reactor trip
setpoint, `` 10 inches below top of feed ring'' is
`` 50 inches below normal water level'' for both the RSG and
OSG. Hence, the functionality of the reactor trip steam generator
level-low setpoint will be unchanged. Furthermore, use of normal water
level as the point of reference instead of top of the feed ring is more
practical and appropriate since it is the frame of reference for steam
generator water level indication used in the Control Room by the
operators.
Therefore, the proposed revision to change the reference setpoint
for steam generator low level reactor trip function will not create the
possibility of a new or different [kind] of accident from any accident
previously evaluated.
B. LCO 3.4.1 and Surveillance Requirement 3.4.1.3
As described above in Item 1, increasing the required minimum RCS
total flow rate has no adverse impact on the plant's safety analyses.
The increase in RCS flow associated with the RSG is within the bounds
previously analyzed for the OSG. The hydraulic forces experienced
around the RCS loop, including the core uplift force, are acceptable.
The change is more restrictive in nature in that more RCS flow will be
required to meet Surveillance Requirement 3.4.1.3 and more RCS flow
ensures enhanced core heat removal.
Therefore, the proposed revision to increase reactor coolant
minimum required total flow rate will not create the possibility of a
new or different type of accident from any accident previously
evaluated.
C. Technical Specification Administrative Control 5.5.9
As described in Item I above, the three sleeving options were
acquired by CCNPP for economic reasons to maintain OSO thermal output
by minimizing the number of tubes plugged. Therefore, deletion of these
repair options from Technical Specification Administrative Control
5.5.9 has no safety significance.
Therefore, the proposed revision will not create the possibility of
a new or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in the margin of safety
A. Technical Specification Table 3.3.1-1, Item 7
As described above in Item 1, the design basis accident affected by
the proposed change is the Loss of Feedwater Flow event. The Steam
Generator Level-Low Reactor Trip Setpoint, in combination with the
Auxiliary Feedwater Actuation System, ensures that adequate secondary
side water inventory exists in both RSGs to remove decay heat following
a Loss of Feedwater Flow event. To ensure that the acceptance criteria
for the Loss of Feedwater Flow event are met with the RSGs, there must
be at least as much mass in RSG at the Safety Analysis water level as
in the OSG. The OSG Safety Analysis water level is 116.4 inches below
normal water level. Using the same method to predict steam generator
inventory, at this water level, OSO has 64,049 lbm water mass and RSG
has 64,115 Ibm water mass. Therefore, the RSG has more post-reactor
trip secondary side inventory than the OSG which ensures the Loss of
Feedwater event acceptance criteria are not challenged.
Therefore, the proposed revision to change the reference setpoint
for steam generator low level reactor trip function does not involve a
significant reduction in the margin of safety.
B. LCO 3.4.1 and Surveillance Requirement 3.4.1.3
As described above in Item 1, increasing the required minimum RCS
total flow rate has no adverse impact on the safety analysis. Crediting
more RCS flow in the safety analysis allows for greater flexibility in
core design and operation. The increase in RCS flow associated with the
RSG is within the bounds previously analyzed for the OSG. The hydraulic
forces experienced around the RCS loop, including the core uplift
force, are acceptable. The change is more restrictive in nature in that
more RCS flow will be required to meet Surveillance Requirement 3.4.1.3
and more RCS flow ensures enhanced core heat removal. The overall core
thermal margin in the safety analysis will remain essentially the same.
Therefore, the proposed revision to increase reactor coolant
minimum required total flow rate does not involve a significant
reduction in the margin of safety.
C. Technical Specification Administrative Control 5.5.9
As described in Item 1C above, the three sleeving options were
acquired by CCNPP for economic reasons to maintain OSG thermal output
by minimizing the number of tubes plugged. Therefore, deletion of these
repair options from Technical Specification Administrative Control
5.5.9 has no safety significance.
Therefore, the proposed revision does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Marsha Gamberoni.
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of amendment request: January 26, 2001.
Description of amendment request: The proposed amendment would
change Technical Specification (TS) Surveillance Requirement (SR)
3.7.9.2, ``Ultimate Heat Sink (UHS),'' by increasing the maximum
allowable temperature of Lake Michigan water from 81.5 deg.F to 85
deg.F. The licensee also proposes to reflect this change in the
associated TS Bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 13801]]
The following evaluation supports the finding that operation of the
facility in accordance with the proposed changes would not:
a. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The UHS is Lake Michigan which is completely passive and is not an
accident initiator in any accident previously evaluated. Therefore,
this change does not involve an increase in the probability of an
accident previously evaluated.
The UHS, by design, mitigates the consequences of accidents by
supplying a repository for the decay heat and other excess energy
removed in the process of cooling the plant equipment. The safety
analysis has been revised to use a maximum UHS water temperature of 85
deg.F. The results of these revised analyses still meet all of the
required acceptance criteria. Therefore, the proposed changes do not
affect any of the results of the FSAR [Final Safety Analysis Report]
Chapter 14 accident analyses. Hence the consequences of accidents
previously evaluated do not change.
Therefore, operation of the facility in accordance with the
proposed changes to the Technical Specifications would not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
b. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed change would not alter the design, configuration, or
method of operation of the plant. The proposed temperature limit has
been verified to be acceptable for UHS operability determinations by
its documented use in plant equipment design considerations, and in the
FSAR Chapter 14 accident analyses. Therefore, operation of the facility
in accordance with the proposed change to the Technical Specifications
would not create the possibility of a new or different kind of accident
from any previously evaluated.
c. Involve a significant reduction in the margin of safety.
The proposed change to the Technical Specifications would impose
temperature limits already in use in equipment designs and as an
initial assumption of the plant accident analyses. The proposed SR
limit has been utilized in the accident analyses since 1994. The
results of these accident analyses meet all of the required acceptance
criteria when using the 85 deg.F UHS water temperature limit.
Therefore, the proposed change to the Technical Specifications would
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based 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: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: Claudia M. Craig.
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of amendment request: February 12, 2001.
