[Federal Register Volume 66, Number 55 (Wednesday, March 21, 2001)]
[Notices]
[Pages 15915-15939]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-6732]


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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 26 through March 9, 2001. The last 
biweekly notice was published on March 7, 2001 (66 FR 13797).

Notice of Consideration of Issuance of Amendments to Facility 
Operating Licenses, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public

[[Page 15916]]

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 20, 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 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).

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of application for amendments: April 26, 2000, as supplemented 
November 6, 2000. This notice supersedes the notice concerning this 
facility that appeared at 65 FR 31356, May 17, 2000.
    Description of amendments request: The proposed amendments would 
revise the maximum Ultimate Heat Sink (UHS) temperature allowed by 
Technical Specification (TS) 3.7.2, ``Service Water (SW) System and 
Ultimate Heat Sink (UHS),'' for the Brunswick Steam Electric Plant 
(BSEP), Unit Nos. 1 and 2. The maximum 24-hour average UHS temperature 
specified in Required Action H.1 would be revised from 89 deg.F to 
90.5 deg.F. To provide consistency with the new maximum 24-hour average 
UHS temperature, these amendments would also: (1) Revise the Condition 
H temperature range from ``>89 deg.F and 92 deg.F'' to 
``>90.5 deg.F and 92 deg.F''; and (2) revise Surveillance 
Requirement 3.7.2.2 to require verification that the UHS temperature is 
90.5 deg.F versus 89 deg.F.

[[Page 15917]]

    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 with the maximum 24 hour average UHS water 
temperature as high as 90.5 deg.F does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The BSEP SW system is designed to provide cooling water for the 
removal of heat from equipment required for a safe reactor shutdown 
following a Design Basis Accident (DBA) or transient. This equipment 
includes the Diesel Generators (DGs), Residual Heat Removal (RHR) 
pump seal coolers, room cooling units for Emergency Core Cooling 
System (ECCS) equipment, and Residual Heat Removal Service Water 
(RHRSW) heat exchangers. The SW system also provides cooling to 
other components, as required, during normal operation. The SW 
system is not an initiator of any previously evaluated accident. The 
safety related components associated with SW cooling have been 
analyzed for a maximum UHS temperature of 92 deg.F. The proposed 
change maintains this maximum UHS temperature. As such, the 
qualification of safety related components is not affected. 
Therefore, the probability of occurrence of a previously evaluated 
accident is not increased.
    The new maximum 24 hour average UHS water temperature limit of 
90.5 deg.F has been evaluated and it was determined that the SW 
system will maintain sufficient heat removal capability. Existing TS 
operability requirements for the UHS ensure that conservatively 
bounding assumptions used in the analysis of the SW system's heat 
removal capability will be met, or the UHS will be declared 
inoperable. As such, the consequences of previously analyzed 
accidents are not affected[.]
    2. Operation with the maximum 24 hour average UHS water 
temperature as high as 90.5 deg.F will not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    Increasing the maximum 24 hour average UHS water temperature 
does not create the possibility of an accident of a different type 
than any evaluated previously in the safety analysis report. UHS 
water temperature does not represent an accident initiator. There is 
no physical change to any plant structure, system, or components. 
Therefore, there is no possibility of an accident of a different 
type.
    Increasing the maximum 24 hour average UHS water temperature 
does not create the possibility of a malfunction of a different type 
than any evaluated previously. The safety related components 
associated with SW cooling have been analyzed for a maximum UHS 
temperature of 92 deg.F. This maximum UHS temperature is maintained 
by the proposed change. As such, this condition does not introduce 
the possibility of a malfunction of a different type than any 
evaluated.
    3. Operation with the maximum 24 hour average UHS water 
temperature as high as 90.5 deg.F does not involve a significant 
reduction in a margin of safety.
    UHS temperature limits are established to ensure that the SW 
system is able to provide sufficient cooling water for the removal 
of heat from equipment, such as the DGs, RHR pump seal coolers, ECCS 
room cooling units, and RHRSW heat exchangers, required for a safe 
reactor shutdown following a DBA or transient. CP&L has performed an 
analysis which demonstrates that this capability is not reduced with 
the increased maximum 24 hour average UHS water temperature limit. 
Existing TS operability requirements for the UHS ensure that 
conservatively bounding assumptions used in the analysis of the SW 
system's heat removal capability will be met, or the UHS will be 
declared inoperable. As such, the ability of the SW system to 
perform its intended safety function is not affected and the margin 
of safety is not reduced.

    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 are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards considerations.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Richard P. Correia.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of amendment request: February 15, 2001.
    Description of amendment request: The proposed amendment revises 
Technical Specifications (TS) 3/4.3.2 ``Engineered Safety Features 
Actuation System Instrumentation,'' 3/4.3.3.1 ``Radiation Monitoring 
Instrumentation,'' 3/4.6.1.1 ``Containment Integrity,'' 3/4.6.1.7 
``Containment Ventilation System,'' 3/4.6.3 ``Containment Isolation 
Valves,'' 3/4.9.4 ``Containment Building Penetrations,'' 3/4.9.9 
``Containment Ventilation System Isolation System,'' and associated 
Bases to clarify and relocate requirements by implementing the guidance 
of pre-approved NUREG-1431, Revision 1.
    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 does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes modify required Actions and Surveillance 
Requirements previously reviewed and approved by the NRC in improved 
Technical Specifications (ITS) and changes to ITS as described in 
TSTF [Technical Specification Traveler Form]-30, TSTF-45, TSTF-46, 
and TSTF-269. These changes are administrative in nature in that 
they do not modify the design or operation of Structures, Systems, 
and Components (SSCs) that initiate or mitigate the consequences of 
an accident.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve new plant components or 
procedures, but only revise existing Technical Specification Actions 
and Surveillance Requirements. These changes do not modify the 
design or operation of Structures, Systems, and Components (SSCs) 
that could initiate an accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed changes modify required Actions and Surveillance 
Requirements previously reviewed and approved by the NRC in improved 
Technical Specifications (ITS) and changes to ITS as described in 
TSTF-30, TSTF-45, TSTF-46, and TSTF-269.

    Therefore, the proposed change 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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Richard P. Correia.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of amendment request: February 28, 2001.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications to incorporate new requirements for 
the Low Pressure Service Water system

[[Page 15918]]

standby pump auto start circuitry, related surveillance requirements, 
and 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:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    No. The Low Pressure Service Water (LPSW) Auto-start circuitry 
provides a means of automatic response to start the standby LPSW 
pump after the running LPSW pump fails to restart following a Loss 
Of Offsite Power (LOOP) event.
    Loss Of Coolant Accidents (LOCA) events actuate the LPSW pumps 
via the Engineered Safeguards Systems. This modification will not 
change this response.
    The LPSW pumps automatically restart following a LOOP event. A 
failure of a running LPSW pump to restart and LPSW header pressure 
not returning to normal operating values following a LOOP event will 
actuate the LPSW Standby Pump Auto-Start circuitry. The circuitry 
will start the LPSW standby pump. When LPSW header pressure returns 
to normal operating values, the auto-start signal will be cleared 
from the LPSW pumps start circuits.
    The modification enhances plant design basis functions by 
ensuring that the standby LPSW pump starts to provide flow. This 
removes the necessity to rely on alternative systems and/or 
components to mitigate design basis events. It will eliminate a 
degraded/non-conforming condition, and will support returning 
affected systems to Maintenance Rule (MR) a(2) status.
    This modification does not involve an 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 kind of accident previously evaluated:
    No. This modification adds LPSW Standby Pump Auto-Start 
circuitry such that if the LPSW pumps fail to restart following a 
LOOP, the standby LPSW pump will start to provide system flow. This 
enhances current plant design. It ensures system flow and eliminates 
reliance on alternative systems and/or components that may or may 
not be safety related to mitigate the design basis event.
    This modification will not create the possibility of a new or 
different kind of accident from any kind of accident previously 
evaluated.
    (3) Involve a significant reduction in a margin of safety.
    No. The proposed change does not adversely affect any plant 
safety limits, set points, or design parameters. The change also 
does not adversely affect the fuel, fuel cladding, Reactor Coolant 
System, or containment integrity. The change will enhance the 
ability to provide flow from the standby LPSW pump following a LOOP. 
It eliminates reliance on alternative systems and/or components to 
mitigate the design basis event should the LPSW pumps fail to 
restart. Therefore, the proposed change does not involve a reduction 
in a margin of safety.
    Duke has concluded, based on the above, that there are no 
significant hazards considerations involved in this amendment 
request.

    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: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: Maitri Banerjee, Acting.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of amendment request: October 30, 2000.
    Description of amendment request: Energy Northwest is requesting a 
revision to the Columbia Generating Station Final Safety Analysis 
Report (FSAR) in regards to the spent fuel storage and spent fuel cask 
handling descriptions. There are significant physical differences 
between the General Electric cask analyzed in the FSAR and the new 
Holtec HI-STORM 100 cask system. The physical description of the 
Columbia Generating Station spent fuel pool as discussed in the FSAR, 
does not accurately reflect the existing configuration. The specific 
changes to the FSAR include:
    1. The FSAR describes two separate pools for spent fuel handling, 
when there is only one pool. The FSAR states that there is a spent fuel 
cask storage and a cask loading pool adjacent to the spent fuel pool. 
There is not a separate spent fuel cask storage and loading pool. There 
is a spent fuel cask loading pit located within the spent fuel pool. 
The proposed change is to eliminate references to separate pools and to 
add a statement that, ``Sufficient redundancy is provided in the 
reactor building crane such that no credible postulated failure of any 
crane component will result in dropping of the fuel cask and rupturing 
the fuel storage pool.''
    2. The FSAR states that limitations on reactor building crane 
travel preclude transporting the spent fuel casks over the spent fuel 
pool. There are no interlocks that prevent crane movement over the 
spent fuel cask pit loading area, which is part of the spent fuel pool. 
There are interlocks that prevent movement over the spent fuel racks. 
The proposed change is to add the statement to the FSAR that, 
``Interlocks on the reactor building crane prevent travel over the 
spent fuel racks.''
    3. The FSAR states that at no time while being transported does the 
fuel cask pass over any safety related equipment. The cask does pass 
over a safety-related conduit associated with a fuel pool cooling level 
instrumentation. The proposed change is to add the statement to the 
FSAR that, ``At no time while being transported does the cask pass over 
any safe shutdown equipment.''
    4. The FSAR discusses cask loading, handling, and features of 
construction associated with the GE IF-300 spent fuel cask rather than 
the Holtec HI-STORM 100 spent fuel cask system, which is the cask 
system that will be used. The proposed change would accurately describe 
the HOLTEC HI-STORM 100 spent fuel cask system.
    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 probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences.
    Accidents previously evaluated in the FSAR that could be 
influenced by these FSAR text changes regarding cask handling and 
spent fuel loading operations include the Spent Fuel Cask Drop 
Accident (FSAR 15.7.5) and the Fuel Handling Accident (FSAR 15.7.4).
    Spent Fuel Cask Drop Accident: Sufficient redundancy is provided 
in the reactor building crane such that no credible postulated 
failure of any crane component will result in dropping of the fuel 
cask and rupturing the fuel storage pool. (Reference: Columbia 
Generating Station FSAR Section 15.7.5, ``Spent Fuel Cask Drop 
Accident''). The drop accident is not deemed credible and the 
revision of the FSAR description will continue to maintain the drop 
accident as incredible. Additionally, as a defense-in-depth measure, 
crane position interlocks prevent lifting a spent fuel cask over the 
spent fuel stored in the pool.
    As the cask is moved in and out of the fuel pool, it passes over 
several cables and conduits supporting plant equipment. They include 
nonsafety-related cables such as those supplying the refueling 
bridge, and spent fuel pool temperature indicator FPC-TE-7. 
Additionally, a safety-related conduit for FPC-LE-5 is included in 
the cask load

