[Federal Register Volume 67, Number 73 (Tuesday, April 16, 2002)]
[Notices]
[Pages 18641-18656]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-8866]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice Applications and Amendments to Facility Operating 
Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 22, 2002 through April 4, 2002. The 
last biweekly notice was published on April 2, 2002 (67 FR 15619).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC's Public Document 
Room (PDR), located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland. The filing of requests for a 
hearing and petitions for leave to intervene is discussed below.
    By May 16, 2002, 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 NRC's PDR, located at One White Flint North, 11555 
Rockville Pike (first floor), Rockville, Maryland. Publicly available 
records

[[Page 18642]]

will be accessible from the Agencywide Documents Access and Management 
Systems (ADAMS) Public Electronic Reading Room on the internet at the 
NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a 
request for a hearing or petition for leave to intervene is filed by 
the above date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland, 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 PDR, located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. Publicly available records 
will be accessible from the Agencywide Documents Access and Management 
Systems (ADAMS) Public Electronic Reading Room on the internet at the 
NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not 
have access to ADAMS or if there are problems in accessing the 
documents located in ADAMS, contact the NRC PDR Reference staff at 1-
800-397-4209, 304-415-4737 or by e-mail to [email protected].

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

    Date of amendments request: January 31, 2002.
    Description of amendments request: The proposed amendments would 
change the method of verifying the boron concentration of each safety 
injection tank. Rather than taking a sample from each tank every 31 
days, the proposed change would require leakage into the tanks to be 
monitored every 12 hours and a sample be taken every 6 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. Would not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Boron concentration is controlled in the safety injection tanks 
(SITs) to prevent either excessive boron concentrations or insufficient 
boron concentrations. Post-loss-of-coolant accident (LOCA) emergency 
procedures directing the operator to establish simultaneous hot and 
cold leg injection are based on the worst case minimum boron 
precipitation time. Maintaining the maximum SIT boron concentration 
within the upper limit ensures that the SITs do not invalidate this 
calculation. The minimum boron requirements of 2300 ppm [parts per 
million] are based on beginning-of-life reactivity values and are 
selected to ensure that the reactor will remain subcritical during the 
reflood stage of a large break LOCA. During a large break LOCA, all 
control element assemblies are assumed not to insert into the core, and 
the initial reactor shutdown is accomplished by void formation during 
blowdown.

[[Page 18643]]

Sufficient boron concentration must be maintained in the SITs to 
prevent a return to criticality during reflood. Level and pressure 
instrumentation is provided to monitor the availability of the tanks 
during plant operation.
    The Technical Specification Surveillance Requirement (SR 3.5.1.4) 
verifies that the boron concentration remains within the required range 
by sampling. Currently, the boron concentration in each SIT is required 
to be verified by taking a sample of the water in the SIT every 31 
days. A containment entry is required to take a sample from each of the 
four SITs. In addition, the boron concentration of the water added to 
the SITs is also sampled at the discharge of the high pressure safety 
injection pump to ensure that the water being added to the SITs is 
within the required boron concentration limits prior to being added. 
All intentional sources of level increase have their boron 
concentrations administratively maintained to ensure that the SIT boron 
concentrations are within Technical Specification limits. However, the 
Reactor Coolant System boron concentration is lower during power 
operation than the boron concentration in the SITs. Two check valves in 
series prevent leakage from the Reactor Coolant System into the SITs.
    This proposed amendment would require inleakage monitoring to be 
done every twelve hours in addition to taking samples from each SIT 
every six months. Samples would continue to be taken to verify the 
inleakage observations remain conservative. In addition, the 
requirement to sample the discharge of the operating high pressure 
safety injection pump prior to filling the SIT would remain.
    As noted above, the SITs are used only to respond to an accident 
and are not an accident initiator. Therefore, the probability of an 
accident has not increased.
    The engineering analysis and risk insights combine to demonstrate 
that the method of SIT boron concentration verification can be changed 
from sampling very 31 days to monitoring inleakage every twelve hours 
and sampling every six months. The inleakage monitoring is based on a 
calculation method that has sufficient conservatism to predict the 
boron concentration of the SITs as shown by sample. Therefore, the SITs 
would remain capable of responding to an accident as described above 
and the consequences of an accident previously evaluated are not 
increased.
    Therefore the probability or consequences of an accident previously 
evaluated are not increased.
    2. Would not create the possibility of a new or different [kind] of 
accident from any accident previously evaluated.
    The proposed change does not alter the function of any equipment, 
nor has it to operate differently than it was designed to operate. All 
equipment required to mitigate the consequences of an accident would 
continue to operate as before. The proposed change alters the method of 
verification of the SIT boron concentration, but not the boron 
concentration requirements themselves.
    Therefore, this change does not create the possibility of a new or 
different [kind] of accident from any accident previously evaluated.
    3. Would not involve a significant reduction in a margin of safety.
    The margin of safety defined by 10 CFR [Code of Federal 
Regulations] Part 100 has not been significantly reduced. The inleakage 
monitoring done to verify the concentration of boron in the SITs, is 
sufficiently conservative to ensure that the boron concentration would 
be underpredicted, leading to attempts to increase the boron 
concentration or a need to sample the affected SIT. Sampling of the 
SITs every six months will continue to be done to ensure that the 
inleakage monitoring remains conservative and representative. Water 
added to the SITs will also continue to be sampled to ensure that it 
meets the minimum boron concentrations. If the boron concentration is 
maintained in the SITs, the system operates as assumed in the Updated 
Final Safety Analysis Report Chapter 14 analyses and the analyses 
continue to meet the dose consequences acceptance criteria given in the 
Updated Final Safety Analysis Report.
    Therefore, this proposed change does not involve a significant 
reduction in [a] margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Joel Munday, Acting.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: February 21, 2002.
    Description of amendment request: The proposed amendment involves 
changes to the Fermi 2 Updated Final Safety Analysis Report (UFSAR) and 
Technical Requirements Manual which is incorporated by reference in the 
UFSAR to eliminate the chlorine detection function from the control 
room heating, ventilation, and air conditioning system. Changes to the 
UFSAR are subject to the requirements of 10 CFR 50.59; however, these 
changes are being submitted for Nuclear Regulatory Commission (NRC) 
review and approval since they involve the elimination of an automatic 
action in accordance with the Nuclear Energy Institute guidance 
document 96-07, Revision 1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The chlorine detection system was originally added to the plant 
design when it was assumed that a chlorine rail car would be located on 
site for use in water treatment purposes; however, one-ton chlorine 
cylinders were used instead. In 1992, the use of chlorine for on site 
water treatment was discontinued. There is no chlorine stored on site 
and no significant amounts are stored at any other facility within the 
5-mile radius of the plant. The only credible accident involving a 
chlorine release that could be carried into the control room is from a 
chlorine rail car accident on the three railroad tracks 3.4 to 3.8-
miles away from the site. The probability of a rail car accident and 
spill of chlorine is not affected by the removal of the chlorine 
detectors located in the normal air intake for the CCHVAC [control room 
heating, ventilation and air conditioning] system; therefore, only the 
consequences of the event must be addressed as a result of the proposed 
change.
    The chlorine detectors in the control room ventilation air intake 
are intended to provide protection to the control room occupants in the 
event of an accidental offsite chlorine release. Detroit Edison has 
performed a probabilistic risk assessment to determine the probability 
of reaching toxic chlorine concentration levels of 10 parts per million 
in the control room as a result of a chlorine railcar accident and 
spill within 5 miles of the plant. The probability analysis took no 
credit for any automatic or manual action to