Description of amendment request: The proposed amendment would
change Technical Specification (TS) Section 5.6.5b, ``Reporting
Requirements--Core Operating Limits Report (COLR),'' by adding a
reference to the existing references of approved analytical methods for
determining core operating limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The following evaluation supports the finding that operation of the
facility in accordance with the proposed changes would not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to the list of methodology documents in
Specification 5.6.5.b. would not increase the probability or
consequence of an accident previously evaluated. Accidents previously
evaluated will be unaffected by the addition of a methodology reference
because they were analyzed using approved methods. The results of these
event analyses met their respective acceptance criteria.
Therefore, operation of the facility in accordance with the
proposed change to the Technical Specifications would not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
(2) Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed change to the list of methodology documents in
Specification 5.6.5.b. would not create the possibility of a new or
different accident than previously analyzed. The proposed change only
adds an approved methodology document. All accidents remain analyzed
using applicable NRC approved methodologies.
Therefore, operation of the facility in accordance with the
proposed change to the Technical Specifications would not create the
possibility of a new or different kind of accident from any previously
evaluated.
(3) Involve a significant reduction in the margin of safety.
The proposed change to the list of methodology documents in
Specification 5.6.5.b. would not reduce the margin of safety. Because
all analyses use approved methodologies and their results satisfy their
respective acceptance criteria, the margin of safety is not reduced.
Therefore, the proposed change to the Technical Specifications
would not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based 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: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
NRC Section Chief: Claudia M. Craig.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: January 24, 2001.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to incorporate the provisions
to perform routine diesel generator (DG) monthly testing by gradually
accelerating the DG to operating speed, as opposed to requiring the DG
to attain rated voltage and frequency within 10 seconds for DG 1A and
DG 1B, and within 13 seconds for DG 1C. In addition, a new TS would be
added to require fast start tests of the DGs on a 184-day frequency.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this proposed
change involve a significant increase in the probability or
consequences of an accident previously evaluated?
[[Page 13802]]
The proposed changes affect the surveillance requirements for the
emergency diesel generators. The emergency diesel generators are onsite
standby power sources intended to provide redundant and reliable power
to ESF [engineered safety feature] systems credited as accident
mitigating features in design basis analyses. As discussed in
Regulatory Guide (RG) 1.9, Revision 3, the proposed changes are
intended to allow slower starts of the diesel generators during testing
in order to reduce diesel generator aging effects due to excessive
testing conditions. As such, the proposed changes should result in
improved diesel generator reliability and availability, thereby
providing additional assurance that the diesel generators will be
capable of performing their safety function. The method of starting the
emergency diesel generators for testing purposes does not affect the
probability of any previously evaluated accident. Although the changes
allow slower starts for the monthly tests, the more rapid start
function assumed in the accident analysis is unchanged and will be
verified on a 184 day frequency. Therefore the accident analysis
consequences are not affected.
Therefore, these changes do not involve a significant increase in
the probability or consequences of any accident previously evaluated.
2. 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?
The proposed changes affect the surveillance requirements for the
onsite ac [alternating current] sources, i.e. the diesel generators.
Accordingly, the proposed changes do not involve any change to the
configuration or method of operation of any plant equipment that could
cause an accident. In addition, no new failure modes have been created
nor has any new limiting failure been introduced as a result of the
proposed surveillance changes.
Therefore, these changes do not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Will operation of the facility in accordance with this proposed
change involve a significant reduction in a margin of safety?
The proposed changes are intended to bring the existing RBS [River
Bend Station] TS requirements for the onsite ac sources in line with
regulatory guidance. Under the proposed changes, the emergency diesel
generators will remain capable of performing their safety function, and
the effects of aging on the diesel generators will be reduced by
eliminating unnecessary testing. The diesel generator start times
assumed in the current accident analyses are unchanged and will be
verified on a 6-month frequency.
Therefore, these changes do not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.
Entergy Nuclear Generation Company, Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: February 5, 2001.
Description of amendment request: The proposed amendment would
change the Safety Limit Minimum Critical Power Ratio (SLMCPR) in
Technical Specification (TS) 2.1.2 from 1.08 to 1.06. The proposed
amendment would also change the parenthetical statements after certain
references listed in TS 5.6.5.b to clarify that the analytical methods
described in General Electric Nuclear Energy documents inclusive of the
latest amendment or revision are used to determine core operating
limits. Also, the proposed amendment would add a new reference to TS
5.6.5.b.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is
presented below:
1. The proposed changes to technical specification do not involve a
significant increase in the probability of an accident previously
evaluated.
The proposed Safety Limit MCPR (SLMCPR), and its use to determine
the Cycle 14 thermal limits, have been derived using NRC approved
methods [See application dated February 5, 2001]. These methods do not
change the method of operating the plant and have no effect on the
probability of an accident initiating event or transient.
The basis of the SLMCPR is to ensure no mechanistic fuel damage is
calculated to occur if the limit is not violated. The new SLMCPR
preserves the margin to transition boiling, and the probability of fuel
damage is not increased.
Therefore, the proposed changes to technical specifications do not
involve an increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes to technical specifications do not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
The proposed changes result only from revised methods of analysis
for the Cycle 14 core reload. These methods have been reviewed and
approved by the NRC, do not involve any new or unapproved method for
operating the facility, and do not involve any facility modifications.
No new initiating events or transients result from these changes.
Therefore, the proposed changes to technical specifications do not
create the possibility of a new or different kind of accident from any
accident previously evaluated.
3. The proposed changes to technical specifications do not involve
a significant reduction in a margin of safety.
The margin of safety will remain the same. The new SLMCPR was
derived using NRC approved methods which are in accordance with the
current fuel design and licensing criteria. The SLMCPR remains high
enough to ensure that greater than 99.9% of all fuel rods in the core
will avoid transition boiling if the limit is not violated, which is
the current margin of safety used to preserve the fuel cladding
integrity.
Therefore, the proposed changes to technical specifications do not
involve a significant reduction in the 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: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts, 02360-5599
NRC Section Chief: James W. Clifford.
Entergy Nuclear Generation Company, Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: February 16, 2001.
Description of amendment request: This amendment would substitute a
surveillance interval of ``Once/
[[Page 13803]]
Operating Cycle'' for the current surveillance interval of ``Each
Refueling Outage,'' for the following instruments in Technical
Specification Table 4.2.F: Containment High Radiation Monitor, Reactor
Building Vent Radiation Monitor, Main Stack Vent Radiation Monitor, and
Turbine Building Vent Radiation Monitor.
