[Federal Register Volume 68, Number 101 (Tuesday, May 27, 2003)]
[Notices]
[Pages 28843-28864]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-12973]


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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, May 2, 2003, through May 15, 2003. The last 
biweekly notice was published on May 13, 2003 (68 FR 25648).

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

[[Page 28844]]

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 Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By June 26, 2003, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714, which is 
available at the Commission's PDR, located at One White Flint North, 
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System's (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, Public File Area 01F21, 11555 
Rockville Pike (first floor), Rockville, Maryland, by the above date. 
Because of continuing disruptions in delivery of mail to United States 
Government offices, it is requested that petitions for leave to 
intervene and requests for hearing be transmitted to the Secretary of 
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the request for 
hearing and petition for leave to intervene should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and because of continuing disruptions in 
delivery of mail to United States Government offices, it is requested 
that copies be transmitted either by means of facsimile transmission to 
301-415-3725 or by e-mail to [email protected]. A copy of the 
request for hearing and petition for leave to intervene should also be 
sent 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

[[Page 28845]]

public inspection at the Commission's PDR, located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management System's (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, 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 amendment request: April 21, 2003.
    Description of amendment request: The licensee proposed to revise 
Sections 3.7 and 4.7, ``Auxiliary Electrical Power,'' of the Technical 
Specifications (TSs) to make them generally consistent with Nuclear 
Regulatory Commission (NRC) guidance set forth in NUREG-1433, 
``Standard Technical Specifications, General Electric Plants, BWR 
[Boiling Water Reactor]/4,'' Revision 2, and with the NRC-approved 
industry guidance identified as Technical Specification Task Force 
(TSTF) traveler TSTF-360, Revision 1. The amendment would also add a 
new Section 6.8.5, ``Station Battery Monitoring and Maintenance 
Program.'' The resulting Sections 3.7, 4.7, and 6.8.5 will be 
explicitly applicable to station batteries B and C, both safety-related 
subsystems, and their associated battery chargers. The proposed 
amendment would revise requirements concerning surveillance, 
monitoring, and maintenance of the subject batteries and chargers.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the three standards of 10 CFR 50.92(c) and performed its own. 
The NRC staff's analysis is presented below:
    The first standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated. The proposed changes, if approved by the NRC, will be made 
in a manner such that conservatism is maintained through compliance 
with applicable NRC regulations and guidance. No hardware design change 
is involved with the proposed amendment, thus there will be no adverse 
effect on the functional performance of any plant structure, system, or 
component (SSC). Consequently, all SSCs will continue to perform their 
design functions with no decrease in their capabilities to mitigate the 
consequences of postulated accidents. Station battery surveillance, 
monitoring, and maintenance were not previously factored into the 
probability of accidents, nor were they factored into scenarios of 
previously analyzed accidents. Consequently, the proposed revision to 
Sections 3.7, 4.7, and 6.8.5 of the TSs will lead to no increase in the 
consequences of an accident previously evaluated, and no increase of 
the probability of an accident previously evaluated.
    The second standard requires that operation of the unit in 
accordance with the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. The proposed amendment is not the result of a hardware 
design change, nor does it lead to the need for a hardware design 
change. There is no change in the methods the unit is operated. As a 
result, all SSCs will continue to perform as previously analyzed by the 
licensee, and previously evaluated and accepted by the NRC staff. 
Therefore, the proposed amendment will not create the possibility of a 
new or different kind of accident from any previously evaluated.
    The third standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
reduction in a margin of safety. Since the licensee did not propose to 
exceed or alter a design basis or safety limit, the proposed amendment 
will not affect in any way the performance characteristics and intended 
functions of any SSC. Therefore, the proposed amendment does not 
involve a significant reduction in a margin of safety.
    Based on the NRC staff's analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendments request: April 15, 2003.
    Description of amendments request: The amendments would revise 
Sections 2.2, ``SL [Safety Limits] Violations,'' for reporting such 
violations to positions in the plant organization; 5.2.1, ``Onsite and 
Offsite Organization,'' for the position responsible for overall safe 
plant operation; and 5.5.1, ``Offsite Dose Calculation Manual (ODCM),'' 
for the position that approves changes to the ODCM, of the Technical 
Specifications (TSs). The revisions would account for the elimination 
of the positions of Vice President, Nuclear Production, and Director, 
Site Chemistry, and the assignment of the responsibilities of these 
positions in the above TS sections to other positions in the plant 
organization. Also, there would be the format change of adding the 
title of Section 2.2 near the top of TS page 2.0-2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    These changes involve minor changes in the organization of 
PVNGS. It is expected that the organizational changes will have a 
positive effect on the conduct of plant operations and safety-
related work. Functions which are necessary to operate the facility 
safely and in accordance with the operating licenses, remain in the 
re-aligned organization and will not affect the safe operation of 
the plant and continue to ensure proper control of administrative 
activities. The Quality Assurance (QA) organization reporting 
structure has not been affected by these changes allowing the QA 
organization to maintain the required authority and organizational 
freedom to identify quality problems; to initiate, recommend, or 
provide solutions; and to verify implementation of solutions. The 
proposed changes will not affect the operation of structures, 
systems, [or] components, and will not reduce programmatic controls 
such that plant safety would be affected. (The changes in the plant 
organization are also not initiators of an accident.) 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.