[[Page 15919]]

path. While a cask drop, which could damage or cut the cable to FPC-
LE-5 is not credible, operator error in which the cable is damaged 
by the cask not clearing the conduit during cask movement may be 
credible. If the cable were damaged, it might inhibit one train of 
the automatic isolation signal for the fuel pool cooling system. The 
automatic isolation of interest occurs on low fuel pool water level, 
isolating the Seismic Category I cooling portion of the system from 
the Seismic Category II cleanup portion of the system. A fuel pool 
low water level coincident with a crane operator damaging the cable 
for FPC-LE-5 is an extremely low probability event. However, in the 
case of a damaged cable for FPC-LE-5, automatic isolation on low 
water level would still occur because a separate, redundant, logic 
train (from FPC-LE-4) would not be affected and would still be 
capable of accomplishing the isolation function described in FSAR 
Section 9.1.3.2.3. The cable for the redundant logic train is not in 
the cask load path. The cable for FPC-LE-5 also carries a signal for 
high/low spent fuel pool water level alarm, which has a redundant 
analogue signal (undamaged in this scenario) from FPC-LS-4.
    Fuel Handling Accident: The fuel handling accident is analyzed 
in FSAR Section 15.7.4. In it, the assumption is made that a failure 
occurs in a fuel assembly lifting mechanism. The accident which 
produces the largest number of failed spent fuel rods is the drop of 
a spent fuel bundle into the reactor core when the reactor pressure 
vessel (RPV) head is off. The analysis assumes the accident occurs 
at the maximum height allowed by the fuel handling equipment above 
spent fuel (34 ft.). Since the same fuel handling mechanism is used 
in both the reactor (the analyzed accident location) and in the fuel 
pool, but at a considerably lower available drop height 
(approximately 3 ft.), the energy available to damage fuel rods is 
significantly less. As a result, the analyzed fuel handling accident 
consequences bound the consequences of a fuel assembly drop in the 
spent fuel pool. Because fuel loaded in a cask will be within 
approximately 1 ft. [foot] of the elevation of a fuel pool rack, 
fuel handling for cask loading is essentially the same as other fuel 
handling within the pool and is also bounded by the FSAR analysis. 
Therefore the consequences of this accident evaluated previously in 
the FSAR will not be increased by the proposed change.
    The proposed change does not entail any physical alteration to 
the present plant configuration. Therefore, individual precursors of 
an accident are unaffected and the probability of an accident 
previously evaluated is not expected to increase. In addition, since 
the functions and capabilities of systems designed to operate safely 
and/or mitigate the consequences of an accident have not changed, 
the consequences of an accident previously evaluated are not 
expected to increase.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Creation of the possibility of a new or different kind of 
accident would require the creation of one or more new precursors of 
that accident. New accident precursors may be created by 
modifications of the plant configuration.
    Information presented in the FSAR describing the spent fuel cask 
safe load path is revised by this amendment. To agree with the 
current plant configuration noted above, the FSAR will need to be 
changed to read, ``At no time while being transported does the cask 
pass over any safe shutdown equipment.'' The objectives referenced 
in RG [Regulatory Guide] 1.13, Rev. 1, and the guidelines of NUREG-
0612 (to prevent impact by heavy loads with safe shutdown equipment) 
will continue to be met. The proposed change does not entail any 
physical alteration to the present plant configuration. There are no 
new precursors of an accident created and no new or different kinds 
of accidents are created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    There are no plant modifications required as a result of the 
proposed FSAR change. The proposed FSAR text changes correct 
inaccuracies partly resulting from incorrect original process 
descriptions. Since then, there have been significant changes to 
spent fuel cask handling and design requirements including the 
necessity for extended dry storage of spent fuel at independent 
spent fuel storage installations. With the proposed FSAR text 
changes incorporated, the FSAR will accurately describe actual plant 
configuration and processes related to spent fuel cask handling and 
the NRC certified Holtec HI-STORM 100 System.
    The Columbia Generating Station reactor building crane is 
single-failure-proof and therefore no credible postulated failure of 
any crane component will result in dropping of the fuel cask and 
rupturing the fuel storage pool. A single-failure-proof crane 
obviates the need for an isolated spent fuel cask transfer pool. In 
addition, safe load paths are defined that keep the spent fuel cask 
away from irradiated fuel and safe shutdown equipment. This is in 
accordance with defense-in-depth approach as described in NUREG-
0612, Section 5.2, ``Bases for Guidelines''.
    The proposed FSAR change contains information about Columbia 
Generating Station spent fuel cask handling that has not been 
previously reviewed and approved by the NRC; however, there is no 
safety significance to this FSAR amendment request. The FSAR text 
corrections are in agreement with applicable regulations and no 
physical alteration to the plant configuration is required.
    Therefore, this change will not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502.
    NRC Section Chief: Stephen Dembek.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of amendment request: February 20, 2001.
    Description of amendment request: The proposed amendment revises 
the Columbia Generating Station Technical Specifications (TS) to remove 
selected operating mode restrictions for performing emergency diesel 
generator (DG) testing. This change will allow the DG testing to be 
performed during any plant operating mode. The proposed change removes 
the restriction associated with the following surveillance requirements 
(SRs) that prohibit performing the required DG testing during Modes 1 
and 2.
    1. SR 3.8.1.9: This SR requires demonstrating that the DG can 
reject its single largest load without the DG output frequency 
exceeding a specific limit.
    2. SR 3.8.1.10: This SR requires demonstrating that the DG can 
reject its full load without the DG output voltage exceeding a specific 
limit.
    3. SR 3.8.1.14: This SR requires starting and then running the DG 
continuously at or near full-load capability for greater than or equal 
to 24 hours.
    The proposed change also removes the restriction associated with 
the following SRs that prohibits performing the required testing during 
Modes 1, 2, and 3.
    1. SR 3.8.1.13: This SR requires demonstrating that the DG non-
emergency (non-critical) automatic trips are bypassed on an actual or 
simulated emergency core cooling system (ECCS) initiation signal.
    2. SR 3.8.1.17: This SR requires demonstrating that the DG 
automatic switchover from the test mode to ready-to-load operation is 
attained upon receipt of an ECCS initiation signal while maintaining 
availability of the offsite source.
    The proposed change also allows the performance of SR 3.8.1.14 to 
satisfy SR 3.8.1.3 (monthly one-hour synchronized and loaded DG run) by 
adding a Note 5 to SR 3.8.1.3 that allows the endurance and margin test 
of SR 3.8.1.14 to be performed in lieu of load-run test in SR 3.8.1.3.
    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 15920]]


    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The DGs and their associated emergency loads are accident 
mitigating features, not accident initiating equipment. Therefore, 
there will be no impact on any accident probabilities by the 
approval of the requested amendment.
    The design of plant equipment is not being modified by these 
proposed changes. As such, the ability of the DGs to respond to a 
design basis accident will not be adversely impacted by these 
proposed changes. The proposed changes do not result in a plant 
configuration change for performance of the additional testing 
different from that currently allowed by the Technical 
Specifications. In addition, experience and further evaluation of 
the probability of a DG being rendered inoperable concurrent with or 
due to a significant grid disturbance support the conclusion that 
the proposed changes do not involve any significant increase in the 
likelihood of a loss of safety bus. Therefore, there would be no 
significant impact on any accident consequences.
    Based on the above, the proposed change to permit certain DG 
surveillance tests to be performed during plant operation will not 
involve a significant increase on accident probabilities or 
consequences.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    No new accident causal mechanisms would be created as a result 
of NRC approval of this amendment request since no changes are being 
made to the plant that would introduce any new accident causal 
mechanisms. Equipment will be operated in the same configuration 
currently allowed by other DG SRs that currently allow testing in 
plant Modes 1, 2 and 3. An interaction between the DG under test and 
the offsite power system that could lead to a consequential loss of 
safety bus during a grid disturbance is not deemed to be credible. 
This amendment request does not impact any plant systems that are 
accident initiators; neither does it adversely impact any accident 
mitigating systems.
    Based on the above, implementation of the proposed changes would 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. The proposed changes to the testing requirements for the 
plant DGs do not affect the operability requirements for the DGs, as 
verification of such operability will continue to be performed as 
required (except during different allowed Modes). Continued 
verification of operability supports the capability of the DGs to 
perform their required function of providing emergency power to 
plant equipment that supports or constitutes the fission product 
barriers. Consequently, the performance of these fission product 
barriers will not be impacted by implementation of this proposed 
amendment.
    In addition, the proposed changes involve no changes to 
setpoints or limits established or assumed by the accident analysis. 
On this and the above basis, no safety margins will be impacted. 
Therefore, implementation of the proposed changes 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: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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 license amendment request 
consists of changes to the Technical Specifications (TSs) to revise the 
reactor vessel pressure/temperature (P/T or P-T) limits specified in TS 
3.4.11, ``RCS [Reactor Coolant System] Pressure and Temperature (P/T) 
Limits,'' for reactor heatup, cooldown, and critical operation, as well 
as for inservice leak and hydraulic tests for the RCS. Also, the 
current RCS P/T Limits in TS Figure 3.4-11, ``Minimum Temperature 
Required Vs. RCS Pressure,'' would be replaced with recalculated RCS P/
T limits based, in part, on an alternate methodology. The alternate 
methodology uses American Society of Mechanical Engineers (ASME) Boiler 
and Pressure Vessel (B&PV) Code (Code) Case N-640, ``Alternative 
Requirement Fracture Toughness for Development of P-T Limit Curves for 
ASME B&PV Code Section XI, Division 1,'' for alternate reference 
fracture toughness for reactor vessel materials in determining the P/T 
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:

    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?
    The proposed changes to the River Bend [Station] reactor coolant 
system (RCS) pressure/temperature (P/T) limits do not modify the 
boundary, operating pressure, materials or seismic loading of the 
reactor coolant system. The proposed changes do adjust the P/T 
limits for radiation effects to ensure that the RPV [reactor 
pressure vessel] fracture toughness is consistent with analysis 
assumptions and NRC [Nuclear Regulatory Commission] regulations. An 
evaluation has been performed justifying the use of the methodology 
contained in Code Case N-640 to determine the P-T curve. The 
proposed P/T limits were determined using this methodology. Thus, 
the proposed changes do not involve a significant increase in the 
probability of occurrence of an accident previously evaluated. 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.
    Therefore, this change does 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 to the reactor pressure vessel pressure-
temperature limits do not affect the assumed accident performance of 
any structure, system or component previously evaluated. The 
proposed changes do not introduce any new modes of system operation 
or failure mechanisms.
    Therefore, this change does 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 methodology for determining the RCS P/T limits ensures that 
the limits provide a margin of safety to the conditions at which 
brittle fracture may occur. The methodology is based on requirements 
set forth in Appendix G and Appendix H of 10 CFR [Part] 50, with 
reference to the requirements and guidance of ASME Section Xl, and 
on guidance provided in Regulatory Guide 1.99, Revision 2. The 
revised P/T limits are also based on this methodology except as 
modified by application of the noted Code Case. Although the Code 
Case constitutes relaxation from the current requirements of 10 CFR 
[Part] 50 Appendix G, the alternatives allowed by the Code are based 
on industry experience gained since the inception of the 10 CFR 
[Part] 50 Appendix G requirements for which some of the requirements 
have now been determined to be excessively conservative. The more 
appropriate assumptions and provisions allowed by the Code Case 
maintain a margin of safety that is consistent with the intent of 10 
CFR [Part] 50 Appendix G, i.e., with regard to the margin originally 
contemplated by 10 CFR

[[Page 15921]]

[Part] 50 Appendix G for determination of RPV/RCS P/T limits.
    Therefore, this change 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 
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 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 amendment request proposes 
that the River Bend Station Operating License be amended to change the 
limit on the Low Power Setpoint Limit specified by Technical 
Specifications 3.1.3 ``Control Rod OPERABILITY,'' 3.1.6 ``Control Rod 
Pattern,'' and 3.3.2.1 ``Control Rod Block Instrumentation'' from less 
than or equal to 20% reactor thermal power to less than or equal to 10% 
reactor thermal power.
    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 changes do not significantly increase the 
probability or consequences of an accident previously evaluated.
    The proposed change revises the setpoint from 20% to 10% rated 
power and does not affect the function, reliability or required 
surveillance frequency of the RPC [Rod Pattern Control] set forth in 
the Technical Specification. It does not constitute a safety 
significant change to the plant design or operation since the RPC 
and associated BPWS [Banked Position Withdrwawal Sequence] will 
continue to ensure site compliance with 10 CFR [Code of Federal 
Regulations Part] 100.
    The RPC limits the incremental worth of control rods during 
reactor startup and shutdown. The BPWS allows continuous withdrawal 
from fully inserted to the fully withdrawn position for the first 
25% of control rod density. The change in LPSP [Low Power Setpoint 
Limit] does not affect any of the parameters or conditions that 
contribute to initiation of the control rod drop accident since it 
is not the precursor of the accident. On this basis, change in the 
low power setpoint will not increase the probability of an accident 
previously evaluated.
    The low power setpoint of the RPC is set so that the resultant 
peak fuel enthalpy due to the postulated rod drop accident shall be 
equal to or less than 280 cal/gm. For operation below the LPSP, 
systems are provided so that the design limit of 280 cal/gm is not 
exceeded for the design basis accident. Conformance to the 280 cal/
gm design limit also ensures that the 10 CFR [Part] 100 offsite dose 
criteria will not be exceeded for the design basis accident. GE 
[General Electric] generic analysis demonstrates the radiological 
effect following a CRDA [Control Rod Drop Accident], for all current 
GE fuel design is within the guidelines set forth in 10 CFR [Part] 
100. No River Bend specific analysis is necessary. On these bases, 
the proposed LPSP reduction does not significantly change the 
consequences of an accident previously evaluated.
    2. The request does not create the possibility of occurrence of 
a new or different kind of accident from any accident previously 
evaluated.
    The LPSP is set so that the resultant peak fuel enthalpy due to 
the postulated rod drop accident at power levels below the LPSP, 
shall be equal to or less than 280 cal/gm, ensuring compliance with 
10 CFR [Part] 100 offsite dose criteria. The proposed change 
implements the reduction in LPSP from 20% to 10% of rated power 
without the addition of new hardware.
    The change in LPSP does not affect any of the parameters or 
conditions that contribute to initiation of any accident since the 
LPSP is not the precursor of any accident. The LPSP is the point at 
which the RPCS [Rod Pattern Control System] switches between the RPC 
and RWL [Rod Withdrawal Limit] function. Periodic verification that 
it is within the allowable value is required. The proposed change 
does not affect the function and the reliability of the RPC, or the 
required surveillance frequency of Technical Specification LCO 
[Limiting Condition for Operation]. Furthermore, the reduction in 
setpoint can be implemented without the addition of new hardware. On 
this basis, reduction in the low power setpoint does not create the 
possibility of occurrence of a new or different accident.
    3. The request does not involve a significant reduction in 
margin of safety.
    Below the LPSP, mitigating systems and procedures are used to 
limit the consequences of a postulated CRDA. These involve a time 
consuming process of a series of controlled rod moves or steps. The 
setpoint change has the potential to impact the margin of safety and 
as such, a series of evaluations and under the worst case scenario 
were performed for a CRDA. NEDO-10527 demonstrates that a CRDA at or 
above 10% of rated power will always result in peak fuel enthalpies 
less than 280 cal/gm. These results assumed the worst single 
operator error, conservative Technical Specification scram times and 
rod drop velocity. This generic analysis also included the effect of 
core and fuel cycle design parameters such as the axial gadolinia 
distributions. The results indicate, that even for this worst case 
scenario, the resultant peak fuel enthalpy will always be less than 
280 cal/gm, ensuring conformance with guidelines set forth in 10 CFR 
[Part] 100. Additional vendor analyses show that ``Above 
approximately 10% power, the RDA cannot exceed 280 cal/gm because of 
the prompt Doppler feedback in the power range and the impossibility 
of achieving high rod reactivity worth with the relatively low rod 
density, even with erroneous rod patterns.'' Finally, the new 
models, which include moderator reactivity feedback, provide 
additional justification for the 10% of rated power LPSP. These 
methods indicate that the existence of any steam flow (i.e., power) 
will result in the CRDA results remaining below the design basis 
limit. Therefore, a LPSP limit of 10% is conservative relative to 
the new models. On these bases, the proposed reduction in the LPSP 
does not change the margin of safety significantly.