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 changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
There are no physical changes to Pilgrim being introduced by the
proposed changes to the specified instruments. The proposed changes do
not modify Pilgrim, i.e., there are no changes in operating pressure,
materials or seismic loading. No plant safety limits, setpoints, or
design parameters are adversely affected by the proposed changes. The
proposed changes do not adversely affect the integrity of the reactor
coolant pressure boundary such that its function in the control of
radiological consequences is affected. The proposed changes do enlarge
the opportunity-period for performing the subject calibrations by
substituting one established Technical Specification definition for
another; hence, the proposed changes are administrative in nature
because they do not change any methodology, interval, configuration or
equipment at Pilgrim.
Thus, the proposed changes do not affect any significant parameter
associated with the instruments or calibration interval; therefore, the
ability of the instruments to perform their designed safety function is
maintained. The change does not impact plant operation. Consequently,
operating Pilgrim in conformance with the proposed changes does not
involve a significant increase in the probability or consequences of an
accident previously evaluated.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change substitutes one Technical Specification
definition for another concerning certain radiation-monitoring
instruments. The ability of these instruments to perform their
designed-function is not affected by this change, and the surveillance
interval remains nominally 24 months. No new modes of operation are
introduced by the proposed changes. No plant safety limits, setpoints,
or design parameters are herein proposed, nor is any adverse
consequence introduced by the proposed changes. The proposed changes
will not create any failure mode not bounded by previously evaluated
accidents. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed changes do not involve a significant reduction in a
margin of safety.
The proposed changes entail the substitution of one Technical
Specification definition for another concerning radiation-monitoring
instruments. This is an administrative change because such substitution
does not modify the operation, configuration, or processes of Pilgrim,
nor does the change modify the nominal 24-month surveillance/
calibration interval currently in force for these instruments.
The substitution of one Technical Specification definition for
another concerning radiation monitoring instruments potentially reduces
personnel exposure from calibration-source radiation because site
population is less during non-refueling periods. No plant safety
limits, setpoints, or design parameters are changed, nor is any adverse
consequence introduced by the proposed changes. Therefore, the proposed
changes do not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts, 02360-5599
NRC Section Chief: James W. Clifford.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: February 6, 2001
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) associated with the reactor
coolant system (RCS) leakage detection systems, to make them consistent
with the requirements in NUREG-1432, ``Standard Technical
Specifications, Combustion Engineering Plants.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1--Does Not Involve a Significant Increase in the Probability
or Consequences of an Accident Previously Evaluated
The aforementioned revisions do not involve any physical change to
plant design. Relocating the requirements associated with the RCS Leak
Detection System from various TSs to ANO-2 [Arkansas Nuclear One, Unit
2] Specification 3.4.6.1 is administrative in nature and does not
affect the accident analyses. The RCS water inventory balance is more
accurate than normal leak detection methods in regard to actual RCS
leak rates, and therefore is an excellent alternative when other leak
detection components may become inoperable. Since the proposed changes
only affect the requirements for the detection of RCS leakage, the
probability that an accident previously evaluated will occur remains
unchanged. The proposed changes do not prevent nor limit the diversity
of acceptable detection of RCS leakage and, therefore, do not
significantly affect the consequences of an accident previously
evaluated since leak rate information will remain available to station
personnel. Although the non-administrative revisions result in less
restrictive requirements, the proposed changes remain within the
acceptability of General Design Criteria (GDC) 30 of Appendix A to 10
CFR [Part] 50 and Regulatory Guide (RG) 1.45, and are consistent with
the philosophies of the RSTS [Revised Standard Technical
Specifications].
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any accident previously
evaluated.
Criterion 2--Does Not Create the Possibility of a New or Different Kind
of Accident From Any Previously Evaluated
The aforementioned revisions do not involve any physical change to
plant design. Relocating the requirements associated with the RCS Leak
Detection System from various TSs to ANO-2 Specification 3.4.6.1 is
administrative in nature and does not affect the accident analyses. The
RCS water inventory balance is more accurate than normal leak detection
methods in regard to actual RCS leak rates, and therefore is an
excellent alternative when other leak
[[Page 13804]]
detection components may become inoperable. The proposed changes do not
prevent acceptable detection of RCS leakage by diverse methods. The
detection of a RCS leak does not cause an accident or prevent an
accident from occurring. Likewise, detecting a RCS leak while in its
initial stages does not create the possibility of a new or different
kind of accident than any previously analyzed. Therefore, a new or
different kind of accident than that previously analyzed is not
expected to result due to the proposed changes of this submittal.
Although the non-administrative revisions result in less restrictive
requirements, the proposed changes remain within the acceptability of
General Design Criteria (GDC) 30 of Appendix A to 10 CFR [Part] 50,
Regulatory Guide (RG) 1.45, and are consistent with the philosophies of
the RSTS.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
Criterion 3--Does Not Involve a Significant Reduction in the Margin of
Safety
The aforementioned revisions do not involve any physical change to
plant design. Relocating the requirements associated with the RCS Leak
Detection System from various TSs to the ANO-2 Specification 3.4.6.1 is
administrative in nature and does not affect the margin of safety. The
RCS water inventory balance is more accurate than normal leak detection
methods in regard to actual RCS leak rates, and therefore is an
excellent alternative when other leak detection components may become
inoperable. Maintaining diverse and accurate RCS leak detection methods
available helps to ensure RCS leaks will be detected within an
acceptable period of time and, therefore, the proposed changes do not
significantly reduce the margin to safety. Although the non-
administrative revisions result in less restrictive requirements, the
proposed changes remain within the acceptability of General Design
Criteria (GDC) 30 of Appendix A to 10 CFR [Part] 50 and Regulatory
Guide (RG) 1.45, and are consistent with the philosophies of the RSTS.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant,
Unit No. 1, St. Lucie County, Florida
Date of amendment request: January 17, 2001
Description of amendment request: The licensee proposes to revise
the Technical Specifications (TS) requirements for the Emergency Diesel
Generator (EDG) 24-hour surveillance test run. Currently, the TS
restrict performance of this test to shutdown periods due to historical
concerns regarding the effects of a potential failure while the EDGs
are paralleled to the off-site power system. The proposed amendment
would allow the surveillance test to be conducted with the plant on-
line. The licensee has performed an analysis, which shows that
conducting the 24-hour EDG test run with the plant on-line results in a
very small change in core damage frequency, and is acceptable under the
guidelines of Regulatory Guide 1.174. The risks incurred by performing
the test on-line will be substantially offset by plant benefits
associated with avoiding unnecessary plant transitions and/or reducing
risks during shutdown operations.