[[Page 18644]]

isolate the control room. The results of the analysis show that the 
total probability of 8.4E-07 per year is below the 1.0E-06 threshold 
specified in Regulatory Guide (RG) 1.78, Revision 1. Therefore, since 
the probability analysis results meet the RG criteria, the elimination 
of the chlorine detection function will not significantly increase the 
consequences of an offsite chlorine release.
    2. The change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The probabilistic risk assessment evaluation demonstrates that the 
likelihood of creating hazardous conditions in the control room as a 
result of a chlorine accident is very small. RG 1.78, Revision 1, 
states that events of such low frequencies do not need to be considered 
in the plant design because the resultant low levels of radiological 
risk are considered acceptable. The probabilistic assessment assumed no 
automatic or manual action to isolate the control room or to filter 
outside air before it is discharged in the control room. The evaluation 
did not rely on any structure, system or component to perform a 
specific function; therefore, the elimination of the chlorine detection 
system does not create the potential for a new or different kind of 
accident from any accident previously evaluated.
    3. The change does not involve a significant reduction in the 
margin of safety.
    The elimination of the chlorine detection system will not affect 
the protection of the control room operators from the hazard of an 
offsite chlorine release. No significant amounts of chlorine are stored 
within 5 miles of the plant and the only chlorine accident risk is from 
a railroad car accident over 3 miles away. The probabilistic evaluation 
demonstrates the low risk associated with a chlorine accident that 
would incapacitate the operators such that their functions in 
mitigating a radiological event are impacted. Since the Regulatory 
Positions in RG 1.78, Revision 1 are satisfied, deletion of the 
chlorine detection system will not result in 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: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: William D. Reckley, Acting.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: February 5, 2002
    Description of amendment request: The proposed amendment would 
revise the surveillance requirements associated with the Containment 
Isolation Valves (CIVs), Reactor Building Closed Cooling Water (RBCCW) 
System, and Service Water (SW) System. The proposed changes would 
remove redundant testing requirements that are already addressed by the 
Inservice Testing (IST) Program, which is required pursuant to 
Technical Specification 4.0.5, and would use Technical Specification 
4.0.5 to control the specific acceptance criteria and frequency of test 
performance. Additional proposed changes would remove the post 
maintenance testing requirements associated with the CIVs, revise the 
wording of the RBCCW and SW Systems Limiting Conditions for Operation, 
and increase the allowed outage times for the RBCCW and SW Systems.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed Technical Specification changes associated with the 
limiting condition for operation requirements, surveillance 
requirements, and allowed outage times will not cause an accident to 
occur and will not result in any change in the operation of the 
associated accident mitigation equipment. The ability of the equipment 
associated with the proposed changes to mitigate the design basis 
accidents will not be affected. The proposed changes to the limiting 
condition for operation requirements will not affect the equipment 
operability requirements. The proposed surveillance requirements are 
adequate to ensure proper operation of the associated accident 
mitigation equipment. Proper operation of the containment isolation 
valves will still be verified, as appropriate, following maintenance 
activities. The proposed allowed outage times are reasonable and 
consistent with standard industry guidelines to ensure the accident 
mitigation equipment will be restored in a timely manner. The design 
basis accidents will remain the same postulated events described in the 
Millstone Unit No. 2 Final Safety Analysis Report, and the consequences 
of those events will not be affected. Therefore, the proposed changes 
will not increase the probability or consequences of an accident 
previously evaluated.
    The additional proposed changes to the Technical Specifications 
(e.g., combining requirements, deleting an expired footnote, and 
renumbering a requirement) will not result in any technical changes to 
the current requirements. Therefore, these additional proposed changes 
will not increase the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes to the Technical Specifications do not impact 
any system or component that could cause an accident. The proposed 
changes will not alter the plant configuration (no new or different 
type of equipment will be installed) or require any unusual operator 
actions. The proposed changes will not alter the way any structure, 
system, or component functions, and will not alter the manner in which 
the plant is operated. There will be no effect on plant operation or 
accident mitigation equipment. The response of the plant and the 
operators following an accident will not be different. In addition, the 
proposed changes do not introduce any new failure modes. Therefore, the 
proposed changes will not create the possibility of a new or different 
kind of accident from any accident previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed Technical Specification changes associated with the 
limiting condition for operation requirements, surveillance 
requirements, and allowed outage times will not cause an accident to 
occur and will not result in any change in the operation of the 
associated accident mitigation equipment. The equipment associated with 
the proposed Technical Specification changes will continue to be able 
to mitigate the design basis accidents as assumed in the safety 
analysis. The proposed surveillance requirements are adequate to ensure 
proper operation of the affected accident mitigation equipment. The 
proposed allowed outage times are reasonable and consistent with 
standard industry guidelines to ensure the accident

[[Page 18645]]

mitigation equipment will be restored in a timely manner. In addition, 
the proposed changes will not affect equipment design or operation, and 
there are no changes being made to the Technical Specification required 
safety limits or safety system settings. The proposed Technical 
Specification changes, in conjunction with existing administrative 
controls (e.g., IST Program), will provide adequate control measures to 
ensure the accident mitigation functions are maintained. Therefore, the 
proposed changes will not result in a reduction in a margin of safety.
    The additional proposed administrative changes to the Technical 
Specifications (e.g., combining requirements, deleting an expired 
footnote, and renumbering a requirement) will not result in any 
technical changes to the current requirements. Therefore, these 
additional changes will not result in a reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: James W. Clifford.