    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 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 request consists of a change 
to Technical Specification 3.6.1.3, ``Primary Containment Isolation 
Valves (PCIVs),'' to permit the operation of the Inclined Fuel Transfer 
System (IFTS) bottom valve after removal of the IFTS primary 
containment isolation blind flange while the containment is required to 
be operable.
    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 changes do not significantly increase the 
probability or consequences of an accident previously evaluated.
    The proposed change permits the operation of the IFTS Bottom 
valve after removal of the inclined fuel transfer system (IFTS) 
primary containment isolation blind flange when primary containment 
operability is required in MODE 1, 2, and 3. This will permit the 
full operation of the IFTS while the plant is operating. With 
respect to the probability of

[[Page 15922]]

an accident, this aspect of the containment structure does not 
directly interface with the reactor coolant pressure boundary. 
Operation of the IFTS bottom valve after the removal of the blind 
flange does not involve modifications to plant systems or design 
parameters that could contribute to the initiation of any accidents 
previously evaluated. Operation of IFTS is unrelated to the 
operation of the reactor, and there is no aspect of IFTS operation 
that could lead to or contribute to the probability of occurrence of 
an accident previously evaluated. Operation of the IFTS bottom valve 
during operation of IFTS system after removal of the blind flange 
does not result in changes to procedures that could impact the 
occurrence of an accident.
    With respect to the issue of consequences of an accident, the 
function of the containment is to mitigate the radiological 
consequences of a loss of coolant accident (LOCA) or other 
postulated events that could result in radiation being released from 
the fuel inside containment. While the proposed change does not 
change the plant design, it does permit an alteration of the 
containment boundary for the IFTS penetration. Altering the 
containment boundary in this case (i.e., Opening the IFTS bottom 
valve) would not result in any additional IFTS components being 
subjected to containment pressure in the event of a LOCA. However, 
the additional post-accident peak pressure load to be imposed upon 
the components in the IFTS if the blind flange is removed is a small 
fraction of their design capability. Therefore, they are considered 
an acceptable barrier to prevent uncontrolled release of post-
accident fission products for this proposed change.
    As discussed in LAR [License Amendment Request] 1999-30, the 
proposed change required examination of two potential leakage 
pathways. The larger is the IFTS transfer tube, itself. The other, 
much smaller one, is a branch line used for draining the IFTS 
transfer tube during its operation. The bottom of the IFTS transfer 
tube is always water sealed, and maintained so by the submergence of 
the water in the transfer tube and in the fuel building spent fuel 
storage pool (the lower pool). The height of this water seal is 
greater than that necessary to prevent leakage from the bottom of 
the transfer tube during accidents that result in the calculated 
peak post-DBA [design basis accident] LOCA pressure, Pa. 
The potential leakage pathway from the drain piping that attaches to 
the transfer tube will be isolated if required, via administrative 
controls on the drain piping isolation valve. Additionally, as 
committed to in LAR 1999-30, the drain piping isolation valve will 
be added to the Primary Containment Leakage Rate Testing Program 
(Technical Specification 5.5.13) to ensure that leakage past this 
valve will be maintained consistent with the leakage rate 
assumptions of the accident analysis. Due to the test methodology, 
the portion of the large transfer tube piping outboard of the blind 
flange (the portion of the tube which becomes exposed to the 
containment atmosphere during the draining portion of the IFTS 
operation) will also be part of the leakage rate test boundary and 
will therefore also be tested. Therefore, no unidentified leakage 
will exist from the piping and components that are outboard of the 
blind flange, and the leakage rate assumptions of the accident 
analysis will be maintained.
    Therefore, the proposed change does not result in a significant 
increase in the probability or the consequences of previously 
evaluated accidents.
    2. The proposed changes would not create the possibility of a 
new or different kind of accident from any previous analyzed.
    The proposed change consists of permitting operation of the IFTS 
Bottom valve after the removal of a the IFTS Blind Flange which is 
not part of the primary reactor coolant pressure boundary nor 
involved in the operation or shutdown of the reactor. Being passive, 
the presence or absence of the IFTS Blind Flanges does not affect 
any of the parameters or conditions that could contribute to the 
initiation of any incidents or accidents that are created from a 
loss of coolant or an insertion of positive reactivity. Realigning 
the boundary of the primary containment to include portions of the 
IFTS is also passive in nature and therefore has no influence on, 
nor does it contribute to the possibility of a new or different kind 
of incident, accident or malfunction from those previously analyzed. 
Furthermore, operation of the IFTS is unrelated to the operation of 
the reactor and there is no mishap in the process that can lead to 
or contribute to the possibility of losing any coolant from the 
reactor or introducing the chance for an insertion of positive or 
negative reactivity, or any other accidents different from and not 
bounded by those previously evaluated.
    Therefore, the proposed change does not result in creating the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed change involves the operation of the IFTS Bottom 
Valve after realignment of the primary containment boundary by 
removing the blind flange which is a passive component. The margin 
of safety that has the potential of being impacted by the proposed 
change involves the dose consequences of postulated accidents which 
are directly related to potential leakage through the primary 
containment boundary. The potential leakage pathways due to the 
proposed change have been reviewed, and leakage can only occur from 
the administratively controlled IFTS transfer tube drain piping, and 
from the IFTS transfer tube itself. A dedicated individual will be 
designated to provide timely isolation of this drain piping during 
the duration of time when this proposed change is in effect. The 
conservatively calculated dose which might be received by the 
designated individual while isolating the drain piping is calculated 
to be 3.8 rem [roentgen equivalent man] TEDE [Total Effective Dose 
Equivalent], which remains within the guidelines of General Design 
Criterion (GDC) 19 (10 CFR [Code of Federal Regulations Part] 50, 
Appendix A, Criterion 19). Furthermore, the drain piping isolation 
valve will be added to the Primary Containment Leakage Rate Testing 
Program (Technical Specification 5.5.13) to ensure that leakage from 
the piping and components located outboard of the blind flange will 
be maintained consistent with the leakage rate assumptions of the 
accident analysis.
    Studies of the capability of the IFTS system to withstand 
containment pressurization under severe accident conditions have 
been conducted. These studies conclude that IFTS, including the 
transfer tube and its valves, has a capability to withstand beyond 
design basis severe accident containment pressures which is greater 
than that of the containment structure itself. The RBS [River Bend 
Station] Emergency Operating Procedures (EOPs) are based on an 
ultimate containment failure pressure capability of 53 psig [pounds 
per square inch gauge], which represents a margin of safety of 38 
psi [pounds per square inch] above the 15 psig containment design 
pressure.
    This capability to withstand containment pressurization under 
severe accident conditions envelops other non-DBA LOCA scenarios, 
such as the small break LOCA. For the large break LOCA, additional 
defense-in-depth is provided by maintaining a water seal greater 
than Pa above the outlet of the IFTS transfer tube in the 
lower pool.
    The RBS base LERF [Large Early Release Frequency] is 5.915E-9/
yr. Removal of the blind flange increases the LERF by 6.315E-9/yr to 
1.223E-8/yr. This increase in LERF is due to the reduced failure 
pressure of the IFTS tube. With the blind flange installed, the IFTS 
tube has a median failure pressure of approximately 80 psig. The 
IFTS tube was evaluated to withstand a pressure of 40 psig, with the 
blind flange removed. This lower IFTS failure pressure increases the 
probability of gross failure versus penetration failure at a given 
containment pressure. This shift in failure probability means that 
some of the less severe pressurization events (i.e. small hydrogen 
deflagrations) have a higher probability of causing a LERF. Based on 
the RBS PRA [Probabilistic Risk Assessment] Analysis, the operation 
of the bottom valve has no affect on LERF.
    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: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3, Westchester County, New York

    Date of amendment request: February 14, 2001.
    Description of amendment request: The amendment would modify the

[[Page 15923]]

Indian Point Nuclear Generating Unit No. 3 (IP3) Technical 
Specifications (TSs) to extend the allowed outage time (AOT) for the 
emergency diesel generators (EDGs) and the associated fuel oil storage 
tanks (FOSTs) from 72 hours to 14 days on a one-time basis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the proposed License amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. The proposed License amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. The EDGs and their associated fuel 
oil systems are not part of any accident initiation; therefore there 
is no increase in the probability of an accident.
    At a minimum, two EDGs are still available with sufficient fuel 
oil supply to mitigate IP3 design basis accidents. The minimum 
safeguards equipment can still be powered even if one EDG and FOST 
is assumed to be lost due to single failure. This has been verified 
by EDG loading calculation, IP3-CALC-ED-00207, ``480V Bus 2A, 3A, 5A 
& 6A and EDGs 31, 32 and 33 Accident Loading''. With the associated 
EDG available and aligned for automatic start capability (although 
declared inoperable) during this EDG FOST outage, further backup to 
the remaining two EDGs is provided. By the design of the overall EDG 
fuel oil system, the associated EDG fuel oil day tank is able to be 
supplied with sufficient fuel oil supply from either of the 
remaining two FOSTs, via their transfer pumps, in order to support 
operation of this associated EDG, if necessary.
    To support fuel oil needs of all three EDGs, if necessary, the 
FSAR [final safety analysis report] describes that additional fuel 
oil supplies are available on the Indian Point site and locally near 
the site. Further EDG fuel oil supplies are available in the region, 
about 40 miles from IP3. Overall, the EDGs are designed as backup AC 
power sources in the event of a Loss of Offsite Power (LOOP). The 
proposed one-time AOT for each EDG/FOST does not change the 
conditions or minimum amount of safeguards equipment assumed in the 
safety analysis for design basis accident mitigation, since a 
minimum of two EDGs is assumed. No changes are proposed as to how 
the EDGs provide plant protection. Additionally, no new modes of 
overall plant operation are proposed as a result of this change. A 
PRA [probablistic risk assessment] evaluation determined that the 
conditional core damage probability (CCDP) for these scenarios is 
less than the threshold value of 1 E-6. Therefore, the proposed one-
time license amendment to TS 3.7.B.1 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. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The proposed change does not introduce any new overall 
modes of plant operation or make any permanent physical changes to 
plant systems necessary for effective accident mitigation. The 
minimum required EDG operation remains unchanged by removal of a 
single FOST for repair. Additionally, added requirements to minimize 
risk associated with loss of offsite power also support this one-
time extended AOT. Also, as previously stated, the EDGs and FOSTs 
are not part of any accident initiation scenario. Therefore the 
proposed one-time license amendment to TS 3.7.B.1 does not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    (3) Does the proposed License amendment involve a significant 
reduction in a margin of safety?
    No. The proposed License amendment does not involve a 
significant reduction in a margin of safety. The minimum safeguards 
loads can be maintained available if needed for design basis 
accident mitigation with two EDGs operable combined with their 
respective FOSTs. The selected, inoperable EDG will be available and 
aligned for automatic start capability (though declared inoperable) 
during this outage. The additional fuel oil needed to support three 
EDGs in this condition is available as indicated in the present 
design and licensing basis. The FSAR describes that this fuel can be 
provided from the Indian Point site, local sources and from a source 
about 40 miles away to support the additional 30,026 gallons TS 
required fuel oil, already existing at the Buchanan substation. 
Therefore, sufficient fuel oil will be available for potential 
events that could occur during this 14-day AOT. The PRA evaluation 
for the case of maintaining the 31, 32 or 33 EDG available (though 
declared inoperable) with its FOST out for repair indicates an 
acceptable safety margin below the risk-informed threshold of 1 E-6.
    The 480VAC electrical distribution system can be fed from a 
number of TS independent 13.8kV and 138kV offsite power sources to 
minimize reliance of IP3 on EDG power sources during the extended 
AOT requested. Additional requirements to minimize risk associated 
with the potential for loss of offsite power sources within this TS 
change also ensure that this extended AOT does not involve a 
significant reduction in safety margin. On this basis, the proposed 
one-time license amendment to TS 3.7.B.1 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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Generating Station, 600 Rocky Hill Road, Plymouth, MA 
02360.
    NRC Section Chief: Marsha Gamberoni.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of amendment request: January 25, 2001.
    Description of amendment request: Entergy Operations, Inc. is 
proposing that the Grand Gulf Nuclear Station (GGNS) Operating License 
be amended to revise the GGNS Technical Specification (TS), 
Surveillance Requirement (SR) 3.1.4.2 to increase the control rod scram 
time testing interval from 120 days to 200 days of full power 
operation. The licensee also proposes to revise the associated TS Bases 
to reflect the proposed revision to the SR.
    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?
    The proposed change will not adversely impact plant operation. 
There will be no change in the method of performing the tests. The 
extended test frequency will provide some positive safety benefits 
by reducing the complexity of half of the control rod sequence 
exchange maneuvers, reducing the likelihood of a reactivity or fuel 
related event.
    The actual rod insertion times and control rod reliability are 
not impacted by this proposed change; only the probability of 
detecting slow rods is impacted. The potential consequence of the 
proposed change is that one or more slow rods that would have been 
detected under the current 120-day frequency, may not be detected 
due to a reduced number of tests under the 200-day frequency.
    Historical data shows that the GGNS control rod insertion 
function is highly reliable and rod insertion tests meet the scram 
time limits 99.84% of the time. Statistical analysis also 
demonstrates that the extended frequency would have little impact on 
the ability to detect slow rods in the sampling tests.
    There is no safety consequence resulting from ``slow'' rods so 
long as the plant does not exceed the Technical Specification 3.1.4 
Limiting Condition of Operation [LCO] requirement of no more than 14 
slow rods in the entire core or no two OPERABLE ``slow''