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 a significant increase in
the probability or consequences of an accident previously evaluated for
the following reasons:
The change relocating the ``during shutdown'' requirement from TS
4.8.1.1.2.e to the individual surveillance requirements under TS
4.8.1.1.2.e is strictly administrative in nature. Therefore, it does
not involve any increase in the probability or consequences of an
accident previously evaluated.
For the change that revises Unit 1 TS 4.8.1.1.2.e.6 to remove the
restriction to perform the EDG 24-hour endurance test during shutdown,
the emergency diesel generators (EDG) and their associated emergency
busses are not accident initiating equipment. Therefore, there will be
no impact on any accident probabilities by the approval of this
amendment. The design of this equipment is not being modified by these
proposed changes. In addition, the ability of the EDGs to respond to a
design basis accident will not be significantly impacted by these
proposed changes. Consequences are no different than presently when an
EDG is out-of-service in the current TS allowed outage time during
operation in Modes 1 and 2.
Therefore, performing the EDG 24-hour endurance test in Modes 1 and
2 does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
(2) Use of the modified specification would not create the
possibility of a new or different kind of accident from any previously
evaluated.
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated for
the following reasons:
No new accident causal mechanisms are created as a result of this
amendment request. Equipment will be operated in the same configuration
with the exception of the plant Mode in which testing is conducted. No
changes are being made to the plant which introduce any new accident
causal mechanisms. This amendment request does not impact any plant
systems that are accident initiators; neither does it adversely impact
accident mitigating systems.
The changes removing the restriction to perform the tests during
shutdown for Unit 1 TS 4.8.1.1.2.e.6, in its simplest form, is just a
request to extend the amount of time the EDG is synchronized to the
grid in Modes 1 and 2 from approximately 18 hours (one hour per month)
to approximately 42 hours per cycle. The existing surveillance
requirement TS 4.8.1.1.2.a.5 requires, in part, that every 31 days each
EDG be demonstrated operable by synchronizing to the grid for at least
an hour. It is simply a time extension of the existing surveillance
requirement. Therefore, performing the EDG 24-hour endurance test in
Modes 1 and 2 does not create the possibility of a new or different
kind of accident from any previously evaluated.
(3) Use of the modified specification would not involve a
significant reduction in a margin safety.
The AC electrical distribution system has been designed to provide
sufficient
[[Page 13805]]
redundancy and reliability to ensure the availability of the EDGs to
provide the required safety function under design basis events to
protect the power plant, the public, and plant personnel.
The proposed changes do not affect the limiting conditions for
operation or their bases that are used in the deterministic analysis to
establish any margin of safety. PSA evaluations were used to evaluate
these changes, and these evaluations determined that the changes are
not risk significant. The proposed activity involves changes to the
allowed plant mode for the performance specific Technical Specification
surveillance requirements.
During the performance of the EDG endurance surveillance test for a
24-hour period, at least one EDG will be available and will adequately
respond within the time necessary to mitigate anticipated operational
occurrences or postulated design basis accidents.
The calculated total change in CDF, including the conservatively
estimated fire risk contribution, is less than 1E-06 per reactor year
and the calculated total change in the LERF, including the
conservatively estimated fire risk contribution, is less than 1E-07 per
reactor year. The change in CDF and LERF is, therefore, within Region
III of Regulatory Guide 1.174 Figures 3 and 4, and is considered very
small. When the full scope of plant risk is considered, the risks
incurred by performing the EDG 24-hour surveillance test during power
operation will be substantially offset by plant benefits associated
with avoiding unnecessary plant transitions and/or reducing risks
during shutdown operations.
The proposed change does not involve a change to the plant design
or operation, and thus, does not affect the design of the EDGs, the
operational characteristics of the EDGs, the interfaces between the
EDGs and other plant systems, or the function or reliability of the
EDGs. Because EDG performance and reliability will continue to be
ensured by the proposed Technical Specification changes, the proposed
changes do not result in a significant reduction of the margin of
safety.
Based on the above, FPL has determined that the proposed amendment
does not involve a significant increase in the probability or
consequences of an accident previously evaluated; or create the
possibility of a new or different kind of accident from any accident
previously evaluated; or involve a significant reduction in a margin of
safety; and therefore, does not involve a significant hazards
consideration.
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.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of amendment request: January 4, 2001.
Description of amendment request: The proposed amendment requests
NRC's approval of the Maine Yankee Atomic Power Company's (MYAPC)
Security Plan, Training and Qualification Plan, and Contingency Plan.
These plans reflect the addition of provisions related to the loading
and storage of spent fuel into the independent spent fuel storage
installation (ISFSI) under construction on owner-controlled property
adjacent to the plant site.
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 approved Security Plan, or Defueled Security Program, currently
implemented is not being changed. The FIT [Fuel in Transit] Security
Program and the ISFSI Security Program are being added to the scope of
the overall security scheme at the Maine Yankee site. The additions to
the overall plan have been evaluated in accordance with 10 CFR 50.54(p)
and 10 CFR 72.212(b)(4) and it has been determined that the
implementation of the ISFSI and FIT Security Programs would not
decrease the effectiveness of the Defueled Security Program, the
Defueled Security Guard Training and Qualification Program, or the
first four categories of the Defueled Safeguards Contingency Program.
The Defueled Security Program Staffing will be augmented as and if
necessary to support Fuel in Transit evolutions. The ISFSI Security
Program staffing will be separate from and parallel to the staffing
requirements of the Defueled Security Program.
The operational and physical venues of the Defueled Security
Program, the FIT Security Program, and the ISFSI Security Program are
separate and distinct. The line of demarcation between the three
programs is clearly defined and not overlapping. The implementation of
any of the programs therefore does not degrade or inhibit the
implementation of the other two programs.