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423, 
Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London County, 
Connecticut

    Date of amendment request: February 14, 2002.
    Description of amendment request: The proposed Technical 
Specification (TS) changes will relocate selected Millstone Units 2 and 
3 TSs related to the Reactor Coolant System (RCS) and Plant Systems to 
the Technical Requirements Manual (TRM). The proposed TSs for Unit 2 
include 3/4.4.9.1, ``Pressure/Temperature Limits,'' 3/4.7.2, ``Steam 
Generator Pressure/Temperature Limitation,'' 3/4.7.5, ``Flood Level,'' 
3/4.7.7, ``Sealed Source Contamination,'' 3/4.7.8, ``Snubbers,'' and 
related Tables, Figures, and Bases sections. The proposed TSs for Unit 
3 include 3/4.4.9.1, ``Pressure/Temperature Limits,'' 3/4.7.2, ``Steam 
Generator Pressure/Temperature Limitation,'' 3/4.7.6, ``Flood 
Protection,'' 3/4.7.10, ``Snubbers,'' 3/4.7.11, ``Sealed Source 
Contamination,'' 3/4.7.14, ``Area Temperature Monitoring,'' and 
corresponding Tables, Figures, and Bases sections.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed technical specification changes will relocate to the 
TRM the following items: surveillance requirements for the withdrawal 
of reactor vessel material irradiation specimens of Millstone Unit Nos. 
2 and 3 which are part of the Pressure/Temperature Limits technical 
specifications, Millstone Unit Nos. 2 and 3 technical specifications 
covering Steam Generator Pressure/Temperature Limitation, Flood Level, 
Sealed Source Contamination, and Snubbers. Also the Millstone Unit No. 
3 technical specification covering Area Temperature Monitoring will be 
relocated to the TRM. Since the relocated requirements remain the same, 
the proposed changes will have no effect on plant operation, or the 
availability or operation of any accident mitigation equipment. 
Therefore, the relocation of the requirements associated with these 
technical specifications will not impact an accident initiator and 
cannot cause an accident. These changes will not increase the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed technical specification changes will relocate the 
requirements of selected Millstone Unit Nos. 2 and 3 technical 
specifications as described above to the TRM. The proposed changes do 
not alter the plant configuration (no new or different type of 
equipment will be installed) or require any new or unusual operator 
actions. Since the requirements remain the same, the proposed changes 
do not alter the way any system, structure, or component functions and 
do not alter the manner in which the plant is operated. The proposed 
changes do not introduce any new failure modes. Therefore, the proposed 
changes will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed technical specification changes will relocate to the 
TRM the following items: surveillance requirements for the withdrawal 
of reactor vessel material irradiation specimens of Millstone Unit Nos. 
2 and 3 which are part of the Pressure/Temperature Limits technical 
specifications, Millstone Unit Nos. 2 and 3 technical specifications 
covering Steam Generator Pressure/Temperature Limitation, Flood Level, 
Sealed Source Contamination, and Snubbers. Also the Millstone Unit No. 
3 technical specification covering Area Temperature Monitoring will be 
relocated to the TRM. Since the proposed changes are solely to relocate 
the existing requirements, the proposed changes will have no effect on 
plant operation, or the availability or operation of any accident 
mitigation equipment. The plant response to the Design Basis Accidents 
will not change. Therefore, there will be no 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: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
    NRC Section Chief: James W. Clifford.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station Unit No. 1, Oswego County, New York

    Date of amendment request: March 15, 2002.
    Description of amendment request: The proposed amendment would 
revise the Nine Mile Point Unit 1 Technical Specifications (TSs), Table 
4.6.4, ``Shock Suppressors (Snubbers),'' consistent with the model 
snubber visual inspection and acceptance requirements conveyed in 
Generic Letter 90-09, ``Alternative Requirements for Snubber Visual 
Inspection and Corrective Actions.''
    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 operation of Nine Mile Point Unit 1 in accordance with the 
proposed amendment will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Snubbers are utilized at Nine Mile Point Unit 1 (NMP1) to ensure 
the

[[Page 18646]]

structural integrity of the reactor coolant system and other safety-
related (as well as certain non-safety related) systems during and 
following a seismic event or other event initiating dynamic loads. The 
proposed change to the snubber visual inspection schedule is based on 
that delineated in NRC [Nuclear Regulatory Commission] Generic Letter 
(GL) 90-09, ``Alternative Requirements for Snubber Visual Inspection 
and Corrective Actions.'' This change does not modify any accident 
initiators or change any equipment or procedures used to limit the 
consequences of any accidents previously evaluated.
    Accordingly, the proposed amendment will not significantly increase 
the probability or consequences of an accident previously evaluated.
    2. The operation of Nine Mile Point Unit 1 in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    No physical modifications are being made to any snubbers or to any 
systems supported by snubbers by this proposed amendment. No method of 
plant or system operation is varied by use of the alternate snubber 
visual inspection schedule delineated in GL 90-09. Only the method 
utilized to determine future surveillance intervals for snubber visual 
inspections based on the previous inspection results is changed by the 
proposed amendment. This method was developed and published by the NRC 
in GL 90-09 for generic application at nuclear power plants.
    Accordingly, the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    In GL 90-09, the NRC staff determined that use of the alternate 
snubber visual inspection schedule by nuclear power plants will 
maintain the same level of confidence as the previous schedule required 
by the plants' Technical Specifications. GL 90-09 also recognized that 
snubber visual inspection is a complementary process to snubber 
functional testing and provides additional confidence in snubber 
operability. Snubber functional testing is not being modified by this 
proposed amendment.
    Therefore, the proposed change will not adversely affect any 
structure, system, component, or function that is safety-related or 
important to safety. Accordingly, the proposed amendment will not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment 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: Joel Munday, Acting.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: March 19, 2002.
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Nuclear Power Plant accident source term used for 
design basis radiological analyses.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Differences between the original source term and the proposed AST 
[accident source term] cannot affect the previously analyzed core 
damage frequency (CDF) and large early release frequency (LERF). Since 
there are no modifications proposed with this request for AST, Limiting 
Safety System Settings and Safety Limits specified in the Technical 
Specifications remain unchanged. Re-analysis of design basis accidents 
as described herein demonstrates that regulatory dose acceptance 
criteria continue to be satisfied. Thus, nothing in this proposal will 
cause an increase in the probability or consequence of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    There are no physical changes to the plant associated with this 
request, and the plant conditions for which [Nuclear Management 
Company] (NMC) evaluated design-basis accidents remain valid. 
Consequently, this proposal introduces no new failure modes. Thus, this 
proposal does not create the possibility of a new or different kind of 
accident.
    3. Involve a significant reduction in the margin of safety.
    The revised design-basis accident offsite and control-room dose-
calculations proposed herein remain within regulatory acceptance 
criteria set forth in 10 CFR 100 and 10 CFR 50 Appendix A, General 
Design Criterion 19. They also use the TEDE [total effective dose 
equivalent] dose acceptance criteria as directed by the Commission. An 
acceptable margin of safety is inherent in the limits described 
thereby. Thus, changes proposed by this request 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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: William D. Reckley, Acting.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of amendment request: February 28, 2002.
    Description of amendment request: The proposed amendment would 
revise technical specifications (TS) 1.1, ``Definitions,'' ``CREFS 
Actuation Instrumentation,'' TS 3.4.16, ``RCS Specific Activity,'' TS 
3.3.5, ``CREFS Actuation Instrumentation,'' TS 3.4.16, ``RCS Specific 
Activity,'' TS 3.7.9, ``CREFS,'' and TS 3.7.13, ``Secondary Specific 
Activity,'' and delete TS 3.9.3, ``Containment Penetrations.''
    The accident source term used in the selection of the design-basis 
offsite and control room dose analysis would be replaced by the 
implementation of an alternative source term.
    The specific TS changes would be as follows: (1) TS 1.1, 
``Definitions:'' Revise the definition of La (containment 
leakage) by changing 0.4 percent to 0.2 percent. (2) TS 3.3.5, ``CREFS 
Actuation Instrumentation:'' Revise table 3.3.5-1 to indicate that 
either RE-101 or RE-235 must be operable to ensure that the control 
room radiation instrumentation necessary to initiate the CREFS 
emergency make-up mode is operable. Add the Control Room Area Monitor 
and Control Room Air Intake trip setpoints to Note ``d'' of table 
3.3.5-1.