[[Page 15924]]

rods occupying adjacent positions. It is highly unlikely that a 
combination of missed detections and known ``slow'' rods would lead 
to the requirement to take action in accordance with TS 3.1.4. 
Therefore, it is highly unlikely that the reduction in test 
frequency would have any impact on plant operation or safety.
    The analysis assumes that all 14 slow rods take 7 seconds to 
reach notch position 13 which is very conservative base on actual 
rod performance. Control rod data shows that rods that have failed 
the time requirements are usually only a fraction of a second 
slower. In the unlikely event that, due to the reduction of test 
frequency, the plant is unknowingly operating with one or two more 
slow rods than the 14 slow control rods permitted by the LCO, the 
consequences would still be insignificant. The low probability of 
MODE 1 operation with excess slow rods combined with the low 
consequence of a few excess slow rods, leads to the conclusion that 
the probability or consequences of accidents previously evaluated 
are not significantly increased.
    Therefore, this change does 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 change will make no change to plant configuration 
or test procedures. The proposed change does not impact the 
operation of the plant except to reduce the number of required tests 
and slightly increase the probability of failing to detect a slow 
control rod. Operating with possibly one or two undetected slow rods 
does not create the possibility of an accident, since sudden control 
rod insertion by scram only occurs during the mitigation of 
accidents.
    Therefore, this change does 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 GGNS accident analyses assume a certain negative reactivity 
time function associated with scrams. So long as the LCO of 
Technical Specification 3.1.4 is met, that is, there are no more 
than 14 slow control rods in the entire core or two OPERABLE 
``slow'' rods occupying adjacent locations, all accident analysis 
assumptions are met and there is no reduction in any margin of 
safety. The proposed change does not impact the Technical 
Specification LCO, or any other allowable operating condition. The 
potential for an increase in the probability of being outside 
acceptable operating conditions due to this proposed change is 
insignificant. Calculations have demonstrated that the likelihood of 
detecting four slow rods with proposed testing frequency over a fuel 
cycle is lower than that with the current testing frequency by a 
negligible amount (2E-O7). The difference is even smaller for 
detecting greater number of slow rods over a cycle. Therefore, since 
there is no impact on allowable operating parameters and the 
likelihood of detecting significant numbers of slow rods is only 
negligibly affected, there is no significant reduction in a margin 
of safety.
    Therefore, this change 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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, LLC, 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 amendment request: November 30, 2000.
    Description of amendment request: The proposed amendment would 
revise the ``Diesel Fuel Oil Testing Program'' in technical 
specifications to relocate the specific American Society for Testing 
Materials (ASTM) standard reference from the Administrative Controls 
Section of TS to a licensee-controlled document, i.e., the Diesel Fuel 
Oil Program in the Technical Requirements Manual (TRM). In addition, 
the ``clear and bright'' test has been expanded to allow a water and 
sediment content test to establish the acceptability of new fuel oil. 
The proposed changes are consistent with changes previously approved by 
the Nuclear Regulatory Commission (NRC).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed changes relocate the specific diesel fuel oil 
related American Society for Testing and Materials (ASTM) Standard 
reference from the Administrative Controls Section of Technical 
Specifications (TS) to a licensee-controlled document, i.e., the 
Diesel Fuel Oil Program in the Technical Requirements Manual (TRM). 
The Braidwood Station and the Byron Station TRM is incorporated by 
reference in the Braidwood and Byron Stations' Updated Final Safety 
Analysis Report (UFSAR). Since any change to these licensee-
controlled documents will be evaluated pursuant to the requirements 
of 10 CFR 50.59, ``Changes, tests and experiments,'' no increase in 
the probability or consequences of an accident previously evaluated 
is involved. In addition, the ``clear and bright'' test used to 
establish the acceptability of new fuel oil for use prior to 
addition to storage tanks has been expanded to allow a water and 
sediment content test to be performed to establish the acceptability 
of new fuel oil in lieu of the ``clear and bright'' test. We 
consider that the quantitative water and sediment test is equivalent 
to the qualitative clear and bright test.
    Relocating the specific ASTM Standard references from the TS to 
a licensee-controlled document (i.e., the Diesel Fuel Oil Program in 
the TRM), and allowing a water and sediment content test to be 
performed to establish the acceptability of new fuel oil, will not 
affect nor degrade the ability of the safety-related diesel 
generators (DGs) (i.e., the Emergency DG and the Auxiliary Feedwater 
pump DG) to perform their specified safety function. Fuel oil 
quality will continue to meet ASTM requirements.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components (SSCs) 
from performing their intended function to mitigate the consequences 
of an initiating event within the acceptance limits assumed in the 
Braidwood and Byron Stations' UFSAR. The proposed changes do not 
affect the source term, containment isolation, or radiological 
release assumptions used in evaluating the radiological consequences 
of an accident previously evaluated in the Braidwood and Byron 
Stations' UFSAR.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind accident from any accident previously evaluated?
    The proposed changes relocate the specific ASTM Standard 
reference from the Administrative Controls Section of TS to a 
licensee-controlled document, i.e., the Diesel Fuel Oil Program in 
the TRM. In addition, the ``clear and bright'' test used to 
establish the acceptability of new fuel oil for use prior to 
addition to storage tanks has been expanded to allow a water and 
sediment content test to be performed to establish the acceptability 
of new fuel oil.
    The changes do not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. In addition, 
the changes do not impose any new or different requirements or 
eliminate any existing requirements. The changes do not alter 
assumptions made in the safety analysis. Therefore, the changes do 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?

[[Page 15925]]

    The proposed changes relocate the specific ASTM Standard 
reference from the Administrative Controls Section of TS to a 
licensee-controlled document, i.e., the Diesel Fuel Oil Program in 
the TRM. Instituting the proposed changes will continue to ensure 
the use of current applicable ASTM Standards to evaluate the quality 
of both new and stored fuel oil designated for use in the safety-
related DGs. The detail associated with the specific ASTM Standard 
reference is not required to be in the TS to provide adequate 
protection of the public health and safety, since the TS still 
retain the requirement for compliance with the applicable ASTM 
Standard. Changes to the TRM are evaluated in accordance with 10 CFR 
50.59. Should it be determined that future changes involve a 
potential reduction in a margin of safety, NRC review and approval 
would be necessary prior to implementation of the changes. This 
approach provides an effective level of control and provides for a 
more appropriate change control process. In addition, the ``clear 
and bright'' test used to establish the acceptability of new fuel 
oil for use prior to addition to storage tanks has been expanded to 
allow a water and sediment content test to be performed to establish 
the acceptability of new fuel oil in lieu of the ``clear and 
bright'' test. The level of safety of facility operation is 
unaffected by the proposed changes since there is no change to the 
TS requirements intended to assure that fuel oil is of the 
appropriate quality for safety-related DG use. The proposed changes 
provide the flexibility needed to maintain state-of-the-art 
technology in fuel oil sampling and analysis methodology.
    Therefore, the 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 amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Vice President, 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348
    NRC Section Chief: Anthony J. Mendiola.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: February 20, 2001.
    Description of amendment request: The proposed amendments would 
increase the allowed outage time from 3 days to 14 days for a single 
inoperable Division 1 or 2 emergency diesel generator.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes include the extension of the completion 
time for the Emergency Diesel Generators (EDGs) from 72 hours to 14 
days to allow on-line preventive maintenance to be performed. The 
EDGs are not initiators of previously evaluated postulated 
accidents. Extending the completion times of the EDGs would not have 
any impact on the frequency of any accident previously evaluated, 
and therefore the probability of a previously analyzed accident is 
unchanged. The proposed change to the completion time for EDGs will 
not result in any changes to the plant activities associated with 
EDG maintenance, but rather will enable a more efficient planning 
and scheduling of maintenance activities that will minimize 
potential adverse interactions with concurrent outage activities.
    The consequences of a previously analyzed event are the same 
during a 72 hour EDG completion time as the consequences during a 14 
day completion time. Thus the consequences of accidents previously 
analyzed are unchanged between the existing TS requirements and the 
proposed change. In the worst case scenario, the ability to mitigate 
the consequences of any accident previously analyzed is preserved. 
The consequences of an accident are independent of the time the EDGs 
are out-of-service. As a general practice, no other additional 
failures are postulated while equipment is inoperable within its TS 
completion time.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously analyzed.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not involve a physical change to the 
plant. No new equipment is being introduced, and installed equipment 
is not being operated in a new or different manner. Therefore, these 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed changes will extend the allowable completion times 
for the Required Actions associated with restoration of an 
inoperable Division 1 or Division 2 EDG. The proposed 14 day EDG 
completion time is based upon both a deterministic evaluation and a 
risk-informed assessment. The availability of offsite power coupled 
with the availability of the opposite unit EDG via the unit cross-
tie breaker and the use of the Configuration Risk Management Program 
(CRMP) provide adequate compensation for the potential small 
incremental increase in plant risk of the EDG extended completion 
time. In addition, the increased availability of the EDGs during 
refueling outage offsets the small increase in plant risk during 
operation. The proposed EDG extended completion times in conjunction 
with the availability of the opposite unit EDG continues to provide 
adequate assurance of the capability to provide power to the 
Engineered Safety Feature (ESF) buses. The risk assessment concluded 
that the increase in plant risk is small and consistent with the 
NRC's Safety Goal Policy Statement, ``Use of Probabilistic Risk 
Assessment Methods in Nuclear Activities: Final Policy Statement,'' 
Federal Register, Volume 60, p. 42622, August 16, 1995, and guidance 
contained in Regulatory Guides (RG) 1.174, ``An Approach for Using 
Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-
Specific Changes to the Licensing Basis,'' dated July, 1998, and RG 
1.177, ``An Approach for Plant-Specific, Risk-Informed Decision 
Making: Technical Specifications,'' dated August, 1998. Together, 
the deterministic evaluation and the risk-informed assessment 
provide high assurance of the capability to provide power to the ESF 
buses during the proposed 14 day EDG completion time.
    Therefore, implementation of the proposed changes will not 
involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Vice President, 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.
    NRC Section Chief: Anthony J. Mendiola.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: February 21, 2001 (TS-265).
    Brief description of amendment: The proposed amendment would revise 
the Crystal River Unit 3 (CR-3) Improved Technical Specifications (ITS) 
3.3.8 to clarify the actions to be taken in the event that one or more 
channels of loss of voltage or degraded voltage Emergency Diesel 
Generator (EDG) start functions become inoperable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91, the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below.