[[Page 28846]]

    The proposed changes will not affect the operation of 
structures, systems, [or] components, and will not reduce 
programmatic controls such that plant safety would be affected. The 
changes in the organization will continue to provide necessary 
oversight and control of administrative processes. [The changes in 
the plant organization are also not initiators of an accident.] 
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.
    These changes are administrative and will not diminish any 
organizational or administrative controls currently in place. The 
proposed changes will not affect the operation of structures, 
systems, [or] components, and will not reduce programmatic controls 
such that plant safety would be affected. Therefore, the proposed 
changes do not involve a significant reduction in a margin of 
safety.
    Based on the above, APS concludes that the activities associated 
with the proposed amendment(s) present no significant hazards 
consideration under the standards set forth in 10 CFR 50.92 
``Issuance of Amendment,'' (c) and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona 
Public Service Company, PO Box 52034, Mail Station 7636, Phoenix, 
Arizona 85072-2024.
    NRC Section Chief: Stephen Dembek.

Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-317, Calvert 
Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

    Date of amendment request: May 1, 2003.
    Description of amendment request: The proposed amendment would 
increase the maximum enrichment limit of the fuel assemblies that can 
be stored in the Unit 1 spent fuel pool by taking credit for soluble 
boron in maintaining acceptable margins of subcriticality. 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.
    The proposed change will increase the maximum enrichment limit 
of the fuel assemblies that can be stored in the Unit 1 spent fuel 
pool (SFP) by taking credit for soluble boron in maintaining 
acceptable margins of subcriticality. The proposed change will 
modify Technical Specification 4.3.1 ``Criticality'' and add 
Technical Specification 3.7.16 ``Spent Fuel Pool Boron 
Concentration.'' The postulated accidents for the SFP are basically 
four types: (1) dropped fuel assembly on top of the storage rack, 
(2) a misloading accident, (3) an abnormal location of a fuel 
assembly, and (4) loss-of-normal cooling to the SFP.
    There is no increase in the probability of a fuel assembly drop 
accident in the SFP when considering the higher enriched fuel or the 
presence of soluble boron in the SFP water. Dropping a fuel assembly 
on top of the SFP storage racks is not credible at Calvert Cliffs 
due to the design of the spent fuel handling machine and due to the 
height of the SFP storage racks. The handling of the fuel assemblies 
has always been performed in borated water and will not change as a 
result of crediting soluble boron in the SFP criticality analysis. 
The proposed change does not change the general design and 
characteristics of the fuel assemblies. Therefore, the proposed 
change does not increase the probability of a fuel assembly drop 
accident.
    There is no increase in the probability of the accidental 
misloading of irradiated fuel assemblies into the SFP storage racks 
when considering the higher enriched fuel or the presence of soluble 
boron in the SFP water for criticality control. Fuel assembly 
placement will continue to be controlled pursuant to approved fuel 
handling procedures.
    Due to the design of the SFP storage racks, an abnormal 
placement of a fuel assembly into the SFP storage racks is not 
possible. Also, the design of the SFP prevents an inadvertent 
placement of a fuel assembly between the outer most storage cell and 
the pool wall. The proposed change does not make any change to the 
design of SFP. Therefore, there is no increase in the probability of 
abnormal placement of a fuel assembly into the SFP storage racks.
    The proposed change will not result in any changes to the SFP 
cooling system, and the fuel assembly design and characteristics are 
not changed by an increase in fuel enrichment. Therefore, there is 
no increase in the probability of a loss of SFP cooling. Also, since 
a high concentration of soluble boron has always been maintained in 
the SFP water, there is no increase in the probability of the loss 
of normal cooling to the SFP water considering the presence of 
soluble boron in the pool water for criticality control.
    There is no increase in the consequences of an accidental drop 
