[Federal Register Volume 66, Number 65 (Wednesday, April 4, 2001)]
[Notices]
[Pages 17962-17977]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-8101]


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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

[[Page 17963]]

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 March 12 through March 23, 2001. The last 
biweekly notice was published on March 21, 2001 (66 FR 15915).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Administrative 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852. The filing of requests for a 
hearing and petitions for leave to intervene is discussed below.
    By May 4, 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

[[Page 17964]]

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).

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of amendment request: March 1, 2001.
    Description of amendment request: The proposed amendment would 
increase the reactor core isolation cooling system surveillance test 
upper pressure limit from 1020 psig to 1045 psig.
    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 any accident previously 
evaluated.
    The Reactor Core Isolation Cooling (RCIC) System is designed to 
operate either automatically or manually following reactor pressure 
vessel (RPV) isolation accompanied by a loss of coolant flow from 
the feedwater system to provide adequate core cooling and control of 
RPV water level. The RCIC System is also designed to provide core 
cooling for a wide range of reactor pressures, from 150 pounds per 
square inch gage (psig) to 1200 psig. The proposed change to the 
Technical Specifications (TS) Section 3.5.3, ``RCIC System,'' 
Surveillance Requirement (SR) 3.5.3.3 to allow the RCIC system high 
pressure test to be performed at a higher reactor pressure (i.e., 
less than or equal to 1045 psig) is consistent with the current 
design and licensing basis for the RCIC system. The change to the 
upper pressure limit for the conduct of this SR will not adversely 
impact the performance characteristics of any structure, system, or 
component that is assumed to initiate a previously evaluated 
accident. Therefore, the proposed change will not result in an 
increase in the probability of an accident previously evaluated.
    The proposed change to the TS SR will not result in reduced 
performance or effectiveness of the reactor coolant pressure 
boundary and therefore will not have an adverse impact on any 
barriers. As such, the RCIC System will still be capable of 
performing its transient and accident mitigation function as assumed 
in the accident analysis. On this basis, the consequences of any 
accident previously evaluated are not affected by the proposed 
change.
    Based on the above, the proposed change does not involve a 
significant increase in the probability or consequences on any 
accident previously evaluated.
    (2) The proposed change would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    (2) The proposed change to the TS SR to allow the RCIC System 
high pressure test to be performed at a higher reactor pressure 
(i.e., less than or equal to 1045 psig) is consistent with the 
current design and licensing basis for the RCIC system. The proposed 
change will not change the method for performing the test and the 
revised test pressure is within the current operating design basis 
of the plant. Since the proposed test pressure is within the design 
basis for the reactor and the RCIC System, performing the SR at the 
new pressure will not prevent the RCIC System from performing its 
required function or result in a failure of the reactor coolant 
pressure boundary. As such, the proposed change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) The proposed change will not involve a significant reduction 
in the margin of safety.
    The RCIC System is designed to operate either automatically or 
manually following RPV isolation accompanied by a loss of coolant 
flow from the feedwater system to provide adequate core cooling and 
control of RPV water level. The RCIC System is also designed to 
provide core cooling for a wide range of reactor pressures, from 150 
psig to 1200 psig. The proposed change to TS SR 3.5.3.3 to allow the 
RCIC System high pressure test to be performed at a higher pressure 
(i.e., less than or equal to 1045 psig) is consistent with the 
current design and licensing basis for the RCIC system. Since the 
test at the higher reactor pressure will continue to provide 
reasonable assurance that the RCIC System will perform its intended 
safety function when called upon during an accident or transient, 
the proposed 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: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW, Washington, DC 20036-5869.
    NRC Section Chief: Anthony J. Mendiola.

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of amendment request: December 29, 2000.
    Description of amendment request: The proposed amendment would 
revise the offsite power sources identified in Technical Specification 
(TS) 3.7.A.3 to remove one listed source and add a different source. In 
addition, the bases would be revised to reflect the availability of the 
offsite sources and also be revised administratively for minor changes.
    Basis for proposed no significant hazardsconsideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed amendment does not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to the Technical Specifications involves the 
removal of the 230 kV line from TS 3.7.A.3.a and addition of the 69 
kV Sands Point line in its place. This ensures that two active 
offsite power sources are connected to the plant to support plant 
operation. Two 230 kV lines (considered one active power source), 
34.5 kV line Q121 and the 69 kV Sands Point line will be normally 
maintained as active sources. Utility system operators will connect 
the Z52 line under certain grid conditions, which provides a backup 
to the normally available sources and improves offsite power 
reliability.

[[Page 17965]]

    Since the number of normally available active offsite power 
sources is maintained at three, the probability of occurrence of a 
loss of offsite power is not adversely impacted and, therefore, the 
proposed change does not significantly increase the probability of 
occurrence of an accident previously evaluated.
    Voltage analyses and voltage regulation studies have reviewed 
various degraded grid and plant operating scenarios. Degraded grid 
studies have considered single contingency events, such as loss of 
transmission lines or transformers, and various plant outages, 
minimum Technical Specification conditions and local system 
blackouts. Voltage regulation studies have considered various plant 
operating modes such as normal operation, accident motor starting 
and loading and shutdown conditions. The studies have concluded that 
adequate voltage will be available to safety-related electrical 
loads under all of the plant operating modes. Therefore, the 
consequences of any accident will not change since the operation of 
safety-related systems are not affected by the change in offsite 
power sources. For a complete loss of offsite power, the standby 
diesel generators are relied upon to provide electrical power to 
safety systems. Since the proposed change to Technical Specification 
3.7.A.3 does not affect the operability of the standby diesel 
generators, the consequences of a loss of offsite power are 
unchanged. Therefore, there is no significant increase in the 
consequences of an accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed change to Technical Specification 3.7.A.3 adds a 
new offsite power source (69 kV Sands Point line) and removes the 
second 230 kV line as an additional source since it is not separate 
from the other 230 kV line. Therefore, the proposed change involves 
the availability of offsite power connections to the plant. A 
potential loss of offsite power is already evaluated and the standby 
diesel generators are relied upon to provide power to accident 
mitigation and safe shutdown equipment in the event all offsite 
power is lost. Therefore, the proposed change to available offsite 
power sources does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    (3) Involve a significant reduction in a margin of safety.
    Technical Specification 3.7.A.3 requires two offsite power lines 
to be fully operational for plant start-up and operation. The 
proposed change to the Technical Specifications involves the removal 
of the 230 kV line from TS 3.7.A.3.a and addition of the 69 kV Sands 
Point line in its place. This ensures that two active offsite power 
sources are connected to the plant to support plant operation. Two 
230 kV lines (considered one active power source), 34.5 kV line Q121 
and the 69 kV Sands Point line will be normally maintained as active 
sources. The removal of the second 230 kV line from TS 3.7.A.3.a has 
no impact on available active sources since the [sic] both 230 kV 
lines are routed on the same towers and are therefore considered as 
a single active source. Analyses have concluded that sufficient 
capacity and capability of offsite power is available when TS 
3.7.A.2 and 3.7.A.3 are met. 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: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Marsha Gamberoni.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: February 27, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.3, ``Engineered Safety 
Features,'' and TS 3.7, ``Auxiliary Electrical Systems,'' to change the 
mode applicability for certain systems from the point of time when the 
reactor is made critical to when the average reactor coolant 
temperature is heated above 350  deg.F. The amendment would also change 
the associated action that must be taken when the TS conditions cannot 
be met to require a plant cooldown to below 350  deg.F. In addition, 
the associated TS statements that incorrectly refer to ``power 
operation'' and ``normal reactor operation'' for these TSs are proposed 
to be corrected. The proposed amendment would also revise the 
applicable TS Bases sections and make some minor formatting and 
editorial 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, which is presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    The proposed changes consist of revisions to the TS requirements 
for certain safeguards equipment and associated auxiliary electrical 
equipment to reflect the requirements of the steam line break 
analyses. The result of these changes will be that these safeguards 
systems will be required to be operable for additional plant 
conditions (with average reactor coolant temperature above 350 
deg.F and the reactor not critical). These operability requirements 
for the safeguards equipment meet the assumptions utilized in the 
IP2 [Indian Point Unit 2] safety analyses and, therefore, will not 
result in a change in the consequences of the accident analyses.
    Additionally, the affected safeguards equipment is not an 
initiator for any accident previously analyzed for IP2. The proposed 
changes do not result in a change to the design or operation of the 
safeguards equipment but extends the plant conditions under which 
this equipment will be required to be operable.
    Therefore, there is no increase in the probability or in the 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    The proposed changes involve revising the TS applicability for 
certain safeguards equipment and associated auxiliary electrical 
systems to require this equipment to be operable with average 
reactor coolant temperature above 350  deg.F. The proposed changes 
do not involve a change to the design or operation of any plant 
system or equipment. The result of the proposed change is an 
increased range of operating conditions under which the safeguards 
equipment will be required to be operable. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The proposed changes reflect the assumptions for safeguards 
equipment operability assumed in the steam line break accident 
analyses. These changes ensure that the affected TS reflect the 
assumptions of the safety analyses but do not result in a change to 
any of the safety analyses or any 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: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Section Chief: Marsha Gamberoni.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of amendment request: January 30, 2001.
    Description of amendment request: The proposed amendment would make 
two changes to reporting requirements in Facility Operating License 
DPR-20. First, the requirement in Section 2.C.(3)b that ``All changes 
in the approved [Fire Protection] program