[[Page 18647]]

(3) TS 3.4.16, ``RCS Specific Activity:'' Revise LCO Action Condition A 
to indicate 1.0 Ci/gm as the maximum reactor coolant dose 
equivalent iodine 131 (DE I-131) value. Revise Figure 3.4.16-1 to 
indicate 60 Ci/gm DE I-131 as the maximum RCS limit for 
operations at or above 80 percent of rated thermal power. Revise SR 
3.4.16.2 to verify 1.0 Ci/gm as the maximum reactor coolant DE 
I-131 value. (4) TS 3.7.9, ``CREFS:'' Delete SR 3.7.9.5. (5) TS 3.7.13, 
``Secondary Specific Activity:'' Revise LCO 3.5.13 and SR 3.7.13 to 
indicate that the secondary specific activity shall be less than or 
equal to 0.1 Ci/gm. (6) TS 3.9.3, ``Containment 
Penetrations:'' Delete Section 3.9.3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. Involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    The Alternative Source Term (AST) and those plant systems affected 
by implementing the proposed changes to the TS are not accident 
initiators and cannot increase the probability of an accident. The AST 
does not adversely affect the design or operation of the facility in a 
manner that would create an increase [in] the probability of an 
accident. Rather, the AST is used to evaluate the dose consequences of 
a postulated accident. The revised dose calculations, except those for 
LOCA, use the values in the proposed TS. The limiting design bases 
accidents at PBNP have been evaluated for implementation of the AST.
    These analyses have demonstrated that, with the proposed changes, 
the dose consequences meet the regulatory acceptance criteria of 10 CFR 
50.67 and RG 1.183. A comparison of the current offsite dose 
calculations to the revised offsite dose calculations indicate that the 
proposed changes will not result in a significant increase in the 
predicted dose consequences for any of the analyzed accidents. 
Therefore, the proposed changes do not involve a significant increase 
in the probability or consequences of any of the selected previously 
analyzed accidents.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed amendment will not create the possibility for a new or 
different type of accident from any accident previously evaluated. 
Changes to the allowable activity in the primary and secondary systems 
do not result in changes to the design or operation of these systems. 
The evaluation of the effects of the proposed changes indicates that 
all design standard and applicable safety criteria limits are met.
    The systems affected by the changes are used to mitigate the 
consequences of an accident that has already occurred. The proposed TS 
changes and modifications do not significantly affect the mitigative 
function of these systems. Equipment important to safety will continue 
to operate as designed. Component integrity is not challenged. The 
changes do not result in any event previously deemed incredible being 
made credible. The changes do not result in more adverse conditions or 
result in any increase in the challenges to safety systems.
    Therefore, the proposed changes do not create the possibility of a 
new or different type of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The implementation of the proposed changes does not significantly 
reduce the margin of safety. These changes have been evaluated in the 
revisions to the analysis of the consequences of the design basis 
accidents for PBNP. The radiological analysis results in concert with 
the proposed TS changes, meet the regulatory acceptance criteria of 10 
CFR 50.67 and RG 1.183. These acceptance criteria have been developed 
for the purpose of use in design basis accident analyses such that 
meeting these limits demonstrates adequate protection of public health 
and safety. The proposed changes will not degrade the plant protective 
boundaries, will not cause a release of fission products to the public 
and will not degrade the performance of any SSCs important to safety.
    Therefore, the proposed changes to the TS would 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 H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: William D. Reckley, Acting.

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

    Date of amendment request: February 2, 2001, supplemented August 
31, 2001.
    Description of amendment request: The proposed amendments would 
revise the technical specifications (TSs) to clarify the plant 
conditions under which various specifications are applicable. The 
licensee stated in its amendment request that a literal reading of the 
current technical specifications wording may result in situations where 
a routine plant shutdown would seem to be prohibited by TSs and, 
thereby, require entry into TS 3.0.C. This amendment request also makes 
several administrative changes to the TSs, including revising 
references to the Chief Nuclear Corporate Officer, capitalizing defined 
terms, and updating references to previously relocated TS paragraphs 
and correcting the List of Figures. The licensee's supplement to the 
amendment request, dated August 31, 2001, proposed a correction of a 
typographical error in TS Table 3.5-2B, Action 33.
    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 operation of the facility with the proposed amendment 
involve a significant increase in the probability or consequences of an 
accident previously evaluated?
    The proposed changes are administrative in nature and clarify 
existing specifications without reducing or altering the requirements 
imposed by existing specifications. The proposed changes do not 
significantly affect any system that is a contributor to initiating 
events for previously evaluated accidents. Neither do the changes 
significantly affect any system that is used to mitigate any previously 
evaluated accidents. Therefore, the proposed changes do not involve any 
significant increase in the probability or consequence of an accident 
previously evaluated.
    2. Does operation of the facility with the proposed amendment 
create the possibility of a new or different kind of accident from any 
accident previously evaluated?
    The proposed changes are administrative in nature and clarify 
existing specifications without reducing or altering the requirements 
imposed by existing specifications. The proposed

[[Page 18648]]