or accidental misloading of a maximum enriched fuel assembly into 
the SFP storage racks, because the criticality analysis demonstrates 
that the pool will remain subcritical following either event, even 
if the pool contains a boron concentration less than the proposed 
Technical Specification limit. The proposed Technical Specification 
limit will ensure that an adequate SFP boron concentration will be 
maintained.
    There is no increase in the consequences of a loss-of-normal SFP 
cooling because the Technical Specification boron concentration 
provides significant negative reactivity. Loss of the SFP water via 
boiling will not result in a loss of soluble boron, since the 
soluble boron is not volatile. Therefore, loss of spent fuel pool 
cooling system without makeup flow is not a mechanism for boron 
dilution. Even in the unlikely event that soluble boron in the SFP 
is completely diluted via unborated makeup flow, a pool completely 
filled with maximum enriched unburned assemblies will remain 
subcritical by a design margin of k-effective not to exceed 0.986.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change will increase the maximum enrichment limit 
of the fuel assemblies that can be stored in the Unit 1 SFP by 
taking credit for soluble boron in maintaining acceptable margins of 
subcriticality. Increasing the maximum enrichment limit does not 
create a new type of criticality accident.
    Soluble boron has been maintained in the SFP water and is 
currently required by procedures. Therefore, crediting soluble boron 
in the SFP criticality analysis will have no effect on normal pool 
operation and maintenance. Crediting soluble boron will only result 
in increased sampling to verify the boron concentration. This 
increased sampling will not create the possibility of a new or 
different kind of accident.
    A dilution of the SFP soluble boron has always been a 
possibility. However, the boron dilution event previously had no 
consequences, since boron was not previously credited in the 
accident analysis. The initiating events that were considered for 
having the potential to cause dilution of the boron in the SFP to a 
level below that credited in the criticality analyses fall into 
three categories: dilution by flooding, dilution by loss-of-coolant 
induced makeup, and dilution by loss-of-cooling system induced 
makeup. The spent fuel pool dilution analysis demonstrates that a 
dilution that could increase the rack k-effective greater than 0.95 
is not a credible event. It is not credible that dilution could 
occur for the required length of time without operator notice, since 
this event would activate the high level alarm and initiate 
Auxiliary Building flooding. In addition, in excess of 1,043,000 
gallons of unborated water must be added to the SFP to reach the 
minimum soluble boron concentration. This is more water volume than 
is contained in both pretreated water storage tanks and also more 
water volume than is contained in the demineralized water storage 
tank and both condensate storage tanks combined. Even in the 
unlikely event that soluble boron in the SFP is completely diluted, 
the SFP will remain subcritical by a design margin of k-effective 
will not exceed 0.986.
    The proposed change will not result in any other change in the 
plant configuration or equipment design. Therefore, the proposed

[[Page 28847]]

change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The Technical Specification changes proposed by this license 
amendment request will provide an adequate safety margin to ensure 
that the stored fuel assembly array of maximum enriched fuel will 
always remain subcritical. Those limits are based on a plant 
specific criticality analysis performed for the Calvert Cliffs Unit 
1 SFP, that include technically supported margins.
    While the criticality analysis utilized credit for soluble 
boron, the SFP rack k-effective will remain less than 0.986 with no 
soluble boron with a 95 percent probability at a 95 percent 
confidence level. This substantial reduction in the SFP soluble 
boron concentration was evaluated and shown not to be credible. 
Soluble boron is used to provide subcritical margin such that the 
spent fuel pool k-effective is maintained less than or equal to 
0.95. Since k-effective is less than or equal to 0.95, the current 
margin of safety is maintained.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposed to determine that the 
amendment 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: Richard J. Laufer.