[[Page 17966]]

shall be reported annually, along with the FSAR [Final Safety Analysis 
Report] revision * * *'' would be changed to state ``All changes to the 
approved program shall be reported along with the FSAR revision as 
required by 10 CFR 50.71(e) * * *'' Secondly, a change would be made to 
Section 2.F, which currently states:

    Except as otherwise provided in the Technical Specifications or 
Environmental Protection Plan, the licensee shall report any 
violations of the requirements contained in Section 2.C of this 
license in the following manner: initial notification shall be made 
within 24 hours to the NRC Operations Center via the Emergency 
Notification System with written follow-up within 30 days in 
accordance with the procedures described in 50.73(b), (c), and (e).

The revised Section 2.F would state:

    The licensee shall report any violations of Section 2.C(1) of 
this license within 24 hours to the NRC Operations Center via the 
Emergency Notification System with written follow-up within 60 days 
in accordance with 10 CFR 50.73(b), (c), and (e).

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The following evaluation supports the finding that operation of 
the facility in accordance with the proposed changes would not:
    a. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Two changes are proposed; both deal solely with clarification of 
reporting requirements contained in the Facility Operating License. 
Since these changes have no effect on the physical plant or its 
operation, they cannot involve a significant increase in the 
probability or consequences of an accident previously evaluated. 
Therefore, operation of the facility in accordance with the proposed 
changes to the Facility Operating License would not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    b. create the possibility of a new or different kind of accident 
from any previously evaluated.
    Two changes are proposed; both deal solely with clarification of 
reporting requirements contained in the Facility Operating License. 
Since these changes have no effect on the physical plant or its 
operation, they cannot create the possibility of a new or different 
kind of accident from any previously evaluated.
    Therefore, operation of the facility in accordance with the 
proposed change to the Technical Specifications [sic, Facility 
Operating License] would not create the possibility of a new or 
different kind of accident from any previously evaluated.
    c. Involve a significant reduction in the margin of safety.
    Two changes are proposed; both deal solely with clarification of 
reporting requirements contained in the Facility Operating License. 
Since these changes have no effect on the physical plant or its 
operation, they cannot involve a significant reduction in a margin 
of safety.
    Therefore, the proposed change to the Technical Specifications 
[sic, Facility Operating License] would not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: February 20, 2001.
    Description of amendment request: The licensee is proposing to add 
a new chapter to the Columbia Generating Station Physical Security Plan 
pertaining to the Independent Spent Fuel Storage Installation (ISFSI) 
security requirements. Specifically, the licensee proposes the 
following changes regarding the ISFSI: (1) Illumination will be 
sufficient to permit adequate assessment of unauthorized penetrations 
or activities within the protected area, (2) personnel access will be 
controlled by a key and lock system administered by the security force, 
(3) personnel identification will be by visual identification using 
plant access picture badges and an ISFSI authorization list, (4) no 
vehicle barrier around the perimeter of the ISFSI, (5) response time 
for valid alarms that only needs to be sufficient to assess the 
situation and the further need for corrective actions, and (6) 
secondary power supply for alarm annunciator equipment and non-portable 
communications equipment will have secondary power from an 
uninterruptible power supply not in the vital area.
    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 individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of facility systems and the ability of plant personnel 
to mitigate those consequences.
    Confinement of all radioactive materials at the Energy Northwest 
ISFSI is provided by the required use of certified spent fuel 
storage casks in accordance with 10 CFR 72.214. The design objective 
of NRC certified spent fuel storage casks is to provide a 
confinement boundary that ensures there are no credible design basis 
events resulting in unacceptable radiological releases to the 
environment. In addition, these spent fuel storage casks are to be 
located within the confines of the Energy Northwest ISFSI which is 
designed as a protected area.
    Since the design objective of the spent fuel storage cask has 
not been altered, there is no increase in individual precursors of 
an accident and the probability of an evaluated accident is not 
increased. The spent fuel casks stored at the Energy Northwest ISFSI 
will be inside a new fenced protected area with access requirements, 
detection aids, alarm devices, communication requirements, and 
observational capabilities commensurate with the activity of passive 
dry cask spent fuel storage that meet or exceed the criteria 
specified under 10 CFR 73.51. Since the Energy Northwest ISFSI 
physical security program will provide a high degree of assurance 
that activities involving spent nuclear fuel do not constitute an 
unreasonable risk, 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 facility configuration, including changes in 
allowable modes of operation or the potential for new or different 
personnel errors.
    The proposed license amendment does not alter the design 
objective of NRC certified spent fuel storage casks. This license 
amendment request does not involve any modifications of the spent 
fuel storage casks or allowable modes of operation and no potential 
exists for the creation of personnel errors that might be new 
accident precursors. Thus, no new precursors of an accident are 
created and there is not a possibility of a new or different kind of 
accident.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Confinement of all radioactive materials and substantial 
physical protection of the spent nuclear fuel is accomplished by the 
required use of an NRC certified spent fuel storage cask as provided 
by Certificate of Compliance listed under 10 CFR 72.214. The spent 
fuel casks stored in the Energy

[[Page 17967]]