changes do not alter the design, function, or operation of any plant 
component and do not install any new or different equipment, therefore 
a possibility of a new or different kind of accident from those 
previously analyzed has not been created.
    3. Does operation of the facility with the proposed amendment 
involve a significant reduction in a margin of safety?
    The proposed changes are administrative in nature and clarify 
existing specifications without reducing or altering the requirements 
imposed by existing specifications. Thus, the proposed change[s] do not 
involve a significant reduction in the margin of safety associated with 
the safety limits inherent in either the princip[al] barriers to a 
radiation release (fuel cladding, RCS [reactor coolant system] 
boundary, and reactor containment), or the maintenance of critical 
safety functions (subcriticality, core cooling, ultimate heat sink, RCS 
inventory, RCS boundary integrity, and containment integrity).
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: William D. Reckley, Acting.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of amendment requests: March 11, 2002.
    Description of amendment requests: The proposed amendment would 
revise the Technical Specifications (TSs) for San Onofre Nuclear 
Generating Station, Units 2 and 3. Specifically, TS Section 1.1, 
Definitions, would be revised to change the definition of response time 
testing as it is applied to the Engineered Safety Features, and the 
Reactor Protective System. The proposed change is based on approved 
Technical Specification Task Force (TSTF) Traveler TSTF-368, Revision 
0, ``Incorporate Combustion Engineering Owners Group (CEOG) Topical 
Report to Eliminate Pressure Sensor Response Time Testing.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment to the Technical Specification (TS) 
Definitions for Engineered Safety Feature (ESF) Response Time and 
Reactor Protective System (RPS) Response Time allows substitution of an 
allocated sensor response time in lieu of measuring sensor response 
time. Response time testing is not an initiator of any accident 
previously evaluated. Further, overall system response time will 
continue to meet Technical Specification requirements. The allocated 
sensor response times allowed in lieu of measurement have been 
determined to adequately represent the response time of the components 
such that the safety systems utilizing those components will continue 
to perform their accident mitigation function as assumed in the safety 
analysis. Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment to TS Section 1.1, ``Definitions,'' allows 
the substitution of an allocated sensor response time in lieu of sensor 
response time testing for selected components. The proposed change does 
not involve a physical alteration of the plant (no new or different 
type of equipment will be installed) or a change in the methods 
governing normal plant operation. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment to TS Section 1.1, ``Definitions,'' allows 
the substitution of an allocated sensor response time in lieu of 
measured sensor response time for certain pressure sensors. The 
allocated pressure sensor response times allowed in lieu of measurement 
have been determined to adequately represent the response time of the 
components such that the safety systems utilizing those components will 
continue to perform their accident mitigation function as assumed in 
the safety analysis. Therefore, the proposed change does not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: March 4, 2002 (TS 00-04).
    Brief description of amendments: The proposed amendment would 
change the Sequoyah (SQN) Unit 1 and 2 Technical Specification (TS) to 
relocate the current requirements for ice condenser ice bed temperature 
and inlet door position monitoring systems to the SQN Technical 
Requirements Manual (TRM). These relocated specifications are 
consistent with the latest version of the improved Standard TS (NUREG-
1431). The affected functions have been evaluated in accordance with 
Title 10 of the Code of Federal Regulations, Section 50.36 (10 CFR 
50.36) for applicability to the criteria for requirements that must be 
retained in the TS. In each case, the four criteria of 10 CFR 50.36 did 
not apply to these functions. This revision will provide better 
consistency between the SQN TS and NUREG-1431.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), 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 revision relocates the ice bed temperature monitoring 
system and the inlet door position monitoring system to the TRM. 
Relocation to the TRM continues to provide an acceptable level of 
applicability to plant operation and requires revisions to be processed 
in accordance with the provisions in 10 CFR 50.59. Evaluations of 
revisions in accordance with 10 CFR 50.59 will continue to ensure that 
these specifications adequately control the

[[Page 18649]]

functions of ice bed temperature and inlet door positions to maintain 
safe operation of the plant. These systems are not postulated to be the 
initiator of a design basis accident. Since there are no changes to 
these functions and their operation will remain the same, the 
probability of an accident is not increased by relocating these 
requirements to the TRM. Additionally, the accident mitigation 
capability and offsite dose consequences associated with accidents will 
not change because these functions will not be altered by the proposed 
relocation. Therefore, the consequences of an accident are not 
increased by this relocation to the TRM and the control of revisions to 
these specifications in accordance with 10 FR 50.59.
    B. The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed revision will not alter the functions for the ice bed 
temperature or inlet door positions such that accident potential would 
be changed. The location of these specifications in the TRM and the 
performance of revisions in accordance with 10 CFR 50.59 will continue 
to maintain acceptable operability requirements. Therefore, the 
possibility of an accident of a new or different kind is not created by 
the proposed relocation and deletion.
    C. The proposed amendment does not involve a significant reduction 
in a margin of safety.
    The proposed specification relocation will not affect plant 
setpoints or functions that maintain the margin of safety. This is 
based on the relocation to the TRM. The TRM continues to maintain the 
same level of operability requirements and surveillance testing to 
adequately ensure functionality of the ice bed temperature monitoring 
system and the inlet door position monitoring system. The TRM is 
controlled in accordance with requirements of 10 CFR 50.59. Therefore, 
the proposed relocation and deletion is acceptable and will not reduce 
the 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 Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: March 4, 2002 (TS 01-03)
    Brief description of amendments: The proposed amendment would 
change the Sequoyah (SQN) Unit 1 and 2 Technical Specifications (TSs) 
to delete one definition and modify several subsections contained in TS 
Section 6.0, Administrative Controls. These proposed changes have been 
prepared based on existing NRC guidance. The changes are being proposed 
in the following areas:
     Definition 1.17--``Member(s) of the Public.'' (NUREG-1431, 
Revision 2)
     TS 6.2.2.g, Overtime. (TS Travelers Form (TSTF)-258, 
Revision 4)
     TS 6.3, Facility Staff Qualifications. (TSTF-258, Revision 
4)
     TS 6.8.4.a.ii, Primary Coolant Sources Outside 
Containment. (TSTF-299)
     TS 6.8.4.f, Radioactive Effluent Controls Program. (TSTF-
258, Revision 4 and TSTF-308, Revision 1)
     TS 6.8.4.i, Deletion of the ``Configuration Risk 
Management Program.'' (10 CFR 50.65)
     The second paragraph in TS 6.9.1.5 associated with 
specific activity limits. (NUREG-1431, Revision 2)
     TS 6.9.1.14, Monthly Reactor Operating Report contents 
revision. (TSTF-258, Revision 4)
     TS 6.12, High Radiation Areas revision. (TSTF-258, 
Revision 4)
     TS 6.15, Deletion of Major Changes To Radioactive Waste 
Treatment Systems (Liquid, Gaseous, and Solid). (NUREG-1431, Revision 
2)
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No. The proposed changes that involve the rewording or 
reformatting of the existing TSs do not involve technical changes. 
Therefore, this change is administrative and does not affect the 
initiators of analyzed events or assumed mitigation of accidents or 
transient events.
    Three of the changes remove programs from TSs based on present 
regulatory controls. Specifically 10 CFR 50.59, 10 CFR 50.65, 10 CFR 
50.71(e), 10 CFR 50.73, and Performance Indicator data. Based on the 
requirements residing in existing regulations it is acceptable to 
remove them from TS. Additionally, any changes to these programs will 
be evaluated based on regulatory requirements, no significant increase 
in the probability or consequences of an accident previously evaluated 
will be allowed.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The proposed changes do not involve a physical 
alteration of the plant (no new or different type of equipment will be 
installed) or changes in methods governing normal plant operation. 
Therefore, the proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The proposed changes will not reduce the margin of 
safety because they have no effect on any safety analysis assumptions. 
Additionally, the proposed programs to be removed from TSs are 
contained in existing plant programs required by existing regulations. 
Since any future changes to these programs will be evaluated, no 
significant reduction in a margin of safety will be allowed.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, TVA concludes that the proposed amendment(s) 
present no significant hazards consideration under the standards set 
forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.
    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.