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: April 17, 2003.
    Description of amendments request: The proposed amendment would (1) 
make 19 specific changes to the Technical Specifications actions 
currently requiring suspension of operations involving positive 
reactivity additions, and (2) revise various notes precluding reduction 
in boron concentration. The proposed changes follow the guidance of 
Technical Specification Task Force (TSTF) Change Traveler 286, Revision 
2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The intent of this change is to clarify those Technical 
Specifications involving positive reactivity additions to the 
shutdown reactor so that small, controlled, safe insertions of 
positive reactivity will be allowed where they are now categorically 
prohibited, posing operational difficulties. These controlled 
activities could result in a slight change in the probability of an 
event occurring as Reactor Coolant System (RCS) manipulations that 
are currently prohibited would now be allowed. However, RCS 
manipulations are rigidly controlled to minimize the possibility of 
a significant reactivity increase. In addition, there is sufficient 
shutdown margin available in these conditions to allow for these 
slight reactivity changes without significantly increasing the 
probability of an accident previously evaluated.
    The proposed change does not permit the shutdown margin required 
by the Technical Specifications to be reduced. While the proposed 
change will permit changes in the discretionary boron concentration 
above the technical specification requirements, this excess 
concentration is not credited in the Updated Final Safety Analysis 
Report safety analysis. Because the initial conditions assumed in 
the safety analysis are preserved, no increase in the consequence of 
an accident previously evaluated would occur. In addition, small 
temperature changes in the RCS impose reactivity changes by means of 
the moderator temperature coefficient of reactivity. These small 
changes are within the required shutdown margin, therefore, there is 
no increase in the consequence of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    This proposed amendment allows for minor plant operational 
adjustments without adversely impacting the safety analysis required 
shutdown margin. It does not involve any change to plant equipment 
or the shutdown margin requirements in the Technical Specifications.
    Therefore, the proposed change will 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 in Modes 3, 4, 5, and 6 is preserved by the 
calculated shutdown margin which prevents a return to criticality. 
The proposed change will permit reductions in the discretionary 
shutdown margin beyond the Technical Specification requirements. 
However, the shutdown margin required by the Technical 
Specifications is not changed. The proposed change only affects 
Reactor Coolant System temperature and boron concentration above the 
calculated shutdown margin. By not impacting the shutdown margin, 
the margin of safety is not affected.
    Therefore, the proposed change 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: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: February 13, 2003.
    Description of amendment request: The proposed amendment would 
allow the use of an alternative source term (AST) methodology in 
accordance with 10 CFR 50.67 based on a reevaluation of the loss-of-
coolant accident (LOCA) design-basis accident (DBA). Using an approved 
AST, the licensee has also proposed changes to increase the allowable 
secondary containment bypass and main steam isolation valve (MSIV) 
leakage limits and eliminate the MSIV leakage control system. The 
licensee also proposed changes to the TS Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The implementation of AST assumptions has been evaluated in a 
revision to the analysis of the Loss of Coolant Accident (LOCA) and 
an update to the analysis of the Fuel Handling Accident (FHA).
    Based upon the results of the analyses, it has been demonstrated 
that, with the requested changes, the dose consequences of these 
limiting Design Basis Accidents (DBAs) are within the regulatory 
guidance provided by the NRC for use with the AST. This guidance is 
presented in 10 CFR 50.67, Regulatory Guide 1.183 [``Alternative 
Radiological Source Terms For Evaluating Design Basis Accidents At 
Nuclear Power Reactors''], and Standard Review Plan (SRP) Section 
15.0.1.
    The requirements for MSIV [main steam isolation valve] Leakage 
Control System operability for eliminating MSIV leakage to the 
environment are being eliminated. This is acceptable because, with 
the application of AST, this system is no longer credited in 
mitigating the consequences of a LOCA or any other DBA.
    The proposed changes also increase the limits on maximum 
allowable leakage from

