[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