[[Page 18650]]

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

    Date of amendment request: February 21, 2002.
    Description of amendment request: The proposed amendment would 
revise Required Actions for Limiting Conditions for Operation (LCOs) 
3.3.1, ``Reactor Trip System (RTS) Instrumentation;'' 3.4.5, ``RCS 
[Reactor Coolant System] Loops--MODE 3;'' 3.4.6, ``RCS Loops--MODE 4;'' 
3.4.7, ``RCS Loops--MODE 5, Loops Filled;'' 3.4.8, ``RCS Loops--MODE 5, 
Loops Not Filled;'' 3.8.2, ``AC Sources--Shutdown;'' 3.8.5, ``DC 
Sources--Shutdown;'' 3.8.8, ``Inverters--Shutdown;'' 3.8.10, 
``Distribution Systems--Shutdown;'' 3.9.3, ``Nuclear Instrumentation;'' 
3.9.5, ``Residual Heat Removal (RHR) and Coolant Circulation--High 
Water Level;'' and 3.9.6, ``Residual Heat Removal (RHR) and Coolant 
Circulation--Low Water Level'' in the Wolf Creek Generating Station 
Technical Specifications (TSs). The Required Actions proposed would 
suspend operations involving positive reactivity additions or RCS boron 
concentration reductions. In addition, the proposed amendment would 
revise Notes, for several of the above LCOs, that preclude reductions 
in RCS boron concentration. This amendment would revise these Required 
Actions and LCO Notes to allow small, controlled, safe insertions of 
positive reactivity, but limit the introduction of positive reactivity 
such that compliance with the required shutdown margin or refueling 
boron concentration limits will still be satisfied.
    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.
    Overall protection system performance will remain within the bounds 
of the previously performed accident analyses since there are no 
hardware changes. The RTS instrumentation and reactivity control 
systems will be unaffected. Protection systems will continue to 
function in a manner consistent with the plant design basis. All 
design, material, and construction standards that were applicable prior 
to the request are maintained.
    The probability and consequences of accidents previously evaluated 
in the USAR [Updated Safety Analysis Report] are not adversely affected 
because the changes to the Required Actions and LCO Notes assure the 
limits on SDM [shutdown margin] and refueling boron concentration 
continue to be met, consistent with the analysis assumptions and 
initial conditions included within the safety analysis and licensing 
basis. The activities covered by this amendment application are routine 
operating evolutions. The proposed changes do not reduce the capability 
of reborating the RCS.
    The equipment and processes used to implement RCS boration or 
dilution evolutions are unchanged and the equipment and processes are 
commonly used throughout the applicable MODES under consideration. 
There will be no degradation in the performance of, or an increase in 
the number of challenges imposed on, safety-related equipment assumed 
to function during an accident situation. There will be no change to 
normal plant operating parameters or accident mitigation performance.
    The proposed changes will not alter any assumptions or change any 
mitigation actions in the radiological consequence evaluations in the 
USAR.
    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.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. This amendment will not affect the normal method of plant 
operation or change any operating limits. The proposed changes merely 
permit the conduct of normal operating evolutions when additional 
controls over core reactivity are imposed by the Technical 
Specifications. The proposed changes do not introduce any new equipment 
into the plant or alter the manner in which existing equipment will be 
operated. The changes to operating procedures are minor, with 
clarifications provided that required limits must continue to be met. 
No performance requirements or response time limits will be affected. 
These changes are consistent with assumptions made in the safety 
analysis and licensing basis regarding limits on SDM and refueling 
boron concentration.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result of 
this amendment. There will be no adverse effect or challenges imposed 
on any safety-related system as a result of this amendment.
    This amendment does not alter the design or performance of the 7300 
Process Protection System, Nuclear Instrumentation System, or Solid 
State Protection System used in the plant protection systems.
    Therefore, the proposed changes do 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 changes do not alter the limits on SDM or refueling 
boron concentration. The nominal trip setpoints specified in the 
Technical Specifications Bases and the safety analysis limits assumed 
in the transient and accident analyses are unchanged. None of the 
acceptance criteria for any accident analysis is changed.
    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. There will be no impact on the overpower limit, 
departure from nucleate boiling ratio (DNBR) limits, heat flux hot 
channel factor (FQ), nuclear enthalpy rise hot channel 
factor (FutriH), loss of coolant accident peak cladding 
temperature (LOCA PCT), peak local power density, or any other margin 
of safety. The radiological dose consequence acceptance criteria listed 
in the Standard Review Plan will continue to be met.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application

[[Page 18651]]

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. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by e-mail to [email protected].

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

    Date of application for amendment: June 7, 2001.
    Brief description of amendment: The amendment revises the Oyster 
Creek Technical Specifications, Section 6.2.2.2.j, to allow either the 
Senior Manager-Operations or an Operations Manager to satisfy the 
Senior Reactor Operator-licensed requirement of this section.
    Date of Issuance: March 25, 2002.
    Effective date: March 25, 2002, and shall be implemented within 30 
days of issuance
    Amendment No.: 226.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 25, 2001 (66 FR 
38757). 
The Commission's related evaluation of this amendment is contained in a 
Safety Evaluation dated March 25, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 6, 2001.
    Brief description of amendment: The amendment allows the 24-month 
capacity test for the Diesel Generator Starting Batteries to be 
performed during plant shutdowns or during the 24-month on-line Diesel 
Generator inspection.
    Date of Issuance: March 27, 2002.
    Effective date: March 27, 2002 and shall be implemented within 30 
days of issuance
    Amendment No.: 227.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31702). 
The Commission's related evaluation of this amendment is contained in a 
Safety Evaluation dated March 27, 2002.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket No. 50-325, Brunswick Steam 
Electric Plant, Unit 1, Brunswick County, North Carolina

    Date of amendment request: September 18, 2001, as supplemented 
December 10, 2001, and March 5, 2002.
    Description of amendment request: The amendment revises the Safety 
Limit Minimum Critical Power Ratio (SLMCPR) values contained in TS 
2.1.1.2, and revises the SLMCPR values from 1.10 to 1.12 for two 
recirculation loop operation, and from 1.11 to 1.14 for single 
recirculation loop operation.
    Date of issuance: March 22, 2002.
    Effective date: March 22, 2002.
    Amendment No.: 220.
    Facility Operating License No. DPR-71: The amendment changes the 
Technical Specifications.
    Date of initial notice in Federal Register: October 17, 2001 (66 FR 

52797). The December 10, 2001, and March 5, 2002, supplements contained 
clarifying information only, and did not change the initial no 
significant hazards consideration determination or expand the scope of 
the initial Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 22, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application of amendments: March 29, 2001.
    Brief description of amendments: The amendments remove the NOTE 
that temporarily waived the upper limits of Technical Specifications 
3.8.1.9; thus, these amendments restore the original requirements of 
Surveillance Requirement 3.8.1.9. In addition, these amendments reduce 
the time delay specified in TS 3.8.1.17 from 12 seconds to 5 seconds. 
These amendments will be implemented when the digital governor 
modifications have been implemented on both Keowee Hydroelectric Units.
    Date of Issuance: March 20, 2002.
    Effective date: As of the date of completion of digital governor 
modifications on both Keowee Hydroelectric Units, and shall be 
implemented within 30 days of the date of completion of such 
modifications, but no later than April 30, 2005.
    Amendment Nos.: 322/322/323.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 2, 2001 (66 FR 
22029). 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated March 20, 2002.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of application for amendment: October 2, 2001, as supplemented 
by letter dated January 31, 2002.
    Brief description of amendment: The amendment changes the technical 
specifications definition of reactor trip system response time and 
engineered safety feature response time to allow use of either an 
allocated or a measured response time for select sensors in these two 
systems.
    Date of issuance: March 26, 2002.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.