The Defueled Program Guard Training and Qualification Plan and the
Defueled Safeguards Contingency plan also have not been changed. A
separate and parallel ISFSI Training and Qualification Plan and
Contingency Plan is included in the ISFSI Security Program. The FIT
program uses the Defueled Program, Training and Qualification Plan and
Contingency Plan. The physical protection systems described in the
ISFSI and FIT Programs are designed to protect against the loss of
control of the facility that could be sufficient to cause a radiation
exposure exceeding the dose as described in 10 CFR 72.106.
Therefore, the ISFSI Program revisions of the Security Plan, Guard
Training and Qualification Plan and the Safeguards Contingency Plan
will not increase the probability or the consequences of an accident
previously evaluated since the previously approved Defueled Training
and Qualification Plan and Contingency Plan remain unchanged.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The FIT and ISFSI Security Programs have no impact on the existing
Defueled Security Program since they operate in different physical and
licensing venues. The accidents considered for the Spent Fuel Pool, the
venue of the Defueled Security Program, are described in the Maine
Yankee Defueled Safety Analysis Report. The accidents considered for
the FIT and ISFSI are contained in the NAC International, Inc. Final
Safety Analysis Report for the UMS Universal Storage System Docket No.
72-1015.
The FIT and ISFSI Security Programs have been crafted to meet or
exceed all of the assumptions of the NAC International FSAR concerning
accident analyses and the programs meet or exceed all of the applicable
requirements of 10 CFR 73.55 with approved exceptions or approved
alternative measures. The physical protection systems described in the
ISFSI and FIT Programs are designed to protect against the loss of
control of the
[[Page 13806]]
facility that could be sufficient to cause a radiation exposure
exceeding the dose as described in 10 CFR 72.106.
The proposed action does not affect plant systems, structures or
components within the venue of the existing Security Plan. The ISFSI
and FIT program additions to the Security Plan, Guard Training and
Qualification Plan and the Safeguards Contingency Plan do not create
the possibility of a new or different kind of accident from any
accident previously evaluated since the previously approved Defueled
Security Plan, Training and Qualification Plan and Contingency plan
remain as is, unaltered.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The addition of a separate, parallel ISFSI and FIT Safeguards
Program, Training and Qualification Plan, and Contingency Plan does not
alter or reduce the effectiveness of the previously approved Defueled
Program. The physical protection systems described in the ISFSI and FIT
Programs are designed to protect against the loss of control of the
facility that could be sufficient to cause a radiation exposure
exceeding the dose as described in 10 CFR 72.106. Therefore, the margin
of safety will not be reduced as a result of the ISFSI and FIT
additions to the Security Plan, or an ISFSI specific addition of a
Guard Training and Qualification Plan or an ISFSI specific addition of
a Safeguards Contingency Plan
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendment involves no significant hazards consideration.
Attorney for licensee: Joseph Fay, Esquire, Maine Yankee Atomic
Power Company, 321 Old Ferry Road, Wiscasset, Maine 04578.
NRC Section Chief: Michael T. Masnik.
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment requests: April 17, 2000, as supplemented
February 2, 2001.
Description of amendment requests: The proposed amendments would
change the Technical Specifications (TSs) for the removal of boric acid
storage tanks (BASTs) from the safety injection (SI) system. These
changes would accomplish two objectives: (1) Eliminate high
concentration boric acid from the SI system and (2) align this specific
Prairie Island TS section with the Standard TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed amendment will not involve a significant increase
in the probability or consequences of accidents previously evaluated.
The proposed change to the CVCS [chemical volume control system]
and SI system (increasing the concentration of boric acid in the RWST
[refueling water storage tank] and eliminating the BAST as a suction
source, respectively) and elimination of or change to associated
Technical Specifications do not affect accident initiation. None of the
equipment being removed from Sections 3.2 or 3.5 of Technical
Specifications are accident initiators. Thus, the proposed changes will
not significantly increase the probability of an accident previously
evaluated.
Consequences are evaluated in terms of off-site and on-site
(control room personnel) dose. Loss of coolant accident (LOCA) dose is
unaffected by the proposed changes because the LOCA analysis input
assumptions are not changed by the changes proposed in this amendment
request. The approved steam line break (SLB) methodology (approved by
the NRC in letter dated January 19, 2000) and the expected dose are
unaffected by the proposed change.
Therefore, the proposed changes will not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
(2) The proposed amendment will not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The proposed changes to the plant and its Technical Specifications
do not introduce any new accident initiators. The proposed changes
reduce the number of automatic component actuations needed to support
Safety Injection accident mitigation functions. The proposed changes
also remove the Technical Specification requirements for the balance of
the CVCS components. These requirements were in Technical
Specifications to support the boration function of CVCS; however, all
boration functions can be met by the safety-related SI system. All the
other functions of the CVCS are either backed up by a safety related
system or are not required to preclude an accident (reference NSP
[Northern States Power] letter of June 14, 1995 and NRC letter of
January 8, 1996).
Therefore, the proposed changes will not create the possibility of
a new or different kind of accident.
(3) The proposed amendment will not involve a significant reduction
in the margin of safety.
The proposed changes do not significantly impact the plant response
to an accident with respect to the ability to protect fission product
barriers. The proposed changes will not result in any significant
increase in fuel cladding damage in the event of a postulated accident
(accident analyses show the proposed changes meet all acceptance
criteria related to maintaining cladding integrity). The proposed
changes will not reduce the integrity of the RCS [reactor coolant
system] (reduction of boric acid concentrations in the SI systems will
not promote any degradation of the components that make up the RCS
pressure boundary). The proposed changes will not result in a reduction
in containment integrity in the event of a postulated accident (the
changes proposed by this amendment do not change the results of the
accident analyses with respect to containment response.)
Therefore, the proposed changes will not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Claudia M. Craig.
Sacramento Municipal Utility District (SMUD), Docket No. 50-312, Rancho
Seco Nuclear Station, Sacramento County, California
Date of amendment request: October 23, 2000.