Northwest ISFSI will be inside a new protected area with access 
requirements, detection aids, alarm devices, communication 
requirements, and observational capabilities that meet or exceed the 
criteria specified in 10 CFR 73.51 for spent fuel stored under a 
specific license.
    Since the Energy Northwest ISFSI will provide a high degree of 
assurance that activities involving spent nuclear fuel do not 
constitute an unreasonable risk, there is not 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.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of amendment request: March 12, 2001.
    Description of amendment request: The proposed amendment would 
modify Technical Specification 3.1.3.4a to reduce the minimum 
requirement for average reactor coolant temperature during the rod 
cluster control assembly (RCCA) drop test from greater than or equal to 
541 deg.F to greater than or equal to 500 deg.F. RCCA drop tests are 
required prior to reactor criticality: (1) For all rods, following each 
removal of the reactor vessel head, (2) for specifically affected 
individual rods, following maintenance work which could affect the drop 
times of those specific rods, and (3) at least every 18 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The probability of occurrence of an accident previously 
evaluated for Turkey Point is not altered by the proposed amendments 
to the Technical Specifications. The proposed changes do not impact 
the integrity of the reactor coolant system pressure boundary (i.e., 
no change in operating pressure, materials, seismic loading, etc.) 
and therefore do not increase the potential for the occurrence of a 
loss of coolant accident (LOCA). The changes do not make any 
physical changes to the facility design, material, or construction 
standards. The probability of any design basis accident (DBA) is not 
affected by these changes, nor are the consequences of any DBA 
affected by these changes. The proposed changes are not considered 
to be an initiator or contributor to any accident currently 
evaluated in the Turkey Point Updated Final Safety Analysis Report 
(UFSAR). Based on the above, Florida Power and Light Company 
concludes that the proposed amendments do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    The Rod Cluster Control Assembly (RCCA) drop test is routinely 
performed each refueling. Decreasing the test temperature will not 
create the possibility of a new or different accident. The proposed 
test conditions remain bounded by the analysis of record since the 
RCCA drop time assumption in the UFSAR accident analysis will not be 
changed. Since no new failure modes are associated with the proposed 
changes, the proposed amendments do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    These Technical Specification changes do not involve a 
significant reduction in margin since the acceptance criterion for 
RCCA drop time will not change. The proposed changes will reduce the 
minimum RCCA rod drop test temperature from greater than or equal to 
541 deg.F to greater than or equal to 500 deg.F. This will slightly 
increase the test drop time, but will be well within the current 
Technical Specifications limit of 2.4 seconds. Therefore, the margin 
to safety as defined by Technical Specifications acceptance 
criterion is not impacted by the proposed amendments.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: March 7, 2001.
    Description of amendment request: The licensee requests allowing a 
one-time interval extension for the Crystal River Unit 3 (CR-3) Type A, 
Integrated Leakage Rate Test (ILRT) for no more than 6 years.
    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 revision to the CR-3 [Improved Technical 
Specifications] ITS adds a one-time extension to the current 
interval for Type A testing. The current test interval of 10 years, 
would be extended on a one-time basis to 16 years from the last Type 
A test. The proposed extension to Type A testing cannot increase the 
probability of an accident previously evaluated since the 
containment Type A testing extension is not a modification to plant 
systems, nor a change to plant operation that could initiate an 
accident. The proposed extension to Type A testing does not involve 
a significant increase in the consequences of an accident since 
research documented in NUREG-1493 found that, generically, very few 
potential containment leakage paths fail to be identified by Type B 
and C tests. In fact, an analysis of 144 ILRT results, including 23 
failures, found that no failures were due to containment liner 
breach. The NUREG concluded that reducing the Type A (ILRT) testing 
frequency to one per twenty years would lead to an imperceptible 
increase in risk. CR-3 provides a high degree of assurance through 
testing and inspection that the containment will not degrade in a 
manner detectable only by Type A testing. Inspections required by 
the Maintenance Rule and American Society of Mechanical Engineers 
(ASME) code are performed in order to identify indications of 
containment degradation that could affect leak tightness. Type B and 
C testing required by the CR-3 ITS will identify any containment 
opening, such as valves, that would otherwise be detected by the 
Type A tests. These factors show that a CR-3 Type A test extension 
will not represent a significant increase in the consequences of an 
accident.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously analyzed.
    The proposed extension to Type A testing cannot create the 
possibility of a new or different type of accident since there are 
no physical changes being made to the plant. There are no changes to 
the operation of the plant that could introduce a new failure mode 
creating the possibility of a new or different kind of accident.
    3. Does not involve a significant reduction in the margin of 
safety.
    The proposed extension to Type A testing will not significantly 
reduce the margin of safety. The NUREG-1493 generic study of the 
effects of extending containment leakage

[[Page 17968]]

testing found that a 20 year extension in Type A leakage testing 
resulted in an imperceptible increase in risk to the public. NUREG-
1493 found that, generically, the design containment leakage rate 
contributes a very small amount to the individual risk, and that the 
decrease in Type A testing frequency would have a minimal affect on 
this risk since most potential leakage paths are detected by Type C 
testing.

    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.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point 
Nuclear Station Unit No. 1, Oswego County, New York

    Date of amendment request: March 6, 2001.
    Description of amendment request: The licensee proposed to amend 
the unit's Technical Specifications (TSs), Section 3.4.4, ``Emergency 
Ventilation System [EVS],'' and Section 3.4.5, ``Control Room Air 
Treatment [CRAT] System,'' to require testing consistent with American 
Society for Testing and Materials (ASTM) Standard D3803-1989 (currently 
the American National Standards Institute (ANSI) standard N510-1980 is 
specified). Concurrently, the licensee proposed to change the charcoal 
bed testing efficiency of the EVS and CRAT from 90 percent to 95 
percent, and requiring the pressure drop across the CRAT System high 
efficiency particulate air (HEPA) filters and charcoal adsorber banks 
to be demonstrated to be less than 1.5 inches of water. The licensee's 
application for amendment is a response to the NRC's Generic Letter 
(GL) 99-02, ``Laboratory Testing of Nuclear-Grade Activated Charcoal.'' 
The associated licensee-controlled TS Bases document would also be 
changed to reflect these TS changes.
    The staff had previously published notices (65 FR 9009, February 
23, 2000, and 65 FR 56955, September 20, 2000) for the licensee's 
November 30, 1999, and August 15, 2000, submittals. The licensee's 
March 6, 2001, submittal supersedes the original submittals in their 
entirety. Hence this notice also supersedes the previous two notices. 
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 
reviewed the licensee's analysis against the three standards of 10 CFR 
50.92(c). The NRC staff's analysis 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 TS changes will require 
that the charcoal filter beds be tested in accordance with an NRC-
approved standard (i.e., ASTM D3803-1989), and to improved 
acceptance criteria. The CRAT and EVS do not involve initiators or 
precursors to an accident previously evaluated, as these systems 
perform only mitigative functions in response to an accident. 
Failure of these systems would result in inability or decreased 
ability to perform their mitigative functions, but would not 
increase the probability of an accident. The proposed testing 
requirements would improve the performance of these systems, and 
would not have any effect in reducing their design functions. 
Therefore, the probability and consequences of an accident 
previously evaluated will not be increased by the proposed TS 
changes.
    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. The proposed TS change will only revise the 
testing requirements. These changes will not involve placing the 
systems in new configurations or operating the systems in different 
manners. Therefore, the proposed changes 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, 
operation procedure, or analysis methodology is changed, proposed TS 
changes will not adversely affect the performance characteristics of 
the CRAT or EVS, nor will they affect the ability of the systems to 
perform their intended functions. 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.