    1. Does not involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The emergency diesel generator (EDG) loss of power start is not 
an initiator of any design basis accident. The EDG loss of power 
start is intended to protect engineered safeguards

[[Page 15926]]

equipment from damage due to sustained undervoltage conditions, and 
to ensure rapid restoration of power to the engineered safeguards 
electrical buses in the event of a loss of offsite power.
    The proposed license amendment clarifies the actions to be taken 
in the event that one or more channels of the undervoltage or 
degraded voltage start Functions become inoperable. The design 
functions of the EDG loss of power start and the initial conditions 
for accidents that require an EDG loss of power start will not be 
effected by the change. Therefore, the change will not increase the 
probability or consequences of an accident previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously analyzed.
    The proposed amendment involves no changes to the design or 
operation of the EDG loss of power start. The proposed changes will 
ensure that the EDGs and engineered safeguards actuation system 
(ESAS) automatic initiation logic perform as assumed in the safety 
analysis in the event of a loss of offsite power. The proposed 
change will not affect other EDG or ESAS functions, and will not 
create any new plant configurations. Therefore, the proposed change 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does not involve a significant reduction in the margin of 
safety.
    The proposed amendment clarifies the actions to be taken in the 
event one or more undervoltage or degraded voltage start Functions 
become inoperable. The proposed changes ensure appropriate actions 
are taken to restore the operability of the EDG loss of power start 
under these conditions. Thus, the proposed amendment will not result 
in a reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: R. Alexander Glenn, Associate General 
Counsel, Florida Power Corporation, MAC-A5A, P.O. Box 14042, St. 
Petersburg, Florida, 33733-4042.
    NRR Section Chief: Richard P. Correia.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 (CR-3) Nuclear Generating Plant, Citrus County, Florida

    Date of amendment request: February 21, 2001 (TS-266).
    Description of amendment request: The changes proposed revise 
various administrative actions, requirements, and responsibilities 
contained in Improved Technical Specifications (ITS) 2.0, Safety 
Limits, and ITS 5.0, Administrative Controls, to reflect the recent CR-
3 Nuclear Operations re-organization and the amended requirements of 10 
CFR 50.72, 10 CFR 50.73 and 10 CFR 50.59.
    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 not involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed license amendment deletes redundant administrative 
requirements contained in ITS 2.0, ``Safety Limits'' and updates 
position titles in ITS 5.0, ``Administrative Controls,'' to reflect 
the current CR-3 Nuclear Operations organization. The design 
functions of the structures, systems and components at CR-3, and the 
initial conditions for the analyzed accidents at CR-3 will not be 
affected by the change. Therefore, the change will not increase the 
probability or consequences of an accident previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously analyzed.
    The changes proposed by this amendment are administrative in 
nature. The proposed amendment involves no changes to the design, 
function or operation of any structure, system or component at CR-3 
and will not result in any new plant configurations. Therefore, the 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does not involve a significant reduction in the margin of 
safety.
    The proposed changes are administrative in nature. The safety 
margins established through the design and facility license, 
including the CR-3 Improved Technical Specifications will not be 
changed by the proposed amendment. In addition, the proposed changes 
will ensure that administrative requirements and responsibilities 
contained in the ITS are consistent with the current CR-3 Nuclear 
Operations organization as described in the CR-3 Final Safety 
Analysis Report and the requirements specified in 10 CFR 50.72, 10 
CFR 50.73 and 10 CFR 50.59. Thus, the proposed amendment will not 
result in a reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: R. Alexander Glenn, Associate General 
Counsel (MAC-BT15A), Florida Power Corporation, P.O. Box 14042, St. 
Petersburg, Florida 33733-4042.
    NRC Section Chief: Richard P. Correia.

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook 
Nuclear Plant, Unit 2, Berrien County, Michigan

    Date of amendment request: January 19, 2001.
    Description of amendment request: The proposed amendment would 
extend surveillance intervals associated with the emergency diesel 
generators and station batteries to preclude a mid-cycle shutdown of 
the unit.
    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 of occurrence or consequences of an accident previously 
evaluated?
    The proposed license conditions do not affect or create any 
accident initiators or precursors. As such, the proposed license 
conditions do not increase the probability of an accident. The 
proposed license conditions do not involve operation of the required 
electrical power sources in a manner or configuration different from 
those previously recognized or evaluated.
    The proposed EDG [emergency diesel generator] engine SR 
[surveillance requirement] revision involves deferral of the 
4.8.1.1.2.e.1 requirement to the next refueling outage and does not 
reduce the required operable power sources of the Limiting Condition 
for Operation, does not increase the allowed outage time of any 
required operable power supplies, and does not reduce the 
requirement to know that the deferred SRs could be met at all times. 
Deferral of the testing does not increase by itself the potential 
that the testing would not be met. The monthly EDG engine starts, 
fuel level checks, and fuel transfer pump checks will continue to be 
performed to provide adequate confidence that the required EDG 
engine will be available if needed. Therefore, it is concluded that 
the required A.C. sources will remain available and the previously 
evaluated consequences will not be increased.
    The deferral of the battery service tests described above to the 
refueling outage does not involve any physical changes to the plant 
or to the manner in which the plant is operated. Therefore, the 
probability of an accident previously evaluated is not increased. 
The weekly and quarterly testing, performance monitoring by the 
system manager, and the current condition of the batteries (e.g., 
above 100 percent capacity) provide assurance that battery condition 
and performance will not deteriorate during the deferral period. 
Therefore, the consequences of the analyzed accidents for CNP [Cook 
Nuclear Plant] will not be increased due to the deferral of these 
station battery SRs.
    Therefore, based on the above discussion, it is concluded that 
the proposed amendment does not involve a significant increase in 
the

[[Page 15927]]

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 license condition does not involve a physical 
alteration of the EDG engines or a change to the way the A.C. power 
system is operated. The proposed license condition does not involve 
operation of the required electrical power sources in a manner or 
configuration different from those previously recognized or 
evaluated. No new failure mechanisms of the A.C. power supplies are 
introduced by extension of the subject SR intervals.
    The proposed license conditions for deferral of the station 
battery SRs listed above to the refueling outage do not involve any 
physical changes to the plant or to the manner in which the plant 
D.C. power systems are operated. No new failure mechanisms will be 
introduced by the SR deferral.
    Therefore, the proposed license condition does not create the 
possibility of a new or different kind of accident from any accident 
previously analyzed.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Deferral of the specified EDG engine SR does not introduce by 
itself a failure mechanism, and past performance of the SR has 
demonstrated reliability in passing the deferred SRs. The required 
operable power supplies have not been reduced. Therefore, the 
availability of power supplies assumed for accident mitigation is 
not significantly reduced and previous margins of safety are 
maintained.
    The deferral of the station battery SRs to the refueling outage 
does not involve any physical changes to the plant or to the manner 
in which the plant is operated. Continuing weekly and quarterly 
testing, performance monitoring, and the current condition of the 
batteries provides assurance that the battery condition and 
performance will be acceptable during the deferral period in that 
degradations that may occur will be detected. Therefore, the 
equipment response to accident conditions during the deferral period 
will not be affected. Thus, the one-time deferral of these 18-month 
battery service test SRs does not involve a significant reduction in 
a margin of safety.
    In summary, based upon the above evaluation, I&M has concluded 
that the proposed amendment involves no significant hazards 
consideration.

    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: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: Claudia M. Craig.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of amendment request: January 3, 2001.
    Description of amendment request: The proposed amendment would 
terminate license jurisdiction for a portion of the Maine Yankee Atomic 
Power Station site, thereby releasing these lands from Facility 
Operating License No. DPR-36. The release of these lands will 
facilitate the donation of this property to an environmental 
organization pursuant to a Federal Energy Regulatory Commission-
approved settlement between Maine Yankee Atomic Power Company and its 
ratepayers. The lands donated will be used to create a nature preserve 
and an environmental education center.
    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 requested license amendment involves release of land 
presently considered part of the Maine Yankee plant site under 
license DPR-36. The land in question is not used for any licensed 
activities. No radiological materials have historically been used on 
this land and the land will not be used to support ongoing 
decommissioning operations and activities.
    Most of the land to be released is outside the Exclusion Area 
Boundary and therefore is not affected by the consequences of any 
postulated accident. A small portion of the land is within the 
Exclusion Area Boundary. Maine Yankee will retain sufficient control 
over activities performed within this land through rights granted in 
the legal land conveyance documents to ensure that there is no 
impact on consequences from postulated accidents. Therefore, the 
release of the land from the Part 50 license will not increase the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The requested amendment involves release of land presently 
considered part of the Maine Yankee plant site under license DPR-36. 
The land is not used for any licensed activities or decommissioning 
operations. The proposed action does not affect plant systems, 
structures or components in any way. The requested release of the 
land does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety defined in the statements of consideration 
for the final rule on the Radiological Criteria for License 
Termination is described as the margin between the 100 mrem/yr 
public dose limit established in 10 CFR 20.1301 for licensed 
operation and the 25 mrem/yr dose limit to the average member of the 
critical group at a site considered acceptable for unrestricted use. 
This margin of safety accounts for the potential effect of multiple 
sources of radiation exposure to the critical group. Additionally, 
the State of Maine, through legislation, has imposed a 10 mrem/yr 
all pathways limit, with no more than 4 mrem/yr attributable to 
drinking water sources. Since the survey results described in 
Attachments III and IV demonstrate compliance with the radiological 
criteria for license termination for unrestricted use and 
demonstrate compliance with the more stringent Maine Standard, 
therefore, the margin of safety will not be reduced as a result of 
the proposed release of the nonimpacted land. In fact, since the 
area is nonimpacted, by definition, there will be no additional dose 
to the average member of the critical group.

    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: Robert A. Gramm.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station Unit No. 2, Oswego County, New York

    Date of amendment request: February 5, 2001.
    Description of amendment request: The licensee proposed to amend 
Section 3.6.1.3, ``Primary Containment Isolation Valves,'' of the 
unit's Technical Specifications (TSs). Surveillance Requirement (SR) 
3.6.1.3.9 currently requires verification of the actuation capability 
of each excess flow check valve (EFCV) at least once per 24 months. One 
proposed change will result in limiting the surveillance to only those 
EFCVs in instrumentation lines connected to the reactor coolant 
pressure boundary. The requirement for testing of EFCVs other than 
those in reactor instrumentation lines is proposed to be relocated to a 
licensee-controlled document. Another proposed change is to revise the 
SR by allowing a representative sample of reactor instrumentation line 
EFCVs to be tested every 24 months, such that each reactor 
instrumentation line EFCV will be tested every 10 years.

[[Page 15928]]

    The associated licensee-controlled TSs Basis document would also be 
changed to reflect the above TSs changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the three standards of 10 CFR 50.92(c). The NRC staff's review 
is presented below:
    The first standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. The proposed changes to SR 3.6.1.3.9 
will result in reduction in the frequency and scope of EFCV testing. 
No hardware design change is involved. While a postulated instrument 
line break accident was analyzed and evaluated as part of the design 
basis, no credit was given to EFCVs to limit or stop radioactive 
water through the ruptured instrument line. The EFCVs were not 
considered precursor of accidents in the unit's design basis. 
Accordingly, the revised scope and frequency of EFCV testing will 
lead to no increase in the consequences of an accident previously 
evaluated, and no increase of the probability of an accident 
previously evaluated.
    The second standard requires that operation of the unit in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. No hardware design change or procedural change 
is involved with the proposed changes to SR 3.6.1.3.9. The amendment 
would only relax the frequency and scope of EFCV testing. Therefore, 
the proposed amendment will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The third standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety. Since no design or 
procedural change is involved, the proposed changes to SR 3.6.1.3.9 
will not affect in any way the performance characteristics and 
intended functions of systems and components (i.e., the instrument 
lines and instruments) served by the EFCVs. Therefore, the proposed 
changes to SR 3.6.1.3.9 do not involve a significant reduction in a 
margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Marsha Gamberoni.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station Unit No. 2, Oswego County, New York

    Date of amendment request: February 27, 2001.
    Description of amendment request: The licensee proposed to amend 
Technical Specifications (TSs) Section 3.3.8.2, ``Reactor Protection 
System (RPS) Electric Power Monitoring--Logic,'' reducing the channel 
calibration allowable values for overvoltage from 133.8 V to 130.2 V 
(for Bus A), and to 129.8 V (for Bus B). The licensee also proposed to 
amend Section 3.3.8.3, ``Reactor Protection System (RPS) Electric Power 
Monitoring--Scram Solenoids,'' reducing the channel calibration 
allowable values for overvoltage from 130.5 V (for Bus A) and 131.7 V 
(for Bus B) to 127.6 V. These proposed changes are in the conservative 
direction, reflecting the results of revisions to calculations to 
correct licensee-identified analysis deficiencies. The proposed reduced 
allowable values would be accompanied by an increase in channel 
calibration frequency from once per 24-months to once per 184 days.
    The associated licensee-controlled TSs Basis document would also be 
changed to reflect the above TSs changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the three standards of 10 CFR 50.92(c). The NRC staff's review 
is presented below:
    The first standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. The proposed changes to Sections 
3.3.8.2 and 3.3.8.3 will be made in a conservative direction. No 
hardware design change is involved, thus there will be no adverse 
effect on the functional performance of any plant structure, system, 
or component (SSC). All SSCs will continue to perform their design 
functions with no decrease in their capabilities to mitigate the 
consequences of postulated accidents. Accordingly, the revised 
allowable values and channel calibration frequencies will lead to no 
increase in the consequences of an accident previously evaluated, 
and no increase of the probability of an accident previously 
evaluated.
    The second standard requires that operation of the unit in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. No hardware design change or procedural change 
is involved with the proposed changes to these sections. The 
amendment does not involve any changes in design or performance of 
any SSC; all SSCs will continue to perform as previously analyzed by 
the licensee and previously accepted by the staff. Therefore, the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The third standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety. Since no design or 
procedural change is involved, the proposed changes to Sections 
3.3.8.2 and 3.3.8.3 will not affect in any way the performance 
characteristics and intended functions of any SSC. Therefore, the 
proposed changes do not involve a significant reduction in a margin 
of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Marsha Gamberoni.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of amendment requests: October 30, 2000.
    Description of amendment requests: The proposed amendments would 
allow modification of the eight double-leaf doors in the auxiliary 
building special ventilation zone. These doors serve as ``blowout 
panels'' in case of a high-energy line break (HELB) accident inside the 
auxiliary building. Currently, these doors are held in place by the 
resistance from the hinges and door center latch. The licensee proposes 
to install additional ``breakaway'' pins on these doors to increase the 
restraining forces upon these doors to minimize nuisance alarms from 
these doors. However, the licensee has determined that this 
modification did not meet the criteria of 10 CFR 50.59 and therefore 
requires prior NRC staff review and approval. These amendments do not 
involve changes to the Operating Licenses or the Technical 
Specifications.
    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 15929]]