Description of amendment request: The proposed amendment (PA-194)
as supplemented by SMUD letter to the USNRC dated January 11, 2001,
would change the Permanently Defueled Technical Specification (PDTS) by
deleting the definitions for ``site boundary'' and ``unrestricted
area;'' revising the definition of the ``site;'' deleting figures D5.1-
1, ``Emergency Planning Zone,'' D5.1-2, ``Site Boundary for Gaseous
Effluent,'' and D5.1-3, ``Site Boundary for Liquid Effluent;'' and
making editorial changes
[[Page 13807]]
to the other PDTSs because of the above proposed changes. The
information proposed for removal from the PDTS is contained in or will
be relocated to other licensee-controlled documents.
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:
SMUD has reviewed the proposed PDTS change against each of the
criteria in 10 CFR 50.92 and has concluded that the amendment request
involves no significant hazards consideration. The following provides
SMUD's analysis of the issue of no significant hazards consideration:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident previously
evaluated?
No. The proposed changes are administrative and involve deleting
the definitions of SITE BOUNDARY and UNRESTRICTED AREA from the
DEFINITIONS section, revising the definition of the site in Section 5.1
``SITE,'' deleting all three figures from the DESIGN FEATURES section
[SMUD proposes, as described in its January 11, 2001, letter, that
these or equivalent figures will be relocated to either the Emergency
Plan or the Offsite Dose Calculation Manual, as appropriate], revising
Sections D6.8.3.a(2) and D6.8.3.a(4) so that the term ``unrestricted
area'' is lower case, and revising Sections D6.8.3.a(8), D6.8.3.a(9),
D6.8.3.a(10), and D6.8.3.b(2) so that the term ``site boundary'' is
lower case.
These changes do not affect possible initiating events for
accidents previously evaluated or alter the configuration or operation
of the facility. Safety limits, limiting safety system settings, and
limiting control systems are no longer applicable to Rancho Seco
Technical Specifications in the permanently defueled mode, and are
therefore not relevant.
The proposed changes do not affect the emergency planning zone, the
boundaries used to evaluate compliance with liquid or gaseous effluent
limits, and have no impact on plant operations. Therefore, the proposed
license amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed license amendment create the possibility of a
new or different kind of accident from any accident previously
evaluated?
No. As described above, the proposed changes are administrative.
The safety analysis for the facility remains complete and accurate.
There are no physical changes to the facility and the plant conditions
for which the design basis accidents have been evaluated are still
valid.
The operating procedures and emergency procedures are not affected.
The proposed changes do not affect the emergency planning zone, the
boundaries used to evaluate compliance with liquid or gaseous effluent
limits, and have no impact on plant operations. Consequently, no new
failure modes are introduced as the result of the proposed changes.
Therefore, the proposed changes will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed license amendment involve a significant
reduction in a margin of safety?
No. As described above, the proposed changes are administrative.
There are no changes to the design or operation of the facility. The
proposed changes do not affect the emergency planning zone, the
boundaries used to evaluate compliance with liquid or gaseous effluent
release limits, and have no impact on plant operations. Accordingly,
neither the design basis nor the accident assumptions in the Defueled
Safety Analysis Report (DSAR), nor the Technical Specification Bases
are affected. Therefore, the proposed changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendment involves no significant hazards consideration.
Attorney for licensee: Dana Appling, Esq., Sacramento Municipal
Utility District, P.O. Box 15830, Sacramento, California 95852-1830.
NRC Section Chief: Michael T. Masnik.
Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364,
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama
Date of amendment request: August 25, 2000.
Description of amendment request: The proposed amendments would
revise the Updated Final Safety Analysis Report (UFSAR) described
offsite dose analyses based on changes to the letdown flow rate and
iodine spike postulated concurrent with a Main Steam Line Break or a
Steam Generator Tube Rupture.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes do not significantly increase the probability
or consequences of an accident previously evaluated in the UFSAR. The
comprehensive engineering review included evaluations or re-analysis of
all accident analyses. Calculations for letdown flow measurement and
indication have verified the acceptability of the analyzed letdown flow
rate. The letdown flow rate does not initiate any accident; therefore,
the probability of an accident has not been increased. All dose
consequences have been analyzed or evaluated with respect to the
proposed changes, and all acceptance criteria continue to be met.
Therefore, these changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously analyzed?
The proposed changes do not create the possibility of a new or
different kind of accident than any accident already evaluated in the
UFSAR. No new accident scenarios, failure mechanisms or limiting single
failures are introduced as a result of the proposed changes. The
changes have no adverse effects on any safety-related system and do not
challenge the performance or integrity of any safety-related system.
Therefore, all accident analyses criteria continue to be met and these
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin of
safety?
The proposed changes do not involve a significant reduction in a
margin of safety. All analyses and evaluations using letdown flow rate
as an input have been revised to reflect the proposed value. The
calculations are based on FNP instrumentation and test methods and
include uncertainty allowances. The evaluations and analyses results [a
small change] demonstrate applicable acceptance criteria are met.
Therefore, the proposed
[[Page 13808]]
changes do not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Section Chief (Acting): Maitri Banerjee.
Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364,
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama
Date of amendment request: December 8, 2000.
Description of amendment request: The proposed amendments would
either delete or modify existing license conditions from the Unit 1 and
Unit 2 Operating Licenses, which have been completed or are otherwise
no longer in effect. These activities have now been completed, and the
license conditions are either obsolete or no longer needed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed amendment deletes license conditions which are
completed or are otherwise obsolete. As such, the change is strictly
administrative. Therefore, this change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously analyzed?
The proposed amendment deals with operating license reporting
conditions and has no effect on the type of accidents that have been
considered at Plant Farley. Therefore, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the change involve a significant reduction in a margin of
safety?
The requirements associated with the deleted license conditions
have been completed; the conditions are therefore obsolete. Removing
these conditions from the license is an administrative and editorial
activity. Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Section Chief (Acting): M. Banerjee.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: January 18, 2001 (ULNRC-04371).