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

    Date of amendment requests: June 8, 2000, as supplemented by letter 
dated January 4, 2001.
    Description of amendment requests: The proposed license amendments 
would revise Section 3.5.5, ``Emergency Core Cooling Systems--Seal 
Injection Flow,'' of the improved Technical Specifications to replace 
the description of the seal injection flow with a description 
consistent with the method used to establish and verify reactor coolant 
pump seal injection flow limits and the method used to calculate the 
seal injection flow in the safety analyses for the Diablo Canyon 
Nuclear Power Plant, Unit Nos. 1 and 2.
    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 emergency core cooling system (ECCS) analyses model the 
reactor coolant pump (RCP) seal injection flow path as a hydraulic 
flow resistance. This proposed change clarifies that RCP seal flow 
is a function of system conditions rather than specifying an actual 
flow rate. The seal flow rate can vary during operation, but the 
hydraulic flow resistance is fixed by positioning the manual seal 
injection throttle valves. The resistance does not change if the 
valve adjustments are not changed. Thus, RCP seal flow variation due 
to changing reactor coolant system (RCS) back pressure following a 
loss of coolant accident (LOCA) is explicitly determined as a result 
of modeling the RCP seal injection flow path resistance.
    The proposed improved Technical Specification change is only a 
clarification and does not impact the way the RCP seal flow is 
established and thus cannot affect RCP seal integrity. The seal flow 
resistance otherwise only affects ECCS flow. Since ECCS flow occurs 
after an accident the proposed change cannot impact the probability 
of an accident.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. The change continues to ensure that the assumed ECCS flow 
is available. Therefore, the proposed change does not involve a 
significant increase in the 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.

[[Page 17969]]

    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. Since the change continues to ensure that the assumed ECCS 
flow is available, no new accident scenarios, transient precursors, 
failure mechanisms, or limiting single failures are introduced. 
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 
any analyzed event. There will be no effect on the manner in which 
safety limits or limiting safety system settings are determined nor 
will there be any effect on those plant systems necessary to assure 
the accomplishment of protection functions. Since the change 
continues to ensure the assumed ECCS flow is available, there will 
be no impact on any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment requests: February 20, 2001. This application 
supersedes the June 19, 2000, application and supplement dated 
September 12, 2000 (published in the Federal Register on October 4, 
2000 [65 FR59223]).
    Description of amendment requests: The proposed license amendments 
would revise Sections 5.5.9, ``Steam Generator (SG) Tube Surveillance 
Program'' and 5.6.10, ``SG Tube Inspection Report,'' of the Diablo 
Canyon Power Plant, Unit Nos. 1 and 2 Technical Specifications (TS), to 
add new surveillance and reporting requirements associated with SG tube 
inspection and repair. The new requirements establish alternate repair 
criteria for axial primary water stress corrosion cracking at dented 
tube support plate intersections.
    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.
    Examination of crack morphology for primary water stress 
corrosion cracking (PWSCC) at dented intersections has been found to 
show one or two microcracks well aligned with only a few uncorroded 
ligaments and little or no other inside diameter axial cracking at 
the intersection. This relatively simple morphology is conducive to 
obtaining good accuracy in nondestructive examination (NDE) sizing 
of these indications. Accordingly, alternate repair criteria (ARC) 
are established based on crack length and average and maximum depth 
within the thickness of the tube support plate (TSP).
    The application of the ARC requires a Monte Carlo condition 
monitoring assessment to determine the as-found condition of the 
tubing. The condition monitoring analysis described in WCAP-15573, 
Revision 0, is consistent with NRC Generic Letter 95-05 
requirements.
    The application of the ARC requires a Monte Carlo operational 
assessment to determine the need for tube repair. The repair bases 
are obtained by projecting the crack profile to the end of the next 
operating cycle and determining the burst pressure and leakage for 
the projected profile using Monte Carlo analysis techniques 
described in WCAP-15573, Revision 0. The burst pressure and leakage 
are compared to the requirements in WCAP-15573, Revision 0. Separate 
analyses are required for the total crack length and the length 
outside the TSP due to differences in requirements. If the projected 
end of cycle (EOC) requirements are satisfied, the tube will be left 
in service.
    A steam generator (SG) tube rupture event is one of a number of 
design basis accidents that are analyzed as part of a plant's 
licensing basis. A single or multiple tube rupture event would not 
be expected in a SG in which the ARC has been applied. The ARC 
requires repair of any indication having a maximum crack depth 
greater than or equal to 40 percent outside the TSP, thus limiting 
the potential length of a deep crack outside the TSP at EOC 
conditions and providing margin against burst and leakage for free 
span indications.
    For other design basis accidents such as a main steam line 
break, main feed line break, control rod ejection, and locked 
reactor coolant pump motor, the tubes are assumed to retain their 
structural integrity.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Implementation of the proposed SG tube ARC does not introduce 
any significant changes to the plant design basis. A single or 
multiple tube rupture event would not be expected in a SG in which 
the ARC has been applied. Both condition monitoring and operational 
assessments are completed as part of the implementation of ARC to 
determine that structural and leakage margin exists prior to 
returning SGs to service following inspections. If the condition 
monitoring requirements are not satisfied for burst or leakage, the 
causal factors for EOC indications exceeding the expected values 
will be evaluated. The methodology and application of this ARC will 
continue to ensure that tube integrity is maintained during all 
plant conditions consistent with the requirements of Regulatory 
Guide (RG) 1.121 and Revision 1 of RG 1.83.
    In the analysis of a SG tube rupture event, a bounding primary-
to-secondary leakage rate equal to the operational leakage limits in 
the Technical Specifications (TS), plus the leak rate associated 
with the double-ended rupture of a single tube, is assumed. For 
other design basis accidents, the tubes are assumed to retain their 
structural integrity and exhibit primary-to-secondary leakage within 
the limits assumed in the current licensing basis accident analyses. 
Steam line break leakage rates from the proposed PWSCC ARC are 
combined with leakage rates from other approved ARC (i.e., voltage-
based ARC and W* ARC). The combined leakage rates will not exceed 
the limits assumed in the current licensing basis accident analyses.
    The 40 percent maximum depth repair limit for free span 
indications provides a very low likelihood of free span leakage 
under design basis or severe accident conditions. Leakage from 
indications inside the TSP is limited by the constraint of the TSP 
even under severe accident conditions, and leakage behavior in a 
severe accident would be similar to that found acceptable by the NRC 
under approved ARC for axial outside diameter stress corrosion 
cracking (ODSCC) at TSP intersections. Therefore, even under severe 
accident conditions, it is concluded that application of the 
proposed ARC for PWSCC at dented TSP locations results in a 
negligible difference in risk of a tube rupture or large leakage 
event, when compared to current 40 percent repair limits or 
previously approved ARC.
    Diablo Canyon Power Plant (DCPP) continues to implement a 
maximum operating condition leak rate limit of 150 gallons per day 
per SG to preclude the potential for excessive leakage during all 
plant conditions.
    The possibility of a new or different kind of accident from any 
previously evaluated is not created because SG tube integrity is 
maintained by inservice inspection, condition monitoring, 
operational assessment, tube repair, and primary-to-secondary 
leakage monitoring.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Tube repair limits provide reasonable assurance that tubes 
accepted for continued service without repair will exhibit adequate 
tube structural and leakage integrity during subsequent plant 
operation. The implementation of the proposed ARC is demonstrated to 
maintain SG tube integrity consistent with the criteria of draft NRC 
Regulatory Guide 1.121. The guidelines of RG 1.121 describe a method 
acceptable to the NRC staff for meeting General Design Criteria

[[Page 17970]]