    1. Does operation of the facility with the proposed amendment 
involve a significant increase in the probability or consequences of 
an accident previously evaluated?
    The proposed change does not significantly affect any system 
that is a contributor to initiating events for previously evaluated 
accidents. The addition of a ceramic latch pin in selected Auxiliary 
Building Special Ventilation Zone (ABSVZ) boundary doors will 
provide a small restraining force to hold the doors closed under 
typical operating conditions, but will snap under the pressures 
produced on the doors by a high-energy line break, thus allowing the 
doors to swing open and provide a relief path for steam discharge 
into the Auxiliary Building compartments during a HELB. Testing has 
established that the ceramic pins will breakaway under a load that 
is significantly lower than the differential pressure loading on the 
boundary doors assumed in the HELB analyses. In addition, improving 
the ability to keep these doors closed under normal operating 
conditions helps to assure maintenance of the ABSVZ boundary 
integrity assumed in the LOCA [loss-of-coolant accident] and offsite 
dose analyses. Thus it is concluded that the proposed changes do not 
involve any significant increase in the probability or consequence 
of an accident previously evaluated.
    2. Does operation of the facility with the proposed amendment 
create the possibility of a new or different kind of accident from 
any accident previously evaluated?
    While the proposed modification alters the design of plant 
equipment, it does not alter the function or the manner of operation 
[of] any plant component and does not install any new or different 
equipment. During a HELB selected ABSVZ boundary doors are required 
to swing open to provide a steam relief path. The use of ceramic 
pins to restrain these doors against inadvertent opening during 
normal operations does not alter the accident mitigation function of 
these doors. Testing has established that these ceramic pins will 
break before the pressure in the Auxiliary Building reaches the 
relief point assumed in the HELB analyses. This situation does not 
create the possibility of a new or different kind of accident from 
those previously analyzed.
    3. Does operation of the facility with the proposed amendment 
involve a significant reduction in a margin of safety?
    Because testing has established that these ceramic pins will 
break before the pressure in the Auxiliary Building reaches the 
relief point assumed in the HELB analyses, the accident mitigation 
function of the ABSVZ boundary doors will be preserved. In the event 
of a HELB the ABSVZ boundary doors will swing open and provide a 
steam relief path. Thus avoiding any increased Auxiliary Building 
compartment pressures that might challenge the requirements on 
ventilation boundary leakage and block wall structural integrity 
established to maintain assurance of control room habitability.
    Thus, the proposed change does not involve a significant 
reduction in the margin of safety associated with the safety limits 
inherent in either the principle barriers to a radiation release 
(fuel cladding, RCS [reactor coolant system] boundary, and reactor 
containment), or the maintenance of critical safety functions 
(subcriticality, core cooling, ultimate heat sink, RCS inventory, 
RCS boundary integrity, and containment integrity).

    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.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: February 14, 2001.
    Description of amendment request: The proposed amendment would make 
minor changes to the Ginna Improved Technical Specifications (ITS) 
format to allow for maintaining, viewing, and publishing them with 
different software package. The proposed amendment would also revise 
the ITS section 5.5.13, ``Technical Specifications Bases Control 
Program,'' to provide consistency with the changes to 10 CFR 50.59 as 
published in the Federal Register (64 FR 53582) dated October 4, 1999.
    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:

Evaluation of Administrative Formatting Changes

    The administrative changes associated with the minor revisions 
in the Ginna Station ITS format to allow for maintaining, viewing, 
and publishing them with different software package do not involve a 
significant hazards consideration as discussed below:
    (1) Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. The proposed 
changes involve minor reformatting of the existing Improved 
Technical Specifications to provide compatibility with the software 
package that is proposed for maintenance of the electronic ITS files 
and do not include any technical issues. As such, these changes are 
administrative in nature and do not impact initiators of analyzed 
events or assumed mitigation of accident or transient events. 
Therefore, the probability or consequences of an accident previously 
evaluated is not significantly increased.
    (2) Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
changes do not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) or changes in 
the methods governing normal plant operation. The proposed changes 
will not impose any new or different requirements. Thus, the 
possibility for a new or different kind of accident from any 
accident previously evaluated is not created.
    (3) Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. The proposed changes will not reduce a margin of safety 
because the changes do not impact any safety analysis assumptions. 
These changes are administrative in nature. As such, no question of 
safety is involved, and the changes do not involve a significant 
reduction in a margin of safety.
    Based upon the preceding information, it has been determined 
that the proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated, 
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. Therefore, it is concluded that the 
proposed changes meet the requirements of 10 CFR 50.92(c) and do not 
involve a significant hazards consideration.

Evaluation of Administrative 10 CFR 50.59 Changes

    The administrative changes associated with the revision to ITS 
section 5.5.13, ``Technical Specifications (TS) Bases Control 
Program,'' to provide consistency with the changes to 10 CFR 50.59 
do not involve a significant hazards consideration as discussed 
below:
    (1) Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. The proposed 
change deletes the reference to unreviewed safety question as 
defined in 10 CFR 50.59. Deletion of the definition of unreviewed 
safety question was approved by the NRC [Nuclear Regulatory 
Commission] with the revision of 10 CFR 50.59. Changes to the TS 
Bases are still evaluated in accordance with 10 CFR 50.59. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    (2) Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
changes do not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) or changes in 
the methods governing normal plant operation. Thus, the possibility 
for a new or different kind of accident from any accident previously 
evaluated is not created.
    (3) Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a

[[Page 15930]]

margin of safety. The proposed changes will not reduce a margin of 
safety because the changes do not impact any safety analysis 
assumptions. Changes to the ITS Bases that result in meeting the 
criteria in paragraph 10 CFR 50.59(c)(2) will still require NRC 
approval pursuant to 10 CFR 50.59. This change is administrative in 
nature based on the revision to 10 CFR 50.59. As such, no question 
of safety is involved, and the changes do not involve a significant 
reduction in a margin of safety.
    Based upon the preceding information, it has been determined 
that the proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated, 
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. Therefore, it is concluded that the 
proposed changes meet the requirements of 10 CFR 50.92(c) and do 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 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, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005.
    NRC Section Chief: Marsha Gamberoni.

Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco 
Nuclear Generating Station, Sacramento County, California

    Date of amendment request: February 20, 2001.
    Description of amendment request: The proposed license amendment 
would eliminate the security plan requirements from the 10 CFR Part 50 
licensed site after the Rancho Seco spent nuclear fuel has been 
transferred from the spent fuel pool to the Independent Spent Fuel 
Storage Installation (ISFSI). Specific changes would include deleting 
Section 2.C(3) ``Physical Protection'' from Rancho Seco Facility 
Operating License No. DPR-54 and deleting all references in the 
Permanently Defueled Technical Specifications to the Rancho Seco 
Nuclear Generating Station security plans.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. The physical structures, systems, and components of the 
Rancho Seco 10 CFR 50 licensed site and the operating procedures for 
their use are unaffected by the proposed change. The elimination of 
the security requirements from the 10 CFR Part 50 licensed site does 
not affect possible initiating events for accidents previously 
evaluated or alter the configuration or operation of the facility.
    Elimination of the security requirements for the 10 CFR Part 50 
license is predicated upon completion of the transfer of all nuclear 
fuel from the spent fuel pool to the ISFSI. The planned 10 CFR 72 
licensing controls for the ISFSI will provide adequate confidence 
that personnel and equipment can perform satisfactorily for normal 
operations of the ISFSI and respond adequately to off-normal and 
accident events. The Rancho Seco Physical Protection Plan (PPP) will 
also provide confidence that security personnel and safeguards 
systems will perform satisfactorily to ensure adequate protection 
for the storage of spent nuclear fuel. Therefore, the proposed 10 
CFR Part 50 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. The proposed change is security related and has no direct 
impact on plant equipment or the procedures for operating plant 
equipment. 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.
    Because the ISFSI site is segregated from the 10 CFR Part 50 
licensed site, licensed security activities under the 10 CFR Part 50 
license will no longer be necessary after all the nuclear fuel has 
been moved. The planned 10 CFR 72 licensing controls for the ISFSI 
will provide adequate confidence that personnel and equipment can 
perform satisfactorily for normal operations of the ISFSI and 
respond adequately to off-normal and accident events. Moreover, the 
ISFSI will be physically separate from the 10 CFR 50 licensed site 
structures and equipment. Therefore, the proposed 10 CFR Part 50 
license amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed license amendment involve a significant 
reduction in a margin of safety?
    No. As described above, the proposed change is security related 
and has no direct impact on plant equipment or the procedures for 
operating plant equipment. There are no changes to the design or 
operation of the facility.
    The assumptions for fuel handling and other accidents are not 
affected by the proposed license amendment. Accordingly, neither the 
design basis nor the accident assumptions in the Defueled Safety 
Analysis Report (DSAR), nor the PDTS 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 
amendment request 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: Stephen Dembek.

TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: May 17, 2000, as supplemented by letters 
dated August 31, 2000, and January 31, 2001.
    Brief description of amendments: The proposed amendments would 
revise the Allowable Values specified in Technical Specification (TS) 
Table 3.3.5-1, ``Loss of Power (LOP) Diesel Generator (DG) Start 
Instrumentation'' to ensure that the 6.9 kiloVolt (kV) and 480 Volt (V) 
undervoltage relays initiate the necessary actions when required. In 
addition, a proposed administrative change to Condition D of TS 3.3.5, 
would eliminate the term ``undervoltage,'' consistent with the proposed 
changes to TS Table 3.3.5-1.
    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:
    The proposed License Amendment Request includes more restrictive 
Allowable Values for the Preferred offsite source bus undervoltage 
function, the Alternate offsite source bus undervoltage function, 
the 6.9 kV Class 1E bus loss of voltage function, the 6.9 kV Class 
1E bus degraded voltage function and the 480 V Class 1E bus degraded 
voltage function. These more restrictive values assure that all 
applicable safety analysis limits are being met. The 480 V low grid 
undervoltage relay allowable value is being lowered to the same as 
the 480 V degraded voltage relays which matches its function. This 
is a less restrictive value but the value still assures that all 
applicable safety analysis limits are being met. Lowering of the 480 
V low grid undervoltage allowable value will minimize unnecessary 
actuations that could challenge plant systems. Changing the 6.9 kV 
and 480 V degraded voltage, 480 V low grid undervoltage, the 6.9 kV 
loss of voltage, and the preferred and alternate bus undervoltage

[[Page 15931]]

Allowable Values in the TSs has no impact on the probability of 
occurrence of any accident previously evaluated. Because all 
accident analyses continue to be met, these changes do not impact 
the consequences of any accident previously evaluated.
    Removal of the lower limit for the 6.9 kV Class 1E bus loss of 
voltage relays does not impact the probability of occurrence of any 
accident previously evaluated. None of the accident analyses are 
affected; therefore, the consequences of all previously evaluated 
accidents remain unchanged.
    The proposed administrative change to Condition D of TS 3.3.5, 
which would eliminate the term ``undervoltage,'' consistent with the 
proposed changes to TS Table 3.3.5-1 is administrative in nature. 
None of the accident analyses are affected; therefore, the 
probability and consequences of all previously evaluated accidents 
remain unchanged.
    None of the changes to TS Table 3.3.5-1 affect plant hardware or 
the operation of plant systems in a way that could initiate an 
accident. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed administrative change to Condition D of TS 3.3.5, 
which would eliminate the term ``undervoltage,'' consistent with the 
proposed changes to TS Table 3.3.5-1 is administrative in nature. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    There were no changes made to any of the accident analyses or 
safety analysis limits as a result of this proposed change. Further, 
the proposed change does not affect the acceptance criteria for any 
analyzed event. Removal of the lower limit for the 6.9 kV Class 1E 
bus loss of voltage relays does not change the margin of safety. 
Each allowable value, as revised, assures the safety analysis limits 
assumed in the safety analyses as discussed in Chapter 15 of the 
Final Safety Analysis Report is maintained. The margin of safety 
established by the Limiting Conditions for Operation also remains 
unchanged. Thus there is no effect on the margin of safety.
    The proposed administrative change to Condition D of TS 3.3.5, 
which would eliminate the term ``undervoltage,'' consistent with the 
proposed changes to TS Table 3.3.5-1 is administrative in nature. 
Thus there is no effect on the margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application request: February 15, 2001 (ULNRC-4391).
    Description of amendment request: The proposed amendment would 
delete paragraph d.1.j(2) in Technical Specification (TS) 5.5.9, 
``Steam Generator (SG) Tube Surveillance Program,'' that requires all 
SG tubes containing an Electrosleeve, a Framatome proprietary process, 
to be removed from service within two operating cycles following 
installation of the first Electrosleeve. This requirement was 
incorporated in TS 5.5.9 in Amendment No. 132 issued May 21, 1999. The 
first Electrosleeve tube was installed in the Fall of 1999 and the two-
cycle allowance will expire in the Fall of 2002.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change would remove the restriction that requires 
all steam generator tubes repaired with Electrosleeves to be removed 
from service at the end of two operating cycles following 
installation of the first Electrosleeve. This would allow all steam 
generator tubes repaired with Electrosleeves to remain in service. 
Reference 2 [licensee's letter dated October 27, 1998] concluded 
that there was no significant increase in the probability or 
consequences of an accident previously evaluated when using the 
Electrosleeve repair method. The two operating cycle restriction was 
invoked because the NRC staff concluded that the UT [ultrasonic] 
methods used to perform NDE [nondestructive examination] for 
inservice inspections of the Electrosleeved tubes could not reliably 
depth size stress corrosion cracks to ensure that structural limits 
are maintained.
    Revision 4 to topical report BAW-10219P [nonproprietary version 
is attached to the application] has addressed the concerns that 
resulted in the restriction of two operating cycles and 
consequently, the probability of an accident previously evaluated is 
not significantly increased. As a result, the consequences of any 
accident previously evaluated are not affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing plant operation. Reference 2 
concluded that the use of the Electrosleeve repair method did not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated when using this method to repair 
steam generator tubes. This proposed change removes the two 
operating cycle limit for the Electrosleeved tubes based on the 
evaluations and justifications of the NDE techniques used to perform 
inservice examinations of the Electrosleeved steam generator tubes 
provided in Revision 4 of the topical report.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not affect the acceptance criteria for 
an analyzed event. The margin of safety presently provided by the 
structural integrity of the steam generator tubes remains unchanged. 
Reference 2 concluded that the use of the Electrosleeve repair 
method did not involve a significant reduction in a margin of safety 
when using this method to repair steam generator tubes. The proposed 
change removes the two operating cycle limit based on the 
evaluations and justifications presented in Revision 4 of the 
topical report.
    Therefore, the proposed change does not involve a reduction in a 
margin of safety.