[[Page 28848]]

secondary containment bypass and main steam isolation valves, and on 
unfiltered inleakage into the Control Room. This is acceptable due 
to the new assumptions used in calculating Control Room and offsite 
dose following the affected design basis accident using the AST 
methodology.
    The proposed changes do not affect the normal design or 
operation of the facility before the accident; rather, once the 
occurrence of an accident has been postulated, the new source term 
is an input to evaluate the consequence. The radiological 
consequences of the analyzed DBAs have been evaluated with 
application of AST assumptions. The results conclude that the 
radiological consequences remain within applicable regulatory 
limits. Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The application of AST does not affect the design, functional 
performance or normal operation of the facility. Similarly, it does 
not affect the design or operation of any component in the facility 
such that new equipment failure modes are created. Elimination of 
the MSIV Leakage Control System cannot create a new accident because 
it is used as a mitigation system to limit MSIV leakage after the 
accident has occurred. Similarly, the use of Standby Liquid Control 
System to buffer suppression pool pH to prevent iodine reevolution 
is another mitigation function credited after the accident has 
occurred and; therefore, cannot create a new accident.
    As such the proposed changes will not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    This proposed license amendment involves changes from the 
original source term developed in accordance with Technical 
Information Document (TID) 14844 to a new AST, as described in 
Regulatory Guide 1.183. The results of the DBA analyses and the 
requested Technical Specification changes, are subject to revised 
acceptance criteria. The analyses have been performed using 
conservative methodologies.
    Safety margins and analytical conservatisms have been evaluated 
and have been found acceptable. The analyzed events have been 
carefully selected and margin has been retained to ensure that the 
analysis adequately bounds postulated event scenario. The dose 
consequences of these limiting events are within the acceptance 
criteria presented in 10 CFR 50.67, Regulatory Guide 1.183 and SRP 
Section 15.0.1.
    The margin of safety is that provided by meeting the applicable 
regulatory limits. The effect of relaxation of these design and 
Technical Specification requirements has been analyzed and doses 
resulting from the design basis accidents have been found to remain 
within the regulatory limits. The changes continue to ensure that 
the doses at the exclusion area and low population zone boundaries, 
as well as the control room, are within the corresponding regulatory 
limits.
    Therefore, operation of Fermi 2 in accordance with the proposed 
changes 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: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: February 13, 2003.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) Section 5.5.10, ``Technical 
Specification (TS) Bases Control Program,'' to be consistent with 
changes made to 10 CFR 50.59, which were published in the Federal 
Register on October 4, 1999 (64 FR 53582), and which became effective 
March 13, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change deletes the reference to ``unreviewed safety 
question'' as defined in 10 CFR 50.59. Deletion of the definition of 
``unreviewed safety question'' was approved by the NRC with the 
revision of 10 CFR 50.59. This change is administrative in nature. 
Consequently, the probability of an accident previously evaluated is 
not significantly increased. Changes to the TS Bases are still 
evaluated in accordance with 10 CFR 50.59. As a result, the 
probability or consequences of any accident previously evaluated are 
not significantly affected. There is no increase in the radiological 
dose at the site boundary for any previously evaluated accident. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration of 
the plant (i.e., no new or different types of equipment will be 
installed) or a change to the methods governing normal plant 
operation. These changes are considered administrative in nature and 
do not modify, add, delete, or relocate any technical requirements 
in the TS. Therefore, this change does not create the possibility of 
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 proposed change will not reduce a margin of safety because 
it has no direct effect on any of the safety analysis assumptions. 
Changes to the TS Bases that result in meeting the criteria in 
paragraph 10 CFR 50.59(c)(2) continue to require NRC approval 
pursuant to 10 CFR 50.59. This change is administrative in nature 
based on the revision to 10 CFR 50.59. Therefore, the proposed 
change does not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: March 31, 2003.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Surveillance Requirement (SR) 
3.7.3.6 associated with the verification of control room emergency 
filtration (CREF) system duct work unfiltered inleakage. This amendment 
request supercedes DECo's previous amendment request dated September 
26, 2002, in its entirety. The September 26, 2002, amendment request 
was previously noticed in the Federal Register on November 26, 2002 (67 
FR 70765).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    This license amendment proposes an alternative test for 
performing the (Control Room Emergency Filtration) CREF system 
surveillance associated with measuring the

[[Page 28849]]