(GDC) 2, 4, 14, 15, 31, and 32 by ensuring the probability or the 
consequences of SG tube rupture remain within acceptable limits. 
This is accomplished by determining the limiting conditions of 
degradation of SG tubing, for which tubes with unacceptable cracking 
should be removed from service.
    Upon implementation of the proposed ARC, even under the worst-
case conditions, the occurrence of PWSCC at the tube support plate 
elevations is not expected to lead to a SG tube rupture event during 
normal or faulted plant conditions. The ARC involves a computational 
assessment to be completed for each indication left in service 
ensuring that performance criteria for tube integrity and leak 
tightness are met until the next scheduled outage.
    As discussed below, certain tubes are excluded from application 
of ARC. Existing tube integrity requirements apply to these tubes, 
and the margin of safety is not reduced.
    In addressing the combined loading effects of a loss-of-coolant 
(LOCA) and safe shutdown earthquake (SSE) on the SGs (as required by 
GDC 2), the potential exists for yielding of the TSP in the vicinity 
of the wedge groups, accompanied by deformation of tubes and a 
subsequent postulated in-leakage. Tube deformation could lead to 
opening of pre-existing tight through wall cracks, resulting in 
secondary to primary in-leakage following the event, which could 
have an adverse affect on the Final Safety Analysis Report (FSAR) 
results. Based on a DCPP analysis of LOCA and SSE, SG tubes located 
in wedge region exclusion zones are susceptible to deformation, and 
are excluded from application of ARC.
    A DCPP tube stress analysis for feed line break (FLB)/steam line 
break (SLB) plus SSE loading determined that high bending stresses 
occur in certain SG tubes at the seventh TSP, because the stresses 
exceed the maximum imposed bending stress for existing test data 
(equal to approximately the lower tolerance limit yield stress). 
These tubes are located in rows 11 to 15 and 36 to 46, and are 
excluded from application of ARC.
    Tube intersections that contain TSP ligament cracking are also 
excluded from application of ARC.
    Based on the above, it is concluded that the proposed license 
amendment request does not result in a significant reduction in 
margin with respect to the plant safety analyses as defined in the 
FSAR or TS.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

Portland General Electric Company, et al., Docket No. 50-344, Trojan 
Nuclear Plant, Columbia County, Oregon

    Date of amendment request: March 6, 2001.
    Description of amendment request: The proposed amendment revises 
Section 5.0, ``Administrative Controls,'' of the Trojan Nuclear Plant 
(TNP or Trojan) Technical Specifications. The first change is 
associated with modification of the TNP organizational structure. 
Specifically, the position of Senior Vice President, Power Supply, will 
be eliminated and the position Trojan Site Executive and Plant General 
Manager will be divided into two separate positions: (1) Trojan Site 
Executive, and (2) General Manager, Trojan. The second change is 
associated with revising language used in the TNP Technical 
Specifications to conform with the language of the revised 10 CFR 
50.59. Phrases which included the wording ``unreviewed safety 
question'' and ``safety evaluation'' will be replaced with wording that 
will continue to conform to the requirements of the revised 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. The requested license amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    As described above [in the licensee's amendment request], the 
changes in management titles and reporting relationships are 
administrative in nature, and as concluded by the NRC in the 
discussion accompanying the final rule, the changes made for 
consistency with the new 10 CFR 50.59 are viewed as editorial in 
nature. As such, these proposed changes do not alter the intent of 
the Possession Only License, and do not modify the present plant 
systems or administrative controls necessary to preserve and protect 
the integrity of the nuclear fuel at the TNP. Since no plant systems 
or administrative controls are changed, the probability or 
consequences of accidents previously evaluated are unaffected. The 
General Manager, Trojan will be located at the site and will provide 
management attention to each of the functional areas in the TNP 
organization during decommissioning of the facility.
    2. The requested license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    As described above [in the licensee's amendment request], the 
changes in management titles and reporting relationships are 
administrative in nature, and as concluded by the NRC in the 
discussion accompanying the final rule, the changes made for 
consistency with the new 10 CFR 50.59 are viewed as editorial in 
nature. As such, these changes do not affect the manner in which 
systems and components are operated or maintained, and do not alter 
the intent of the Possession Only License. There are no new accident 
scenarios or failure modes created by the requested administrative/
editorial changes. Therefore, the requested changes do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The requested license amendment does not involve a 
significant reduction in a margin of safety.
    As described above [in the licensee's amendment request], the 
changes in management titles and reporting relationships are 
administrative in nature, and as concluded by the NRC in the 
discussion accompanying the final rule, the changes made for 
consistency with the new 10 CFR 50.59 are viewed as editorial in 
nature. As such, these changes do not affect the manner in which 
systems and components are operated or maintained, do not alter the 
intent of the Possession Only License, and do not adversely impact 
previously accepted margins of safety. Therefore, the requested 
amendment 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: Douglas R. Nichols, Esq., Portland General 
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
    NRC Section Chief: Robert A. Gramm.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant, 
Unit 1, Hamilton County, Tennessee

    Date of application for amendments: March 9, 2001 (TS 01-01).
    Brief description of amendments: The proposed amendment would 
change the Sequoyah Nuclear Plant (SQN) Technical Specification section 
on reactor core design (Section 5.3) by adding a provision for 
including a limited number of lead test assemblies in the core.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority, the licensee, has provided its analysis of the issue of no 
significant hazards consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the

[[Page 17971]]