    The reference to ``Reference 2'' in the criteria above is a 
reference to the licensee's letter dated October 27, 1998, and the no 
significant hazards consideration (NHSC) in that letter, which was 
published in the Federal Register (63 FR 66604) on December 2, 1998. 
This NHSC is applicable to the current application because it applies 
to the use of Electrosleeved steam generator tubes, the subject of the 
current application.
    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: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Section Chief: Stephen Dembek.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Units No. 1 and No. 2, Surry County, Virginia

    Date of amendment request: December 7, 2000. This amendment request 
supersedes the November 29, 1999, request in its entirety. The November 
29, 1999, request was noticed on March 22, 2000 (65 FR 15388).

[[Page 15932]]

    Description of amendment request: The proposed changes will modify 
the Technical Specifications (TS) in Section 3.23 for the Main Control 
Room and Emergency Switchgear Room Ventilation and Air Conditioning 
Systems; TS Surveillance Requirement Section 4.20 for the Control Room 
Air Filtration System; and TS Surveillance Requirement Section 4.12 for 
the Auxiliary Ventilation Exhaust Filter Trains. The proposed changes 
will revise the above Surveillance Requirements for the laboratory 
testing of the carbon samples for methyl iodide removal efficiency to 
be consistent with American Society for Testing and Materials (ASTM) 
Standard D3803-1989, ``Standard Test Method for Nuclear-Graded 
Activated Carbon,'' with qualification as the laboratory testing 
standard for both new and used charcoal adsorbent used in the 
ventilation system.
    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--Operation of Surry Units 1 and 2 in accordance with 
the proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes only modify surveillance testing 
requirements and do not affect plant systems or operation and 
therefore do not increase the probability or the consequences of an 
accident previously evaluated. The proposed surveillance 
requirements adopt ASTM D-3803-1989, with qualification, as the 
laboratory method for testing samples of the charcoal adsorber for 
methyl iodide removal efficiency consistent with NRC's Generic 
Letter 99-02. This method of testing charcoal adsorbers provides an 
acceptable approach for determining methyl iodide removal efficiency 
and ensuring that the efficiency assumed in the accident analysis is 
still valid at the end of the operating cycle. There is no change in 
the method of plant operation or system design with this change.
    Criterion 2--The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed changes only modify surveillance testing 
requirements and do not impact plant systems or operations and 
therefore do not create the possibility of an accident or 
malfunction of a different type than evaluated previously. The 
proposed surveillance requirements adopt ASTM D3803-1989, with 
qualification, as the laboratory method for testing samples of the 
charcoal adsorber for methyl iodide removal efficiency. This change 
is in response to NRC's request in Generic Letter 99-02. There is no 
change in the method of plant operation or system design. There are 
no new or different accident scenarios, transient precursors, nor 
failure mechanisms that will be introduced.
    Criterion 3--The proposed license amendment does not involve a 
significant reduction in a margin of safety.
    The proposed changes only modify surveillance test requirements 
and do not impact plant systems or operations and therefore do not 
significantly reduce the margin of safety. The revised surveillance 
requirements adopt ASTM D3803-1989, with qualification, as the 
laboratory method for testing samples as the charcoal adsorber for 
methyl iodide removal efficiency. The 1989 edition of this standard 
imposes stringent requirements for establishing the capability of 
new and used activated carbon to remove methyl iodide from air and 
gas streams. The results of this test provide a more conservative 
estimate of the performance of nuclear-graded activated carbon used 
in nuclear power plant HVAC systems for the removal of methyl 
iodide. The laboratory test acceptance criteria contain a safety 
factor to ensure that the efficiency assumed in the accident 
analysis is still valid at the end of the operating cycle.
    This evaluation concludes that the proposed amendment to the 
Surry Units 1 and 2 Technical Specifications does not involve a 
significant increase in the probab[ility] or consequences of a 
previously evaluated accident, does not create the possibility of a 
new or different kind of accident and 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Donald P. Irwin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Section Chief (Acting): M. Banerjee.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: December 12, 2000, as supplemented 
January 8 and February 22, 2001.
    Description of amendment request: The proposed changes would revise 
Technical Specification (TS) 3.17.4 and 3.17.5 and the appropriate 
Bases. The proposed changes will acknowledge the establishment of seal 
injection for the reactor coolant pump in an isolated and drained loop 
as a prerequisite for the vacuum-assisted backfill technique. Also, the 
proposed changes include additional limiting conditions for operation 
and surveillance requirements for the sources of borated water used 
during loop backfill, and revised reactivity controls for an isolated-
filled loop.
    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 Technical Specification limiting conditions for 
operation and surveillance requirements ensure that the initiation 
of seal injection in order to allow a partial vacuum to be 
established in an isolated and drained loop will not create the 
potential for an inadvertent/undetected introduction of under-
borated water into an isolated loop prior to returning the isolated 
loop to service. The proposed Technical Specification controls 
prevent any additions of makeup or seal injection that would violate 
the existing shutdown margin requirements for the active portion of 
the Reactor Coolant System. Thus, adequate Technical Specification 
controls are established to preclude an inadvertent/undetected 
positive reactivity addition event. Therefore, there is no increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    There are no modifications to the plant as a result of the 
changes. The proposed Technical Specification Limiting Conditions 
for Operation and Surveillance Requirements ensure that the 
initiation of seal injection will not create an undetected positive 
reactivity addition. No new accident or event initiators are created 
by the initiation of seal injection for the RCP [reactor coolant 
pump] in the isolated loop in order to establish a partial vacuum in 
that isolated and drained loop. Therefore, the proposed changes do 
not create the possibility of any accident or malfunction of a 
different type previously evaluated.
    3. Does the change involve a significant reduction in the margin 
of safety as defined in the bases on any Technical Specifications.
    The proposed changes have no effect on safety analyses 
assumptions. Rather, the proposed changes acknowledge the 
establishment of seal injection for the RCP in the isolated and 
drained loop as a prerequisite for the vacuum-assisted backfill 
technique. The proposed Technical Specification Limiting Conditions 
for Operation and Surveillance Requirements ensure that the 
initiation of seal injection in order to allow a partial vacuum to 
be established in an isolated and drained loop will not create the 
potential for an inadvertent/undetected introduction of under-
borated water into an isolated loop prior to returning the isolated 
loop to service. Adequate Technical Specifications controls

[[Page 15933]]

are established to preclude an inadvertent/undetected positive 
reactivity addition event. In addition, the proposed controls 
prevent any additions of makeup or seal injection that would violate 
the existing shutdown margin requirements for the active portion of 
the Reactor Coolant System. Therefore, the proposed changes do not 
result in a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams, 
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
23219.
    NRC Section Chief (Acting): M. Banerjee.

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.

Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear 
Plant, Unit 2, Limestone County, Alabama

    Date of application for amendments: February 5, 2001 (TS-413).
    Brief description of amendments: Changes the Reactor Vessel 
Material Surveillance schedule to allow a one-cycle delay in removal of 
the second capsule.
    Date of publication of individual notice in the Federal Register: 
February 28, 2001 (66 FR 12818).
    Expiration date of individual notice: March 30, 2001.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland 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).

Exelon Generation Company, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: September 5, 2000, as 
supplemented January 17, 2001.
    Brief description of amendments: The amendments revised 
Surveillance Requirement 4.6.3.4 to allow a representative sample of 
reactor instrumentation line excess flow check valves (EFCVs) to be 
tested every 24 months, instead of testing each EFCV every 24-months.
    Date of issuance: As of date of issuance and shall be implemented 
within 30 days.
    Effective date: February 23, 2001.
    Amendment Nos.: 148 and 110.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 10, 2001 (66 FR 
2021).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 23, 2001.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: November 9, 2000.
    Brief description of amendment: By letter dated November 9, 2000, 
FirstEnergy Nuclear Operating Corporation (FENOC), requested a 
Technical Specification change for Davis-Besse Nuclear Power Station 
(DBNPS), Unit 1. The proposed Technical Specification (TS) changes 
would relocate Technical Specification 3/4.4.9.2, Reactor Coolant 
System--Pressurizer, to the Davis-Besse Nuclear Power Station (DBNPS) 
Technical Requirements Manual (TRM). The TRM is a DBNPS controlled 
document which has been incorporated into the Davis-Besse Updated 
Safety Analysis Report (USAR).
    Date of issuance: February 27, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 245.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81919).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 27, 2001.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio

    Date of application for amendment: April 5, 2000, as supplemented 
by letter dated January 15, 2001.
    Brief description of amendment: This amendment implements technical 
specification (TS) changes associated with thermo-hydraulic stability 
monitoring. New TS 3.3.1.3, ``Oscillation Power Range Monitor

[[Page 15934]]

(OPRM) Instrumentation,'' is added, providing the minimum operability 
requirements for the OPRM channels, the Required Actions when they 
become inoperable, and appropriate surveillance requirements. The 
amendment also removes monitoring guidance from TS 3.4.1, 
``Recirculation Loops Operating,'' that will no longer be necessary due 
to the activation of the OPRM instrumentation, and updates TS 5.6.5, 
``Core Operating Limits Report (COLR),'' to require the applicable 
setpoints for the OPRMs to be included in the COLR.
    Date of issuance: February 26, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 118.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 31, 2000 (65 FR 
34745).
    The supplemental information contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 26, 2001.
    No significant hazards consideration comments received: No.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: June 1, 2000.
    Brief description of amendment: The Technical Specification (TS) 
Section 3.4.14, ``RCS Leak Detection Instrumentation, Surveillance 
Requirements,'' was changed to extend the calibration interval of the 
containment sump monitor to 24 months.
    Date of issuance: March 7, 2001.
    Effective date: March 7, 2001.
    Amendment No.: 195.
    Facility Operating License No. DPR-72: Amendment revised the TSs.
    Date of initial notice in Federal Register: July 12, 2000 (65 FR 
43048).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 7, 2001.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant, 
Unit No. 1, St. Lucie County, Florida

    Date of application for amendment: October 30, 2000.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) Limiting Condition For Operation 3.9.4.b to allow 
both doors of the containment personnel airlock to be open during core 
alterations if: (1) at least one personnel airlock door is capable of 
being closed, (2) the plant is in Mode 6 with at least 23 feet of water 
above the fuel in the reactor core, and (3) a designated individual is 
available outside the personnel airlock to close the door.
    Date of Issuance: February 27, 2001.
    Effective Date: February 27, 2001.
    Amendment No.: 172.
    Facility Operating License No. DPR-67: Amendment revised the TS.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81920).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 27, 2001.
    No significant hazards consideration comments received: No.

GPU Nuclear, Inc. and Saxton Nuclear Experimental Corporation, Docket 
No. 50-146, Saxton Nuclear Experimental Facility (SNEF), Bedford 
County, Pennsylvania

    Date of application for amendment: November 30, 2000 and 
supplemented on January 18, 2001.
    Brief description of amendment: The amendment changes the Amended 
Facility License to reflect the change in the legal name of GPU Nuclear 
Corporation to GPU Nuclear, Inc. wherever it appears in the license.
    Date of Issuance: March 8, 2001.
    Effective date: The license amendment is effective as of its date 
of issuance.
    Amendment No.: 17.
    Amended Facility License No. DPR-4: The amendment revised the 
Amended Facility License.
    Date of initial notice in Federal Register: January 10, 2001 (66 FR 
2010). The Commission's related evaluation of the amendment is 
contained in a safety evaluation dated March 8, 2001.
    No significant hazards consideration comments received: No.

GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear 
Station, Unit 2, Dauphin County, Pennsylvania

    Date of amendment request: November 5, 1999, as supplemented by 
electronic mail dated March 22 and letter dated September 28, 2000.
    Brief description of amendment request: The amendment revises 
technical specification requirements to submit biennial reports every 
24-months instead of prior to March 1 of every other year. It also 
eliminates the requirements to notify the Nuclear Regulatory Commission 
(NRC) of exceeding environmental limits and changes to environmental 
permits such as the National Pollution Discharge Elimination System 
permit. The licensee's November 5, 1999, submittal proposed revising 
technical specifications dealing with eliminating notifying the NRC for 
exceeding limits of minor permits where there is no identifiable 
environmental or public health concerns and exceptional occurrences 
(unusual or important events, exceeding limit of relevant permits). 
Since additional information would be required to continue this part of 
the review, the licensee withdrew this portion of their original 
application dated November 5, 1999, and replaced it in its entirety 
with a supplemental letter dated September 28, 2000.
    Date of issuance: March 1, 2001.
    Effective date: Immediately, to be implemented within 120 days.
    Amendment No.: 55.
    Facility Operating License No. DPR-73: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: January 12, 2000 (65 FR 
1924). The September 28, 2000, supplemental letter replaced in its 
entirety the licensee's original application dated November 5, 1999. 
The supplement did not expand the scope of the original request, nor 
did it change the proposed no significant hazards consideration 
finding. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 1, 2001.
    No significant hazards consideration comments received: No.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone 
Nuclear Power Station, Unit No. 3, New London County, Connecticut

    Date of application for amendment: June 30, 2000, as supplemented 
on September 22 and November 20, 2000; and January 26 and February 1, 
2001.
    Brief description of amendment: This amendment changes the 
Millstone Nuclear Power Station, Unit No. 3 licensing basis. The 
amendment authorizes changes to the Final Safety Analysis Report (FSAR) 
regarding the installation of a new sump pump system in the engineered 
safety features building.
    Date of issuance: February 26, 2001.
    Effective date: As of the date of issuance and shall be implemented

[[Page 15935]]

within 30 days from the date of issuance.
    Amendment No.: 195.
    Facility Operating License No. NPF-49: Amendment authorizes changes 
to the FSAR.
    Date of initial notice in Federal Register: October 18, 2000 (65 FR 
62388).
    The September 22 and November 20, 2000, and January 26 and February 
1, 2001, letters provided clarifying information that did not change 
the scope of the amendment or the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 26, 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 29, 1999, as 
supplemented November 10 and December 15, 2000.
    Brief description of amendment: The amendment revises the Kewaunee 
Nuclear Plant Technical Specifications to incorporate requested changes 
per Generic Letter 99-02, ``Laboratory Testing of Nuclear-Grade 
Activated Charcoal,'' dated June 3, 1999.
    Date of issuance: February 28, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 152.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 13, 2000 (65 
FR 77921).
    The supplemental information contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 28, 2001.
    No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: July 28, 2000, as supplemented by letter 
dated December 14, 2000.
    Brief description of amendment: The amendment revises Sections 
2.1.4, 3.1, 3.17, Table 3-13, Table 3-14, and associated Bases of the 
Fort Calhoun Station Technical Specifications to allow the installation 
of ABB Combustion Engineering leak tight sleeves as an alternative tube 
repair method to plugging defective steam generator tubes.
    Date of issuance: March 1, 2001.
    Effective date: March 1, 2001, and shall be implemented within 30 
days from the date of issuance.
    Amendment No.: 195.
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 18, 2000 (65 FR 
62388).
    The December 14, 2000, supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noticed, and did not change the staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 1, 2001.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of application for amendments: November 30, 2000.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 5.5.14, ``Technical Specifications (TS) Bases 
Control Program'' to reflect the changes made to 10 CFR 50.59 as 
published in the Federal Register on October 4, 1999 (Volume 64, Number 
191, ``Changes, Tests, and Experiments,'' pages 53582 through 53617). A 
conforming change is made to TS 5.5.14 to replace the word ``involve'' 
with the word ``require,'' as it applies to changes to the TS Bases 
without prior NRC approval.
    Date of issuance: March 2, 2001.
    Effective date: March 2, 2001, and shall be implemented within 60 
days from the date of issuance.
    Amendment Nos.: Unit 1-145; Unit 2-144
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81928)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 2, 2001.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of application for amendments: December 6, 2000.
    Brief description of amendments: The amendments revised Section 5.0 
of the Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2 Technical 
Specifications to change management titles from (a) ``Vice President, 
Diablo Canyon Operations and Plant Manager'' to ``plant manager,'' (b) 
``Senior Vice President and General Manager--Nuclear Power Generation'' 
to ``specified corporate officer,'' (c) ``Radiation Protection 
Director'' to ``radiation protection manager,'' and (d) ``Operations 
Director'' to ``operations manager.''
    Date of issuance: March 7, 2001.
    Effective date: March 7, 2001, and shall be implemented within 30 
days from the date of issuance.
    Amendment Nos.: Unit 1-146; Unit 2-145.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7685).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 7, 2001.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of application for amendments: February 29, 2000 (submitted by 
PP&L, Inc., the licensee before July 1, 2000).
    Brief description of amendments: The amendments incorporated a 
reference to Supplement 3 ``Application Enhancements'' for the approved 
Topical Report PL-NF-90-001-A, ``Application of Reactor Analysis 
Methods for BWR [Boiling Water Reactor] Design and Analysis,'' into TS 
5.6.5, Core Operating Limits Report.
    Date of issuance: February 28, 2001.
    Effective date: As of date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 189 and 163.
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 18, 2000 (65 FR 
62390).
    The Commission's related evaluation of the amendments is contained 
in a

[[Page 15936]]

Safety Evaluation dated February 28, 2001.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric 
Station, Unit 2, Luzerne County, Pennsylvania

    Date of application for amendment: March 20, 2000 (submitted by 
PP&L, Inc., the licensee before July 1, 2000), as supplemented December 
1, 2000, and January 22, 2001 (submitted by PPL Susquehanna, LLC, the 
licensee on and after July 1, 2000).
    Brief description of amendment: The amendment revised the minimum 
critical power ratio safety limits.
    Date of issuance: March 6, 2001.
    Effective date: As of date of issuance and shall be implemented 
upon startup following the Unit 2 tenth refueling and inspection 
outage.
    Amendment Nos.: 164.
    Facility Operating License No. NPF-22. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 13, 2000 (65 
FR 77924).
    The supplemental letters provided additional information but did 
not change the initial no significant hazards consideration 
determination or expand the amendment beyond the scope of the initial 
notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 6, 2001.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of application for amendments: November 3, 2000, as 
supplemented February 1, 2001.
    Brief description of amendments: The amendments revise Technical 
Specification 5.5.11, ``Technical Specification Bases Control 
Program,'' to provide consistency with the changes to 10 CFR 50.59 
which were published in the Federal Register (64 FR 53582) on October 
4, 1999.
    Date of issuance: March 6, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 224 and 165.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 13, 2000 (65 
FR 77925).
    The supplement dated February 1, 2000, provided clarifying 
information that did not change the scope of the November 3, 2000, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 6, 2001.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of application for amendments: November 16, 2000, as 
supplemented on January 11, 2001.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) 5.5.14, ``Technical Specification Bases 
Control Program'' to provide consistency with the changes to 10 CFR 
50.59 as published in the Federal Register (64 FR 53582) dated October 
4, 1999. Specifically, the amendments remove the term ``unreviewed 
safety question'' from TS 5.5.14.b.2. In addition, two editorial 
corrections are also made on page 5.5-18.
    Date of issuance: March 1, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 118 and 96.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 13, 2000 (65 
FR 77927).
    The supplemental letter dated January 11, 2001, provided clarifying 
information that did not change the scope of the November 16, 2000, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 1, 2001.
    No significant hazards consideration comments received: No.

STP Nuclear Operating Company, Docket No. 50-499, South Texas Project, 
Unit 2, Matagorda County, Texas

    Date of amendment request: February 21, 2000, as supplemented by 
letters dated January 24 and 30, and February 28, 2001. The January 24 
and 30, and February 28, 2001 letters, provided additional clarifying 
information that was within the scope of the original application and 
Federal Register notice and did not change the staff's initial proposed 
no significant hazards consideration.
    Brief description of amendments: The Amendment revises the 
Technical Specifications (TSs) approving the application of the 3-volt 
repair criteria to the methodology for repair of steam generator (SG) 
tubes. The new criteria will apply for Unit 2 Cycle 9 only.
    Date of issuance: March 8, 2001.
    Effective date: The Amendment is effective on the date of issuance.
    Amendment No.: 114.
    Facility Operating License No. NPF-80: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 22, 2000 (65 FR 
15386).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 8, 2001.
    No significant hazards consideration comments received: No.

TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: September 6, 2000, as supplemented by 
letters dated December 14, 2000, and January 25, 2001.
    Brief description of amendments: The amendment changes Comanche 
Peak Electric Station (CPSES), Units 1 and 2, Technical Specification 
(TS) 5.5.9, ``Steam Generator (SG) Tube Surveillance Program,'' to 
permit installation of laser welded tubes sleeves in CPSES Unit 1 steam 
generator as an alternative to plugging defective tubes, and TS 5.6.10, 
``Steam Generator Tube Inspection Report,'' is revised to address 
reporting requirements for repaired tubes. Also an editorial correction 
is made to Table 5.5-2.
    Date of issuance: February 20, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 83 and 83.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 1, 2000 (65 FR 
65350).
    The supplemental letters dated December 14, 2000, and January 25, 
2001, provided additional information that clarified the application, 
did not

[[Page 15937]]

expand the scope of the application, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 20, 2001.
    No significant hazards consideration comments received: No.

TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: December 6, 2000.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 5.5.14, ``Technical Specifications (TS) Bases 
Control Program'' and TS 5.5.17, ``Technical Requirements Manual 
(TRM)'' to reflect the changes made to 10 CFR 50.59 as published in the 
Federal Register on October 4, 1999 (Volume 64, Number 191, ``Changes, 
Tests, and Experiments,'' pages 53582 through 53617). A conforming 
change is made to TS 5.5.14 and 5.5.17 to replace the word ``involve'' 
with the word ``require,'' as it applies to changes to the TS Bases or 
TRM without prior Nuclear Regulatory Commission approval.
    Date of issuance: March 5, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 84 and 84.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 10, 2001 (66 FR 
2024).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 5, 2001.
    No significant hazards consideration comments received: No.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 5, 2001.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia

    Date of application for amendments: March 29, 2000, as supplemented 
December 6, 2000, and March 1, 2001.
    Brief Description of amendments: These amendments revise TS 
Sections 3.19 and 4.1. The changes specify the requirements for two 
redundant trains of bottled air, specify remedial actions when one 
train or both trains are inoperable, eliminate the extension of the 
allowed outage and remedial action time of 8 hours to 24 hours 
currently permitted by TS 3.19.B, specify remedial actions for an 
inoperable control room pressure boundary, and include additional 
surveillance testing requirements. The Bases sections for TS 3.19 and 
TS 4.1 are revised for consistency with the respective TS.
    Date of issuance: March 9, 2001.
    Effective date: March 9, 2001.
    Amendment Nos.: 223 and 223.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: August 9, 2000 (65 FR 
48761). The December 6, 2000, and March 1, 2001, supplements contained 
clarifying information only, and did not change the initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 9, 2001.
    No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: December 7, 2000.
    Brief description of amendment: The amendment deletes Technical 
Specifications (TS) Section 5.5.3, ``Post Accident Sampling System,'' 
for Wolf Creek Generating Station and thereby eliminates the 
requirements to have and maintain the post-accident sampling system. 
The amendment also revises TS Section 5.5.2, ``Primary Coolant Sources 
Outside Containment,'' to reflect the elimination of PASS.
    Date of issuance: March 2, 2001.
    Effective date: March 2, 2001, and shall be implemented on or 
before December 1, 2001.
    Amendment No.: 137.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 10, 2001 (66 FR 
2026).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 2, 2001.
    No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: December 8, 2000.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 5.5.14, ``Technical Specifications (TS) Bases 
Control Program'' to reflect the changes made to 10 CFR 50.59 as 
published in the Federal Register on October 4, 1999 (Volume 64, Number 
191, ``Changes, Tests, and Experiments,'' pages 53582 through 53617). A 
conforming change is made to TS 5.5.14 to replace the word ``involves'' 
with the word ``requires,'' as it applies to changes to the TS Bases 
without prior NRC approval.
    Date of issuance: March 2, 2001.
    Effective date: March 2, 2001, and shall be implemented within 60 
days from the date of issuance.
    Amendment No.: 138.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 10, 2001 (66 FR 
2027).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 2, 2001.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Sinificant 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

[[Page 15938]]

of the licensee's application and of the Commission's proposed 
determination of no significant hazards consideration. The Commission 
has provided a reasonable opportunity for the public to comment, using 
its best efforts to make available to the public means of communication 
for the public to respond quickly, and in the case of telephone 
comments, the comments have been recorded or transcribed as appropriate 
and the licensee has been informed of the public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, 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 20, 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

[[Page 15939]]

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).

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: February 1, 2001.
    Description of amendment request: The amendment removes 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 issuance: March 1, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 116.
    Facility Operating License No. (DPR-22): Amendment revises the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (66 FR 10535, dated February 15, 2001). The notice 
provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. No comments have been received. The notice also 
provided for an opportunity to request a hearing by March 19, 2001, but 
indicated that if the Commission makes a final NSHC determination, any 
such hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a Safety Evaluation dated March 1, 2001.

    Attorney for licensee: Jay Silberg, Esq., at Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

    Dated at Rockville, Maryland this 13th day of March 2001.
    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-6732 Filed 3-20-01; 8:45 am]
BILLING CODE 7590-01-P