Control Room Envelope (CRE) unfiltered inleakage. The CREF system 
provides a configuration for mitigating radiological consequences of 
accidents; however, it does not involve the initiation of any 
previously analyzed accident. Similarly, the implementation of 
compensatory measures to address the failure of the surveillance to 
meet the design basis unfiltered inleakage limits is required to 
mitigate the consequences of a radiological release. Therefore, the 
proposed changes cannot increase the probability of any previously 
evaluated accident.
    The CREF system provides a radiologically controlled environment 
from which the plant can be safely operated following a radiological 
accident. Design basis accident analyses conclude that radiological 
consequences are within the regulatory acceptance criteria. The 
current TS surveillance (SR 3.7.3.6) measures inleakage from four 
sections of CREF system duct work outside the CRE that are at 
negative pressure during accident conditions. The proposed Tracer 
Gas test provides a measurement of CRE inleakage from all potential 
sources including the four sections of duct work. Measuring the CRE 
inleakage using Tracer Gas testing has no effect on the CREF system 
function. The results of Tracer Gas testing will be evaluated 
against the assumptions in the approved Alternative Source Term 
(AST) design basis accident analyses and compensatory measures will 
be implemented, as necessary, to ensure compliance with 10 CFR 
50.67. If compliance with 10 CFR 50.67 cannot be demonstrated or if 
compensatory measures have been in place for more than 18 months, a 
conservative plant shutdown will be required to minimize risk. 
Therefore, the proposed changes do not significantly increase the 
radiological consequences of any previously evaluated accident.
    Based on the above, the proposed changes do not significantly 
increase the probability or consequences of any accident previously 
evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not alter the design function or 
operation of the system involved. The CREF system will still provide 
protection to control room occupants in case of a significant 
radioactive release. The revised TS surveillance requirements 
provide an alternative test method that has been widely accepted for 
the measurement of CRE unfiltered inleakage. The proposed changes do 
not introduce any new modes of plant or CREF system operation. 
Therefore, the proposed changes do not create the potential for a 
new or different kind of accident from any accident previously 
evaluated.
    3. The changes do not involve a significant reduction in the 
margin of safety.
    The proposed changes to the Fermi 2 TS surveillance requirements 
do not affect the radiological release from a design basis accident 
nor the postulated dose to the control room occupants as a result of 
the accident. The alternate surveillance test requirements provide 
an acceptable approach for the measurement of CRE inleakage. Safety 
margins and analytical conservatisms are included in the analyses to 
ensure that all postulated event scenarios are bounded. The proposed 
TS requirements continue to ensure that the radiological 
consequences at the control room are below the corresponding 
regulatory guidelines and that compliance with 10 CFR 50.67 and GDC 
(General Design Criterion)-19 is not affected. Therefore, the 
proposed changes 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: L. Raghavan.

Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone 
Power Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: April 7, 2003
    Description of amendment request: The proposed amendment would move 
selected Technical Specification (TS) parameters to the Core Operating 
Limits Reports (COLR). Specifically, the changes proposed affect TSs 
2.2, ``Limiting Safety System Settings, Table 2.2-1;'' 3/4.1.1.1.1, 
``Reactivity Control Systems, Boration Control, SHUTDOWN MARGIN--Modes 
3, 4, and 5 Loops Filled;'' 3/4.1.1.2, ``Reactivity Control Systems, 
SHUTDOWN MARGIN--Cold Shutdown--Loops Not Filled;'' 3/4.2.5, ``Power 
Distribution Limits, DNB Parameters;'' 3/4.3.5, ``Instrumentation, 
SHUTDOWN MARGIN Monitor;''
3/4.9.1.1, ``Refueling Operations, Boron Concentration;'' Section 
6.9.1.6.a, ``Core Operating Limits Report, Core Operating Limits;'' and 
Section 6.9.1.6.b, ``Core Operating Limits Report, The Analytical 
Methods Used to Determine the Core Operating Limits,'' and the 
corresponding pages and Bases sections.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR), Sec.  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 relocation of cycle-specific core operating limits from the 
technical specifications to the COLR has no influence or impact on 
the probability or consequences of a Design Basis Accident. 
Adherence to the COLR and methodologies acceptable for establishing 
COLR parameters continues to be controlled by Technical 
Specifications. The proposed amendment still requires exactly the 
same actions to be taken when or if limits are exceeded. Each 
accident analysis addressed in the Final Safety Analysis Report 
(FSAR) will be examined with respect to the changes in cycle-
dependent parameters, which are obtained from application of the 
Nuclear Regulatory Commission (NRC) approved reload design 
methodologies, to ensure that the transient evaluation of new core 
designs are bounded by previously accepted analysis. This 
examination, which will be performed in accordance with the 
requirements of 10 CFR 50.59, ensures that future designs will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed change to add new document references to Technical 
Specification Sections 6.9.1.6.b.16 and 6.9.1.6.b.17 are required to 
identify the most recent methodology to be used in the Millstone 
Unit No. 3 Small Break Loss of Coolant Accident (SBLOCA) analysis. 
Section 6.9.1.6.b.18 is added to describe NRC approved Overpower DT 
and Overtemperature DT trip function methodology. The use of these 
methodologies demonstrates that the acceptance criteria for SBLOCA 
events and Overpower DT and Overtemperature DT are met. This change 
has no impact on plant equipment operation. Since these changes only 
affect the method of analysis, they cannot affect the likelihood or 
consequences of accidents. Therefore, these changes will not 
increase the probability or consequences of an accident previously 
evaluated.
    Deleting the revision number and the date from the documents 
contained in Technical Specification Section 6.9.1.6.b.1 and in 
Technical Specification Sections 6.9.1.6.b.4 through 6.9.1.6.b.10 
has no impact on the actual analytical methods used to determine the 
core operating limits, nor does it affect the likelihood or 
consequences of accidents. Therefore, this change 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.
    As stated earlier, the relocation of the cycle-specific 
variables to the COLR, adding new document references and deleting 
the revision number and the date in Technical Specification Section 
6.9.1.6.b have no influence or impact, nor does it contribute in any 
way to the probability or consequences of an accident. No safety 
related equipment, safety function, or plant operations will be 
altered as a result of this proposed change. The cycle specific 
variables are calculated using NRC-approved methods and submitted to 
the NRC to allow the Staff to continue to trend the values of these 
limits. The Technical Specifications will continue to require 
operation within the required core operating limits and appropriate 
actions will be taken when or if limits are exceeded. Therefore the 
proposed amendment does not in any way create the possibility of a 
new or