probability or consequences of an accident previously evaluated.
    The lead test assemblies (LTAs) are identical to the other Mark-
BW fuel assemblies with the exception of the initial uranium 
isotopic composition change. This composition change does not effect 
the chemical properties or affect the thermal-hydraulic performance 
of the fuel. The change in composition does change the neutronic 
response of the fuel. However the operational behavior of the fuel 
is accurately predicted by the NRC approved methodologies used for 
reload core design and analysis as demonstrated in the Topical 
Report [Framatome Cogema Fuels Topical Report BAW-2328], and the 
successful operation during SQN Unit 2 Cycle 10. Therefore, the LTAs 
do not significantly increase the probability of accidents while in 
the reactor.
    A preliminary reload design analysis performed, based upon the 
tentative use of the LTAs in SQN Unit 1 operating Cycle 12 fuel load 
pattern, shows that the LTAs will not become the most limiting fuel 
assemblies in the core during the cycle. Additionally, the peak pin 
criteria will be analyzed for each reload pattern to ensure that the 
LTAs do not become the most limiting peak pin at any time during 
their residence in the core.
    The potential effects of the LTAs on plant operation and safety 
are evaluated for each reload core design. The key core safety 
analysis parameters are examined each cycle to ensure each parameter 
remains bounded by the more limiting values used in the safety 
analysis of record and that there is no increase in the probability 
of occurrence for any design basis accident described in the Final 
Safety Analysis Report (FSAR).
    The impacts of the LTAs on the radiological consequences for all 
postulated events have been evaluated. The total calculated source 
term and the source-term activity of isotopes, which significantly 
contribute to operator and off-site accident exposure levels, were 
shown to be less than standard fuel assemblies with the same burnup, 
therefore, it will not increase the consequences of any accident 
previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The fuel assembly design for the LTAs is identical to the 
standard fuel assemblies. The main difference between the LTAs and 
the production fuel is that the initial concentration of the 
U234 and U236 isotopes will be higher in the 
LTA fuel pellets than that typically found in standard fuel. These 
isotopic differences will not affect the chemical, mechanical, or 
thermal properties of the fuel pellet.
    The LTAs meet the same design criteria and licensing basis 
criteria as the standard fuel assemblies and were manufactured with 
the same processes. The LTA skeleton is identical to the standard 
skeleton, which ensures that the loadings associated with normal 
operation, seismic events, loss-of-coolant accident (LOCA) events, 
and shipping and handling are not affected.
    Pressure and temperature safety limits will be maintained the 
same as those for the current operating cycle, thus ensuring that 
the fuel will be maintained within the same range of safety 
parameters that form the basis for previous accident evaluations. No 
new performance requirements are being imposed on any system or 
component that exceed design criteria or cause the core to operate 
in excess of design basis operating limits. No credible scenario has 
been identified, which could jeopardize equipment that could cause 
or intensify an accident sequence or mitigate events. Therefore, the 
LTAs will not create the possibility of accidents or equipment 
malfunctions of a different type than previously evaluated while in 
the reactor.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The LTAs will not adversely affect reactor neutronic or thermal-
hydraulic performance. The LOCA acceptance criteria with LTAs 
installed in the core will continue to be met. The acceptance 
criteria for departure from nucleate boiling (DNB) events with the 
LTAs installed in the core will also continue to be met. Other 
acceptance criteria have also been demonstrated to remain within 
acceptable limits. The total calculated source-term activity and the 
source-term activity of isotopes, which significantly contribute to 
operator and off-site accident exposure levels of the LTAs, was 
determined to be less than that for the standard fuel assembly with 
the same burnup. All previously evaluated events remain bounding and 
valid. For these reasons, the proposed amendment does not involve a 
significant reduction in a margin of safety.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of amendment request: March 2, 2001.
    Description of amendment request: The proposed amendment would 
revise Watts Bar Nuclear Plant (WBN) Unit 1 Technical Specifications 
(TS) Section 5.6, ``TS Bases Control Program,'' to adopt NRC-approved 
Technical Specification Task Force (TSTF) item TSTF-364, Revision 0. 
TSTF-364 revises the Industry Standard TS consistent with the recent 
revision to 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:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change is an administrative modification of 
existing TS requirements for the TS Bases Control Program to 
reference changes pursuant to 10 CFR 50.59 rather than ``unreviewed 
safety question.'' This change has no affect on the current review 
and approval process for changes to the Final Safety Analyses Report 
[FSAR] and Bases. Changes to the TS Bases are still evaluated in 
accordance with 10 CFR 50.59. As such, there is no effect on 
initiators of analyzed events or assumed mitigation of accidents or 
transients. Therefore, the proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    B. The proposed amendment 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, add any new equipment, or require any existing equipment 
to be operated in a manner different from the present design. 
Therefore, the proposed amendment does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change is an administrative modification of 
existing TS requirements for the TS Bases Control Program to 
reference changes pursuant to 10 CFR 50.59 rather than ``unreviewed 
safety question''. This change has no affect on the current review 
process for changes to the FSAR and Bases, and will not reduce a 
margin of safety because it has no effect on any safety analyses 
assumptions. Changes to the TS Bases are still evaluated in 
accordance with 10 CFR 50.59. For these reasons, the proposed 
amendment 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

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

    Date of application request: February 16, 2001 (ULNRC-04390).

[[Page 17972]]

    Description of amendment request: The amendment would add the word 
``Senior'' to the title ``Vice President and Chief Nuclear Officer'' in 
paragraph c to Technical Specification 5.2.1, ``Onsite and Offsite 
Organizations.'' The new title would be ``Senior Vice President and 
Chief Nuclear Officer.''
    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 revises Callaway Plant management 
organization by changing the title, Vice President and Chief Nuclear 
Officer to Senior Vice President and Chief Nuclear Officer; creating 
Vice President-Nuclear, to add another corporate level of oversight 
for plant site activities and nuclear staff supervision; and 
centralizing the Operations, Operations Support, and Engineering 
functions under the Vice President-Nuclear. These are administrative 
changes. [The proposed change does not change any plant safety 
limit, plant operations, or the plant design related to any accident 
previously evaluated.]
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of 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 revises Callaway Plant management 
organization by changing the title, Vice President and Chief Nuclear 
Officer to Senior Vice President and Chief Nuclear Officer; creating 
the title Vice President-Nuclear, to add another corporate level of 
oversight for plant site activities and nuclear staff supervision; 
and centralizing the Operations, Operations Support, and Engineering 
functions under the Vice President-Nuclear. These are administrative 
changes. [The proposed change does not involve an initiator of an 
accident.]
    Therefore, the proposed revision will not create 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 proposed change revises Callaway Plant management 
organization by changing the title, Vice President and Chief Nuclear 
Officer to Senior Vice President and Chief Nuclear Officer; creating 
the title Vice President-Nuclear, to add another corporate level of 
oversight for plant site activities and nuclear staff supervision; 
and centralizing the Operations, Operations Support, and Engineering 
functions under the Vice President-Nuclear. These are administrative 
changes.
    Therefore, the proposed change to the Technical Specifications 
do not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, D.C. 20037.
    NRC Section Chief: Stephen Dembek.

Yankee Atomic Electric Co., Docket No. 50-29, Yankee Nuclear Power 
Station (YNPS) Franklin County, Massachusetts

    Date of amendment request: November 22, 2000.
    Description of amendment request: The requested amendment would 
relocate certain administrative requirements from the YNPS Defueled 
Technical Specifications to the YNPS Decommissioning Quality Assurance 
Program (YDQAP). Additional editorial changes to titles and 
designations are also proposed.
    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 administrative nature of the changes will not affect any 
important to safety systems or components or their mode of 
operation. Relocation of TS administrative Sections 6.5, 6.7 and 6.9 
to the YDQAP does not result in changes to either system design or 
operating strategies. Relocation of these administrative 
requirements to the YDQAP has no affect on accident initiators or 
mitigation. Therefore, the proposed administrative changes will 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 accident from any previously evaluated.
    The proposed changes do not modify plant operation, systems, or 
components. Relocation of TS administrative Sections 6.5, 6.7 and 
6.9 to the YDQAP does not affect any of the parameters or conditions 
that could contribute to the initiation of any accident. No new 
accident scenarios are created as a result of relocating the 
aforementioned administrative requirements to the YDQAP. In 
addition, no important to safety equipment or functions are altered 
as a result of this proposed change. Therefore, the proposed 
administrative changes will not create the possibility of a new or 
different accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The changes are administrative in nature involving the 
relocation of administrative requirements from one licensing 
document to another licensing document currently containing related 
requirements. Relocation of TS administrative Sections 6.5, 6.7 and 
6.9 to the YDQAP does not affect plant operation, systems, or 
components. The proposed administrative changes do not represent a 
change in initial conditions, system response time, or in any other 
parameter affecting the course of an accident analysis supporting 
the Bases of any Technical Specification. Therefore, the proposed 
administrative changes will not involve a significant reduction in 
the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One 
International Place, Boston, Massachusetts 02110-2624.
    NRC Section Chief: Stephen Dembek.

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.

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

    Date of amendment request: April 14, 2000, as supplemented by 
letters dated June 2, July 28, and December 1, 2000, and January 31, 
2001.
    Brief description of amendment request: The proposed amendment 
would change the surveillance requirements for laboratory testing of 
the charcoal adsorbers for the control

[[Page 17973]]

room, the spent fuel pool storage area and the safety injection pump 
rooms. In addition, the amendment would delete the laboratory testing 
requirements for the containment charcoal adsorbers. The changes comply 
with the guidance of Generic Letter 99-02, ``Laboratory Testing of 
Nuclear-Grade Activated Charcoal.''
    Date of publication of individual notice in Federal Register: March 
5, 2001 (66 FR 13355).
    Expiration date of individual notice: April 4, 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).