Description of amendment request: The proposed amendment deletes
Section 5.5.3, ``Post Accident Sampling,'' from the administrative
controls section of the Technical Specifications (TS). The proposed
amendment deletes requirements from the TS (and, as applicable, other
elements of the licensing bases) to maintain a Post Accident Sampling
System (PASS). Licensees were generally required to implement PASS
upgrades as described in NUREG-0737, ``Clarification of TMI [Three Mile
Island] Action Plan Requirements,'' and Regulatory Guide 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI Unit 2. Requirements related to
PASS were imposed by Order for many facilities and were added to or
included in the TS for nuclear power reactors currently licensed to
operate. Lessons learned and improvements implemented over the last 20
years have shown that the information obtained from PASS can be readily
obtained through other means or is of little use in the assessment and
mitigation of accident conditions.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 11, 2000 (65 FR 49271) on possible
amendments to eliminate PASS, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on October 31, 2000 (65 FR 65018). The licensee affirmed the
applicability of the following NSHC determination in its application
dated January 18, 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated.
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result of
the TMI-2 accident. The specific intent of the PASS was to provide a
system that has the capability to obtain and analyze samples of plant
fluids containing potentially high levels of radioactivity, without
exceeding plant personnel radiation exposure limits. Analytical results
of these samples would be used largely for verification purposes in
aiding the plant staff in assessing the extent of core damage and
subsequent offsite radiological dose projections. The system was not
intended to and does not serve as a function for preventing accidents
and its elimination would not affect the probability of accidents
previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual benefit
to post accident mitigation. Past experience has indicated that there
exists in-plant instrumentation and methodologies available in lieu of
a PASS for collecting and assimilating information needed to assess
core damage following an accident. Furthermore, the implementation of
Severe Accident Management Guidance (SAMG) emphasizes accident
management strategies based on in-plant instruments. These strategies
provide guidance to the plant staff for mitigation and recovery from a
severe accident. Based on current severe accident management strategies
and guidelines, it is determined that the PASS provides little benefit
to the plant staff in coping with an accident.
The regulatory requirements for the PASS can be eliminated without
[[Page 13809]]
degrading the plant emergency response. The emergency response, in this
sense, refers to the methodologies used in ascertaining the condition
of the reactor core, mitigating the consequences of an accident,
assessing and projecting offsite releases of radioactivity, and
establishing protective action recommendations to be communicated to
offsite authorities. The elimination of the PASS will not prevent an
accident management strategy that meets the initial intent of the post-
TMI-2 accident guidance through the use of the SAMGs, the emergency
plan (EP), the emergency operating procedures (EOP), and site survey
monitoring that support modification of emergency plan protective
action recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident from any Previously Evaluated
The elimination of PASS related requirements will not result in any
failure mode not previously analyzed. The PASS was intended to allow
for verification of the extent of reactor core damage and also to
provide an input to offsite dose projection calculations. The PASS is
not considered an accident precursor, nor does its existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within the
containment building.
Therefore, this change does not create the possibility of a new or
different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
The elimination of the PASS, in light of existing plant equipment,
instrumentation, procedures, and programs that provide effective
mitigation of and recovery from reactor accidents, results in a neutral
impact to the margin of safety. Methodologies that are not reliant on
PASS are designed to provide rapid assessment of current reactor core
conditions and the direction of degradation while effectively
responding to the event in order to mitigate the consequences of the
accident. The use of a PASS is redundant and does not provide quick
recognition of core events or rapid response to events in progress. The
intent of the requirements established as a result of the TMI-2
accident can be adequately met without reliance on a PASS.
Therefore, this change does not involve a significant reduction in
the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852.
Publicly available records will be accessible and electronically from
the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room).
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: September 14, 2000.
Brief description of amendments: The amendments add two analytical
methods to the list of approved core operating limit analytical methods
in Technical Specification 5.6.5.b.
Date of issuance: February 8, 2001.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 241 and 215.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 18, 2000 (65 FR
62383).
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated February 8, 2001.
No significant hazards consideration comments received: No.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: November 22, 1999, as
supplemented on September 11, 2000.
Brief description of amendment: The amendment revises Technical
Specification Sections 4.5.D, ``Containment Air Filtration System,''
4.5.E, ``Control Room Air Filtration System,'' 4.5.F, ``Fuel Storage
Building Air Filtration System,'' and 4.5.G, ``Post-Accident
Containment Venting System,'' to address the testing requirements in
Generic Letter 99-02, ``Laboratory Testing of Nuclear-Grade Activated
Charcoal.'' The laboratory testing of the engineered safeguards
features ventilation system charcoal samples will meet the requirements
of the American Society for Testing and Materials Standard D3803-1989.
Date of issuance: February 21, 2001.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 215.
[[Page 13810]]
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 15, 2000 (65
FR 69059).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 21, 2001.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: October 30, 2000.
Brief description of amendment: The amendment revises Surveillance
Requirement 3.6.1.3.8 to allow a representative sample of reactor
instrument line excess flow check valves (EFCVs) to be tested every 24
months such that each reactor instrument EFCV will be tested at least
once every 10 years. The amendment also limits the surveillance
requirement to only the reactor instrument line EFCVs.
Date of issuance: February 20, 2001.
Effective date: February 20, 2001, and shall be implemented within
30 days from the date of issuance.
Amendment No.: 170.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 29, 2000 (65
FR 71135).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 20, 2001.
No significant hazards consideration comments received: No.
Entergy Nuclear Generation Company, Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: November 22, 1999, as
supplemented on November 21, 2000.
Brief description of amendment: This amendment approves changes
related to Technical Specification (TS) Sections 3.7.B.1 and 3.7.B.2,
``Containment Systems.'' TS Section 5.0, ``Administrative Controls,''
was also modified to reflect the addition of an omitted page from a
previous amendment.
Date of issuance: February 13, 2001.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 187.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 5, 2000 (65 FR
17913).
The November 21, 2000, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated February 13, 2001.
No significant hazards consideration comments received: No.
Exelon Generation 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: February 15, 2000, as
supplemented on July 26, 2000. The July 26, 2000, letter provided
clarifying information that did not change the scope of the February
15, 2000, application or the initial proposed no significant hazards
consideration determination.
Brief description of amendments: The amendments allow the use of
the Westinghouse core monitoring system know as Best Estimate Analyzer
for Core Operations Nuclear.
Date of issuance: February 13, 2001.
Effective date: February 13, 2001.
Amendment Nos.: 116, 116, 110, and 110.