[[Page 18652]]

    Amendment No.: 239.
    Facility Operating License No. NPF-6: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 

55016). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 26, 2002.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of application for amendment: October 2, 2001.
    Brief description of amendment: The amendment relocates the 
Technical Specification requirement that the reactor core be 
subcritical for a minimum of 175 hours prior to discharge of more than 
70 assemblies to the spent fuel pool, to the technical requirements 
manual.
    Date of issuance: April 1, 2002.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 240.
    Facility Operating License No. NPF-6: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 

55016). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 1, 2002.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: September 20, 2001.
    Brief description of amendment: The amendment allows the one-time 
extension of the intervals for selected Technical Specification (TS) 
surveillance requirements associated with the volume control tank, 
residual heat removal system, emergency diesel generators, and shock 
suppressors (snubbers). In addition, the amendment: (1) Corrects the 
channel functional test interval in Items 3 and 4 of TS Table 4.10-2 
and Items 4 and 5 of Table 4.10-4, (2) deletes the alternate inspection 
requirements for the steam generator snubbers, (3) removes the 
reference to a prior one-time extension of checks, calibrations, and 
tests for certain instrument channels in TS Table 4.1-1 that is no 
longer applicable.
    The amendment would enable the tests to be performed during the 
next refueling outage starting no later than November 19, 2002.
    Date of issuance: March 27, 2002.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 225.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 

55014). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 27, 2002.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: December 11, 2000, as 
supplemented on November 5 and December 7, 2001.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.1.F.2.a, ``Primary to Secondary Leakage,'' and TS 
4.13.A.3.f, ``Steam Generator Tube Inservice Surveillance,'' based on 
the prior replacement of the steam generators (SGs). Specifically, the 
changes (1) revise the primary to secondary leakage limits and (2) 
delete the requirements associated with tube sleeve repair, SG tube 
denting, and F* repair classification and criteria. The associated TS 
Bases have been modified accordingly. In addition, the amendment 
includes several related administrative changes.
    Date of issuance: April 2, 2002.
    Effective date: As of the date of issuance to be implemented within 
31 days.
    Amendment No.: 226.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7673). The November 5 and December 7, 2001, letters provided clarifying 
information that did not expand the application beyond the scope of the 
notice or change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 2, 2002.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: July 9, 2001, as supplemented by letters 
dated October 23, 2001, January 17, and February 1, 2002.
    Brief description of amendment: This Technical Specification (TS) 
change removes TS requirements that will no longer be applicable 
following replacement of the part-length control element assemblies 
with five-element full-length control element assemblies (CEAs) and 
removal of the four-element CEAs on the core periphery.
    Date of issuance: March 21, 2002.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 182.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41617). The supplement letters dated October 23, 2001, January 17, and 
February 1, 2002, contained clarifying information only, and did not 
change the initial no significant hazards consideration determination, 
or expand the scope of the initial application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 21, 2002.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: September 21, 2001, as supplemented by 
letters dated December 10, 2001, and January 16 and 21, 2002.
    Brief description of amendment: This amendment authorizes changes 
to the Waterford Steam Electric Station, Unit 3, Operating License and 
Technical Specifications associated with an increase in the licensed 
power level from 3,390 Megawatts thermal (MWt) to 3,441 MWt. These 
changes are made possible by increased feedwater flow measurement 
accuracy to be achieved by utilizing high accuracy ultrasonic flow 
measurement instrumentation.
    Date of issuance: March 29, 2002.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 183.
    Facility Operating License No. NPF-38: The amendment revised the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 

55017). The supplement letters dated December 10, 2001, and January 16 
and 21, 2002, contained clarifying information only, and did not change 
the initial no significant hazards consideration determination, or 
expand the scope of the initial application.

[[Page 18653]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 29, 2002.
    No significant hazards consideration comments received: No.

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 application for amendments: January 16, 2002 as 
supplemented February 7, 2002.
    Brief description of amendments: The amendments modified Technical 
Specification Surveillance Requirement 4.8.1.1.2.g.7 to permit 
performance of the required emergency diesel generator functional 
testing during power operation as an alternative to its performance 
during shutdown.
    Date of issuance: March 21, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos: 221 and 215.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 5, 2002 (67 FR 

5328). The licensee's February 7, 2002, supplemental information did 
not affect the original no significant hazards consideration 
determination, and did not expand the scope of the request as noticed 
on February 5, 2002.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 21, 2002.
    No significant hazards consideration comments received: No.

North Atlantic Energy Service Corporation, et al., Docket No. 50-443, 
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: August 6, 2001, as supplemented on 
November 2, 2001, and February 2, 2002.
    Description of amendment request: The amendment changes the 
Technical Specifications Sections 1.9, ``Core Alterations,'' 1.14, 
``Engineered Safety Features Response Time,'' and 1.29, ``Reactor Trip 
Response Time.''
    Date of issuance: April 3, 2002.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 81.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 59509). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 3, 2002.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of application for amendments: November 1, 2001.
    Brief description of amendments: These amendments revise the 
Technical Specifications to allow a one-time extension of the allowed 
outage time for the control room emergency filtration system (CREFS) 
from 7 days to 30 days. The licensee requested this one-time change in 
order to implement modifications to CREFS.
    Date of issuance: March 29, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 203 and 208.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 59510). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 29, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 14, 2001.
    Brief description of amendment: The amendment removes requirements 
for having the equipment hatch closed with four (4) bolts, and one door 
of the personnel access lock (PAL) closed during core alterations and 
refueling operations. The technical specifications (TS) for other 
containment penetrations were modified to be closed by an operable 
ventilation isolation actuation signal from one gaseous radiation 
monitor during core alterations and refueling operations. The amendment 
also modified the requirements for radiation monitors during core 
alterations and refueling operations. The TS Bases that were affected 
by the changes described above were modified. This amendment is based 
upon the alternate source term design basis site boundary and control 
room dose analyses previously reviewed and approved by the staff by 
Amendment No. 201 on December 14, 2001.
    Date of issuance: March 26, 2002.
    Effective date: March 26, 2002, and shall be implemented within 60 
days from the date of its issuance. The implementation of the amendment 
requires the commitments made by the licensee in Attachment 4 of its 
December 14, 2001, letter and as discussed in the staff's safety 
evaluation. These commitments are to be in place prior to any core 
alterations or refueling operations.
    Amendment No.: 204.
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2926). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 26, 2002.
    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: December 14, 2001, as supplemented by 
letter dated March 21, 2002.
    Brief description of amendment: The amendment revised Technical 
Specification 3.7(4) to allow the surveillance tests to be performed on 
a refueling frequency. In addition, the staff reviewed the 
documentation to correct the docket concerning inconsistencies in the 
1973 Fort Calhoun Station (FCS) Safety Evaluation Report (SER) 
associated with the 13.8 kV transmission line capability associated 
with TS 3.7(4) in accordance with OPPD's request.
    Date of issuance: March 26, 2002.
    Effective date: March 26, 2002 , to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 205.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2927). The March 21, 2002, supplemental letter provided additional 
information that clarified the application, 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 26, 2002.
    No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of application for amendments: March 21, 2001, as supplemented 
by letter dated January 11, 2002.