[[Page 28850]]

different kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes have no impact on plant equipment 
operation. The proposed changes do not revise any setpoints assumed 
in the analyses and do not affect the acceptance criteria for SBLOCA 
analyses. Therefore, the proposed 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., Waterford, CT 06141-5127.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: April 10, 2003.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) for the low temperature 
overpressure protection system. Currently, TS Surveillance Requirement 
(SR) 3.4.12.5 requires performance of a channel functional test for the 
power-operated relief valve within 12 hours of decreasing reactor 
coolant system (RCS) temperature to <= 325 [deg]F and every 31 days 
thereafter. The proposed amendments would revise TS SR 3.4.12.5 to 
allow the first performance of this surveillance to be within 31 days 
prior to decreasing RCS temperature to <= 325 [deg]F. The proposed 
amendments also would revise the frequency of the channel calibration 
in TS SR 3.4.12.7 from 18 months to 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:

    Pursuant to 10 CFR 50.91, Duke Power Company (Duke) has made the 
determination that this amendment request involves a No Significant 
Hazards Consideration by applying the standards established by the 
NRC regulations in 10 CFR 50.92. This ensures 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:
    This is a revision to the Technical Specification (TS) 
surveillance requirement (SR) for performing the channel functional 
test (CFT) for the pressurizer [power] [..] operated relief valve 
(PORV). As such, changing the requirement to perform the first CFT 
before entering the Low Temperature Overpressure Protection (LTOP) 
region, rather than after LTOP is required, eliminates removing the 
PORV from service, in the mode of applicability for the performance 
of the CFT. This change will decrease the probability of a low 
temperature overpressurization of the reactor vessel, thereby 
increasing safety and reducing risk, by maintain(ing) both trains 
(active and passive) of the LTOP System operable. The change to the 
frequency for performance of SR 3.4.12.7 is being done to ensure the 
calibration is performed in a time frame supported by current 
analysis. The method of test is not changed, only the frequency. 
This reduction in frequency will not significantly increase the 
probability or consequences of any accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated:
    This revision will not impact the LTOP evaluation analysis. The 
timeframe to perform the CFT for the PORV will not change the 
operation of the PORV or its function during accident conditions. No 
new or different accidents result from performing the CFT prior to 
entering LTOP conditions. The revision to SR 3.4.12.7 only changes 
the frequency of the testing. The method of test is not changed. 
This change has no effect on the possibility of a new or different 
kind of accident.
    (3) Involve a significant reduction in a margin of safety:
    The proposed revision will perform the CFT within 31 days prior 
to entering LTOP conditions, rather than performing the test once 
LTOP conditions are entered. This allows the CFT, which causes the 
PORV to be inoperable for a short period of time, to be performed 
prior to reaching the plant conditions where the PORV is relied upon 
for LTOP. Performing the CFT within 31 days prior to decreasing RCS 
temperature to < 325 [deg][F], rather than after entering these 
conditions, will not change the margin of safety. Oconee 
calculations show that a recalibration interval of 6 months for the 
Reactor Coolant System (RCS) low range pressure instrumentation 
results in a single-sided 95/95 probability confidence limit of 9.4 
psig. This result is bounded by the instrument uncertainty assumed 
in the LTOP evaluation analysis. The frequency change for SR 
3.4.12.7 from 18 months to 6 months does not affect the method of 
test performance. It only decreases the allowed time between 
performances to reflect current Oconee analysis. This will not 
significantly reduce 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: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: John A. Nakoski.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of amendment request: July 5, 2002, as supplemented August 13, 