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: April 5, 2000 (65 FR
17909).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 13, 2001.
No significant hazards consideration comments received: No.
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point
Nuclear Station, Unit 1, Oswego County, New York
Date of application for amendment: September 26, 2000.
Brief description of amendment: The amendment changes the Technical
Specifications to (1) allow reactor vessel hydrostatic tests, leakage
tests, scram time tests and excess flow check valve tests be performed;
(2) require containment building integrity be maintained; and (3)
establish a limit and a surveillance requirement on reactor coolant
radioactive iodine activity, when coolant temperature is above 215
deg.F, the reactor is not critical, and primary containment integrity
has not been established.
Date of issuance: February 20, 2001.
Effective date: As of the date of issuance to be implemented within
30 days of issuance.
Amendment No.: 170.
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 1, 2000 (65 FR
65344).
The staff's related evaluation of the amendment is contained in a
Safety Evaluation dated February 20, 2001.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: November 10, 2000.
Brief description of amendment: The amendment revised several
sections of the Kewaunee Nuclear Power Plant (KNPP) Technical
Specifications (TSs). These sections include administrative changes,
Table 4.1-1, and Sections 1.0, 6.4, and 6.10.
Date of issuance: February 12, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 151.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 13, 2000 (65
FR 77923).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 12, 2001.
No significant hazards consideration comments received: No.
Portland General Electric Company, et al., Docket No. 50-344, Trojan
Nuclear Plant, Columbia County, Oregon
Date of application for amendment: August 5, 1999, as supplemented
by letters dated November 23, 1999, December 27, 1999, May 4, 2000,
October 19, 2000, and November 22, 2000.
Brief description of amendment: The amendment revised the Facility
Operating (Possession Only) License to annotate approval of the Trojan
Nuclear Plant License Termination Plan.
Date of issuance: February 12, 2001.
Effective date: February 12, 2001, and shall be implemented within
30 days of the effective date.
[[Page 13811]]
Amendment No.: 206.
Facility Operating License No. NPF-1: The amendment changes the
Facility Operating (Possession Only) License.
Date of initial notice in Federal Register: December 29, 1999 (64
FR 73083). The November 23, 1999, December 27, 1999, May 4, 2000,
October 19, 2000, and November 22, 2000, supplemental letters provided
additional clarifying information, did not expand the scope of the
application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 12, 2001.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: August 4, 2000 (TS 99-20).
Brief description of amendments: Deletes Sequoyah License Condition
for Shift Technical Advisor and revises Technical Specifications (TSs)
that specify shift manning requirements.
Date of issuance: February 16, 2001.
Effective date: February 16, 2001.
Amendment Nos.: 266 and 257.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the Operating Licenses and TSs.
Date of initial notice in Federal Register: September 6, 2000 (65
FR 54088).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 16, 2001.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: November 21, 2000 (ULNRC-04346).
Brief description of amendment: The amendment changes Table 3.3.2-
1, ``Engineered Safety Feature Actuation System [ESFAS]
Instrumentation,'' of the Technical Specifications. The change adds
Surveillance Requirement (SR) 3.3.2.10 for the following two ESFAS
instrumentation in the table: item 6.f, loss of offsite power, and item
6.h, auxiliary feedwater pump suction transfer on suction pressure--
low.
Date of issuance: February 12, 2001.
Effective date: February 12, 2001, and shall be implemented prior
to entering Mode 3 from Mode 4 during the startup from Refuel Outage
11, including the revision of the FSAR to reflect the ESFAS response
times in accordance with the application.
Amendment No.: 141.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 27, 2000 (65
FR 81931).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 12, 2001.
No significant hazards consideration comments received: No.
Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: February 1, 2001.
Brief description of amendment request: The amendment would remove
the inservice inspection requirements of Section XI of the American
Society of Mechanical Engineers Boiler and Pressure Vessel Code from
the Monticello Technical Specifications and relocates them to a
licensee-controlled program.
Date of publication of individual notice in Federal Register:
February 15, 2001 (66 FR 10535).
Expiration date of individual notice: March 1, 2001.
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
[[Page 13812]]
opportunity for public comment. If comments have been requested, it is
so stated. In either event, the State has been consulted by telephone
whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, located at One White Flint North, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852, 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 April 6, 2001, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, located at One White Flint North,
11555 Rockville Pike (first floor), Rockville, Maryland 20852, 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 20555-001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, located at One White Flint North, 11555 Rockville Pike
(first floor), Rockville, Maryland 20852, by the above date. A copy of
the petition should also be sent to the Office of the General Counsel,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-001, 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).
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of application for amendment: February 14, 2001, as
supplemented February 16 and 19, 2001. The February
[[Page 13813]]
16 and 19, 2001, letters provided additional clarifying information
which did not change the initial proposed no significant hazards
consideration determination or expand the amendment beyond the scope of
the original notice (Harrisburg, PA, Patriot News, February 18-20,
2001).
Brief description of amendment: The amendment allows a one-time
exception to the system configuration and maintenance requirements in
Technical Specification (TS) 3.3.2 related to the nuclear service river
water (NR) system at TMI-1, in order to allow an up to 14-day repair of
a leaking underground concrete pipe. The requirements of TS 3.3.1.4 to
have two NR pumps OPERABLE are unchanged. During the 14-day repair
period, the NR pumps flow will be realigned to pass through a portion
of the nonseismic secondary services river water system.
Date of issuance: February 23, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 229.
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes.
The NRC published a public notice of the proposed amendment, issued
a proposed finding of no significant hazards consideration and
requested that any comments on the proposed no significant hazards
consideration be provided to the staff by the close of business on
February 23, 2001. The notice was published in the Harrisburg, PA,
Patriot News, from February 18 through February 20, 2001.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, consultation with the State of Pennsylvania, and
final no significant hazards consideration determination are contained
in a Safety Evaluation dated February 23, 2001.
Attorney for licensee: Edward J. Cullen, Jr., Esquire, PECO Energy
Company, 2301 Market Street (S23-1), Philadelphia, PA 19103.
NRC Section Chief: Marsha Gamberoni.
Dated at Rockville, Maryland this 27th day of February 2001.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 01-5216 Filed 3-6-01; 8:45 am]
BILLING CODE 7590-01-P