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of application for amendment: August 9, 2000, as supplemented 
February 22, 2001. The February 22, 2001, supplement provided 
additional clarifying information and did not change the initial 
proposed no significant hazards consideration determination or expand 
the amendment beyond the scope of the original notice.
    Brief description of amendment: The amendment approved a revision 
to the Updated Final Safety Analysis Report (UFSAR) to reflect a 
revised steam generator tube failure accident analysis which includes 
the dose resulting from the postulated post-accident steam release 
through the main steam safety valves. The existing radiological dose 
calculations described in the UFSAR do not account for this release.
    Date of issuance: March 9, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 230.
    Facility Operating License No. DPR-50. Amendment authorized UFSAR 
revision.
    Date of initial notice in Federal Register: October 18, 2000 (65 FR 
62382).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 9, 2001.
    No significant hazards consideration comments received: No.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of application for amendments: June 8, 2000, as supplemented 
by the letters of January 3 and March 13, 2001.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 5.6.5, ``Core Operating Limits Report,'' to add a 
methodology using the CASMO-4 and SIMULATE-3 Codes to the list of 
analytical methods used to determine core operating limits contained in 
TS 5.6.5.b. The amendments allow the use of the CASMO-4 and SIMULATE-3 
methodology to perform nuclear design calculations; however, as stated 
in the supplemental letter of January 3, 2001, the licensee agreed that 
the introduction of significantly different or new fuel designs will 
require further validation of the physics methods in CASMO-4/SIMULATE-3 
for application to Palo Verde Units 1, 2, and 3, and will require 
review by the NRC staff.
    Date of issuance: March 20, 2001.
    Effective date: March 20, 2001,and shall be implemented within 45 
days of the date of issuance, including putting the condition mentioned 
above on the use of the new methodology, that was given in the 
licensee's letter of March 13, 2001, in the Updated Final Safety 
Analysis Report for Palo Verde.
    Amendment Nos.: Unit 1-132, Unit 2-132, Unit 3-132.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 4, 2000 (65 FR 
59219).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 20, 2001.
    No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: January 27, 2000, as 
supplemented on June 15, 2000, and November 21, 2000.
    Brief description of amendments: The amendments revise Technical 
Specifications 3.9.3 and 3.9.4 by modifying the conditions of 
containment closure during core alterations, fuel handling and the loss 
of shutdown cooling. The amendments also revise the way the personnel 
air lock and the containment purge system are operated during 
maintenance activities on the shutdown cooling system.
    Date of issuance: March 12, 2001.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 242 and 216.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 8, 2000 (65 FR 
12288).
    The June 15, 2000, and November 21, 2000, submittals provided 
clarifying information that did not change the original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated March 12, 2001.
    No significant hazards consideration comments received: No.

[[Page 17974]]

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: September 14, 2000.
    Brief description of amendments: The amendments revise the reactor 
coolant heatup and cooldown curves in the Technical Specifications.
    Date of issuance: March 15, 2001.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 243 and 217.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 18, 2000 (65 FR 
62382).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated March 15, 2001.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: August 10, 2000.
    Brief description of amendment: This amendment revises Required 
Actions suspending operations involving reactivity additions and 
revises various Limiting Condition for Operation Notes precluding 
reduction in boron concentration.
    Date of issuance: March 14, 2001.
    Effective date: March 14, 2001.
    Amendment No. 190.
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 6, 2000 (65 
FR 54084).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 14, 2001.
    No significant hazards consideration comments received: No.
    Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van 
Buren County, Michigan
    Date of application for amendment: December 8, 2000.
    Brief description of amendment: The amendment deletes Technical 
Specification Section 5.5.3, ``Post Accident Sampling Program,'' for 
Palisades and thereby eliminates the requirements to have and maintain 
the post-accident sampling system for the plant.
    Date of issuance: March 7, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 193.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7679).
    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.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: September 28, 2000.
    Brief description of amendment: The amendment changes the Arkansas 
Nuclear One, Unit 1 technical specifications to revise the safety-
related 4160 Volt (V) bus loss-of-voltage and 480 V bus degraded 
voltage relay allowable values.
    Date of issuance: March 12, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 211.
    Facility Operating License No. DPR 51: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 13, 2000 (65 
FR 77918).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 12, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 21, 2000, as 
supplemented by letters dated June 29, September 1, October 26, and 
December 22, 2000, and February 22, 2001.
    Brief description of amendment: The amendment provides for a full-
scope implementation of the alternative source term, as described in 
NUREG-1465, ``Accident Source Terms for Light-Water Nuclear Power 
Plants,'' Regulatory Guide 1.183, ``Alternative Radiological Source 
Terms for Evaluating Design-Basis Accidents at Nuclear Power 
Reactors,'' and 10 CFR 50.67, ``Accident source term.''
    Date of issuance: March 14, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 145.
    Facility Operating License No. NPF-29: The amendment revises the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: March 22, 2000 (65 FR 
15380).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 14, 2001.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of application for amendments: February 29, 2000, as 
supplemented by letter dated January 11, 2001.
    Brief description of amendments: The amendments reduced the number 
of safety valves required for overpressure protection at Dresden, Unit 
2, by removing from Technical Specifications (TS) Section 3.6.E, the 
safety valve function and setpoint of the Target Rock safety/relief 
valve (SRV). The amendments also moved the remaining safety valve lift 
pressure setpoints from TS Section 3.6.E to TS Section 4.6.E, changed 
the number of required safety valves from nine to eight, and removed 
footnote ``c'' of Unit 3 TS Section 4.6.E.
    Date of issuance: March 23, 2001.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 184 and 179.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 21, 2001 (66 
FR 11055).
    The January 11, 2001, letter is within the scope of the original 
notice and did not change the original no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated March 23, 2001.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, 
Beaver Valley Power Station, Unit No. 1, Shippingport, Pennsylvania

    Date of application for amendment: July 21, 2000, as supplemented 
by letters dated December 1, and December 13, 2000, and January 29, 
2001.
    Brief description of amendment: This amendment approves revisions 
to the Main Steam Line Break (MSLB) design-basis accident dose 
consequence

[[Page 17975]]