[[Page 18654]]

    Brief description of amendments: The amendments revise the 
operating license of each unit to delete those license conditions that 
have been completed and are no longer required and to make other 
corrections and editorial changes.
    Date of issuance: March 27, 2002.
    Effective date: March 27, 2002, to be implemented within 30 days of 
issuance.
    Amendment Nos.: Unit 2-185; Unit 3-176.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: April 18, 2001 (66 FR 
20009). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 27, 2002. The January 11, 
2002, supplemental letter provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendments request: August 25, 2000, as supplemented by 
letter dated November 2, 2001.
    Brief description of amendments: The amendments revise the Updated 
Final Safety Analysis Report described offsite dose analyses based on 
changes to the letdown flow rate and iodine spike postulated concurrent 
with the Main Steam Line Break or a Steam Generator Tube Rupture.
    Date of issuance: April 4, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 154/146.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: March 7, 2001 (66 FR 
13807). The supplement dated November 2, 2001, provided clarifying 
information that did not change the scope of the August 5, 2000, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 4, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 17, 2001, as 
supplemented December 14, 2001, and February 6, 2002.
    Brief description of amendments: The amendments revised the 
pressure-temperature limits for the reactor pressure vessel.
    Date of issuance: March 28, 2002.
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment Nos.: 275 and 233.
    Facility Operating License Nos. DPR-52 and DPR-68: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 19, 2001 (66 
FR 48291). The December 14, 2001, and February 6, 2002, letters 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination or expand 
the scope of the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 28, 2002.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear 
Plant, Unit 3, Limestone County, Alabama

    Date of application for amendment: November 1, 2001, as 
supplemented March 15, 2002.
    Brief description of amendment: The amendment revised the safety 
limit minimum critical power ratio values in Technical Specification 
2.1.1.2.
    Date of issuance: March 29, 2002.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 234.
    Facility Operating License No. DPR-68: Amendment revised the 
technical specifications.
    Date of initial notice in Federal Register: January 8, 2002 (67 FRN 
933). The March 15, 2002, letter provided clarifying information that 
did not change the scope of the original amendment request or the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 29, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 19, 2002, as supplemented by 
letters dated August 15, August 31, November 20, and December 17, 2001.
    Brief description of amendments: The application, as supplemented, 
requested that the antitrust conditions, contained in Appenix C of 
Facility Operating Licenses Nos. NPF-87 and NPF-89 for Comanche Peak 
Steam Electric Station, Units 1 and 2, respectively, be deleted.
    Date of issuance: March 22, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 94 and 94.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
delete Appendix C from the Licenses.
    Date of initial notice in Federal Register: August 20, 2001 (66 FR 
43595). The supplemental letters provided clarifying information that 
did not change the staff's proposed no significant hazards 
consideration determination or expand the application beyond the scope 
of the Federal Register notice.
    The Commission's related evaluation of the amendments are contained 
in a Safety Evaluation dated March 22, 2002, and its attachment.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 25, 2001, as supplemented by 
letter dated February 18, 2002.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 4.2.1, ``Fuel Assemblies,'' for Comanche Peak Steam 
Electric Station (CPSES), Units 1 and 2, to allow the use of 
ZIRLOTM test assemblies and to further allow, ``* * * A 
limited number of lead test assemblies * * *be placed in non-limiting 
core regions.''
    Date of issuance: March 26, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 95 and 95.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR

[[Page 18655]]

64306). The supplemental letter dated February 18, 2002, provided 
clarifying information that did not change the staff's original no 
significant hazards consideration determination or expand the scope of 
the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 26, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 3, 2001, as supplemented by 
letters dated October 22 and December 18, 2001, and March 7, 2002.
    Brief description of amendment: The amendment relocates certain 
reactor coolant system cycle-specific parameter limits from the 
technical specifications (TS) to the Core Operating Limits Report 
(COLR), and thus expands the COLR. Additionally, TS 5.6.5, ``Core 
Operating Limits Report (COLR),'' is revised to allow topical reports 
to be identified by title and number only.
    Date of issuance: March 28, 2002.
    Effective date: March 28, 2002, and shall be implemented, including 
relocating the requirements from the TSs to the COLR, as specified in 
the licensee's letters of April 3, October 22, and December 18, 2001, 
and March 7, 2002, and the Safety Evaluation attached to Amendment No. 
144, prior to the startup from Refueling Outage 12, which is scheduled 
for the spring of 2002.
    Amendment No.: 144.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 2, 2001 (66 FR 
22036) and February 5, 2002 (67 FR 5342). The March 7, 2002, 
supplemental letter provided additional clarifying information that did 
not expand the scope of the application as noticed and did not change 
the staff's proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 28, 2002.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland. Publicly available records will be 
accessible from the Agencywide Documents Assess and Management Systems 
(ADAMS) Public Electronic Reading Room on the internet at the NRC Web 
site, http://www.nrc.gov/NRC/ADAMS/index.html. If you do not have 
access to ADAMS or if there are problems in accessing the documents 
located in ADAMS, contact the NRC Public Document room (PDR) Reference 
staff at 1-800-397-4209, 304-415-4737 or by e-mail to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By May 16, 2002, the licensee 
may file a request for a hearing with respect to issuance of the 
amendment to the subject facility operating license and any person 
whose

[[Page 18656]]

interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, located at One White Flint North, 
11555 Rockville Pike (first floor), Rockville, Maryland 20852, and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room). If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, by the above date. A copy of 
the petition should also be sent to the Office of the General Counsel, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-001, and to 
the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-318, Calvert 
Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

    Date of application for amendment: April 1, 2002.
    Brief description of amendment: The amendment increases the allowed 
outage time of one train of the control room emergency ventilation 
system from 14 to 21 days (for the loss of the emergency power supply 
only). This is a one-time change to support corrective maintenance and 
inspections of the 1A diesel generator during the Unit 1 refueling 
outage.
    Date of issuance: April 4, 2002.
    Effective date: As of the date of issuance.
    Amendment No.: 227.
    Renewed License No. DPR-69: Amendment revised the Technical 
Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No. The Commission's related evaluation of the 
amendment, finding of emergency circumstances, and final determination 
of no significant hazards consideration, are contained in a Safety 
Evaluation dated April 4, 2002.

    For the Nuclear Regulatory Commission.
    Dated at Rockville, Maryland, this 8th day of April 2002.
Ledyard B. Marsh,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 02-8866 Filed 4-15-02; 8:45 am]
BILLING CODE 7590-01-P