analysis as documented in the Updated Final Safety Analysis Report 
(UFSAR) and a technical specification (TS) change. The changes to the 
MSLB accident dose consequence analysis include revisions to input 
parameter values and assumptions. The TS change reduces the limit on 
reactor coolant system specific activity in technical specification 3/
4.4.8. The revisions are in accordance with the methodology described 
in Nuclear Regulatory Commission Generic Letter 95-05, ``Voltage-Based 
Repair Criteria for Westinghouse Steam Generator tubes by Outside 
Diameter Stress Corrosion Cracking.''
    Date of issuance: March 12, 2001.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment No: 236.
    Facility Operating License No. DPR-66: Amendment revised the 
Technical Specifications and approved changes to the UFSAR.
    Date of initial notice in Federal Register: February 7, 2001 (66 FR 
9382).
    Information from the July 21, and December 13, 2000, letters was 
used for the staff's initial proposal to determine that the amendment 
request involves a no significant hazards consideration determination. 
The December 1, 2000, and January 29, 2001, letters provided 
supplemental information applicable to this amendment request but did 
not change the initial proposed no significant hazards consideration 
determination or expand the amendment request beyond the scope of the 
original notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 12, 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: December 11, 2000, as 
supplemented by letter dated February 15, 2001.
    Brief description of amendment: This amendment revised the existing 
Minimum Critical Power Ratio (MCPR) Safety Limit contained in Technical 
Specification 2.1.1.2 by increasing the limit for two recirculation 
loop operation from 1.09 to 1.10.
    Date of issuance: March 12, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 119.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 10, 2001 (66 FR 
2013). The supplemental letter contained clarifying information that 
was within the scope of the original Federal Register notice and did 
not change the initial no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 12, 2001.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of application for amendments: November 28, 2000, as 
supplemented January 17, 2001, and February 15, 2001.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) to permit, as an alternative to the 
current dedicated Shift Technical Advisor (STA), a single, qualified 
individual to simultaneously serve as an STA and a Senior Reactor 
Operator, and either option would be permitted on a shift-by-shift 
basis.
    Date of Issuance: March 14, 2001.
    Effective Date: March 14, 2001.
    Amendment Nos.: 113 and 173.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the TS.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81922). The letters dated January 17, 2001, and February 15, 2001, 
contained clarifying information that did not affect the original 
proposed no significant hazards determination, or expand the scope of 
the request as noticed.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 14, 2001.
    No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: June 15, 1999, as supplemented by letter 
dated November 14, 2000.
    Brief description of amendment: The amendment authorized revision 
of the Updated Safety Analysis Report (USAR) to allow the use of the 
service water system to directly supply cooling water to the reactor 
equipment cooling system during a loss-of-coolant accident event.
    Date of issuance: March 13, 2001.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 185.
    Facility Operating License No. DPR-46: Amendment authorized 
revision to the USAR.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38030). The November 14, 2000, supplemental letter provided clarifying 
information that was within the scope of the original Federal Register 
notice and did not change the staff's initial no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 13, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 20, 2000.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) to (1) include the automatic reactor water cleanup 
(RWCU) system isolation feature, (2) restore the dose equivalent 
iodine-131 limit to 2 microcuries per gram, (3) change the RWCU reactor 
water level automatic isolation signal from Low to Low-Low reactor 
water level and add TSs for the high pressure coolant injection (HPCI) 
and reactor core isolation cooling low steam line pressure isolation 
instrumentation, (4) delete the HPCI 150,000 lb/hr low range high flow 
isolation instrumentation and adds a time delay to the 300,000 lb/hr 
upper range high flow isolation instrumentation, and (5) change the 
suppression chamber water allowable water level from volume units to 
level units.
    Date of issuance: March 7, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 117.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 23, 2000 (65 FR 
51361).
    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.

[[Page 17976]]

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

    Date of application for amendment: January 10, 2001.
    Brief description of amendment: The amendment removes the standby 
liquid control (SLC) pump flow surveillance requirement to recycle 
demineralized water to the test tank and changes the testing frequency 
of the SLC pump capacity test from monthly to quarterly.
    Date of issuance: March 8, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 118.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 7, 2001 (66 FR 
9386).
    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.

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

    Date of amendment request: September 5, 2000, as supplemented by 
letters dated September 28, 2000, December 1, 2000, and December 11, 
2000.
    Brief description of amendment: The amendment revised Sections 1.1, 
1.3, 2.10, 3.10, and 5.9 and associated Bases of the Fort Calhoun 
Station, Unit No. 1 (FCS) technical specifications. The amendment 
allows use of NRC-approved Siemens Power Corporation (SPC) 
methodologies for determining reactor core operating limits in 
conjunction with use of SPC fabricated nuclear fuel. Additionally, the 
revised SPC fuel assembly growth model for FCS Cycle 20 core reload was 
reviewed and approved.
    Date of issuance: March 14, 2001
    Effective date: March 14, 2001, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 196.
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81925).
    The September 28, December 1 and 11, 2000, supplemental letters 
provided additional clarifying information, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 14, 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: July 31, 2000.
    Brief description of amendments: The amendments revised the main 
steam isolation valve leakage rate surveillance requirements.
    Date of issuance: March 9, 2001.
    Effective date: As of date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 190 and 165.
    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 Safety Evaluation dated March 9, 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: November 16, 2000.
    Brief description of amendments: The amendments eliminated response 
time testing requirements for certain reactor protection system and 
isolation actuation system instrumentation.
    Date of issuance: March 12, 2001.
    Effective date: As of date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 191 and 166.
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 10, 2001 (66 FR 
2022).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 12, 2001.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant , Units 1, 2, and 3, Limestone County, 
Alabama

    Date of application for amendments: October 6, 2000.
    Description of amendment request: The amendments revise the 
Technical Specifications (TS) to specify required actions and 
completion times applicable to conditions when two low-pressure coolant 
injection pumps, each in a different subsystem, are inoperable.
    Date of issuance: March 12, 2001.
    Effective date: March 12, 2001.
    Amendment Nos: 240, 269, 229.
    Facility Operating License Nos. DPR-33, DPR-52, and DPR-68. 
Amendments revise the TS..
    Date of initial notice in the Federal Register: November 15, 2000 
(65 FR 69066) and re-noticed February 7, 2001 (66 FR 9387).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 12, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 21, 2000 as 
supplemented by a February 9, 2001 reply to a request for additional 
information.
    Brief description of amendment: It revises the minimum critical 
power ratio safety limits specified in the facility Technical 
Specifications (TS) for two-loop and single-loop operation.
    Date of issuance: March 13, 2001.
    Effective date: March 13, 2001.
    Amendment No.: 270
    Facility Operating License No. DPR-52: Amendment revises the TS.
    Date of initial notice in Federal Register: December 13, 2000 (65 
FR 77927). The letter dated February 9, 2001, contained clarifying 
information that did not affect the original proposed no significant 
hazards determination, or expand the scope of the request as noticed.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 13, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 22, 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

[[Page 17977]]

it applies to changes to the TS Bases without prior NRC approval.
    Date of issuance: March 15, 2001
    Effective date: March 15, 2001, and shall be implemented within 60 
days from the date of issuance.
    Amendment No.: 142.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81931). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 15, 2001.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia

    Date of application for amendment: June 22, 2000, as supplemented 
November 15, 2000.
    Brief description of amendment: These amendments revise Technical 
Specification (TS) 3.1.2.7, TS 3.1.2.8, TS 3.5.1, TS 3.5.5, TS 3.6.2.2, 
and TS 3.9.1 to increase the boron concentration limits in the 
refueling water storage tank, casing cooling tank, safety injection 
accumulators, and the reactor coolant system during refueling.
    Date of issuance: March 20, 2001.
    Effective date: As of the date of issuance and shall be implemented 
at the end of the Fall 2001 refueling outage for Unit 1, and at the end 
of the Fall 2002 refueling outage for Unit 2.
    Amendment Nos.: 225 and 206
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments change 
the Technical Specifications.
    Date of initial notice in Federal Register: July 26, 2000 (65 FR 
46018). The November 15, 2000, supplement 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 20, 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: December 19, 2000.
    Brief Description of amendments: These amendments revise Table 3.7-
4, item 7, and Technical Specification 3.6.B. The changes revise the 
range of allowable values for the 4160-volt bus loss-of-voltage and 
degraded voltage relay settings.
    Date of issuance: March 12, 2001.
    Effective date: March 12, 2001.
    Amendment Nos.: 224 and 224.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: January 10, 2001 (66 FR 
2025).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 12, 2001.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland this 27th 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. 01-8101 Filed 4-3-01; 8:45 am]
BILLING CODE 7590-01-P