2002.
    Description of amendment request: The proposed amendment would 
relocate portions of Technical Specification (TS) 3/4.6.B, ``Primary 
System Boundary--Coolant Chemistry,'' from the TSs to the Updated Final 
Safety Analysis Report (UFSAR). The portions of the TS that would be 
relocated to the UFSAR are the reactor coolant chemistry requirements 
for conductivity and chloride concentration. Specifically, TSs 3/
4.6.B.2, 3/4.6.B.3, and 3.6.B.4 would be relocated to the UFSAR.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Response: No. The proposed change is administrative in nature 
and does not involve the modification of any plant equipment or 
affect basic plant operation. Conductivity and chloride limits are 
not assumed to be an initiator of any analyzed event, nor are these 
limits assumed in the mitigation of consequences of accidents.
    Therefore, the proposed change does 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 change does not involve any physical 
alteration of plant equipment and does not change the method by 
which any safety-related system performs its function. As such, no 
new or different types of equipment will be installed, and the basic 
operation of installed equipment is unchanged. The methods governing 
plant operation and testing remain consistent with current safety 
analysis assumptions. Therefore, the proposed change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The proposed change represents the relocation of 
current Technical

[[Page 28851]]

Specification requirements to the UFSAR, based on regulatory 
guidance and previously approved changes for other stations. The 
proposed change is administrative in nature, does not negate any 
existing requirement, and does not adversely affect existing plant 
safety margins or the reliability of the equipment assumed to 
operate in the safety analysis. As such, there are no changes being 
made to safety analysis assumptions, safety limits or safety system 
settings that would adversely affect plant safety as a result of the 
proposed change. Margins of safety are unaffected by requirements 
that are retained, but relocated from the Technical Specifications 
to the UFSAR. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J.M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts, 02360-5599.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: May 1, 2003.
    Description of amendment request: The proposed amendment would 
modify the surveillance testing requirements for the containment spray 
system (CSS) by deleting the requirement to verify the position of 
valves that are locked, sealed, or otherwise secured in their correct 
position and replacing the quantitative allowable pump degradation 
value with a requirement to verify the pumps perform in accordance with 
the Inservice Testing Program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Analyzed events are assumed to be initiated by the failure of 
plant structures, systems, or components. Altering the surveillance 
requirements for the CSS does not increase the probability that a 
failure leading to an analyzed event will occur. The CSS components 
are passive until an actuation signal is generated. This change does 
not increase the failure probability of the CSS components. 
Therefore, the probability of occurrence for a previously analyzed 
accident is not significantly increased.
    The CSS is primarily designed to mitigate the consequences of a 
loss of coolant accident (LOCA) or main steam line break (MSLB) 
accident. The proposed change does not affect any of the assumptions 
used in the deterministic LOCA or MSLB analyses. Hence the 
consequences of accidents previously evaluated do not change.
    Therefore, the change associated with modifying the CSS 
surveillance requirements does not involve an 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 change does not change the design or configuration 
of the plant. No new equipment is introduced, nor will any installed 
equipment be operated in a new or different manner. No changes are 
proposed to the plant's operating parameters or setpoints at which 
protective or mitigative actions are initiated. Additionally, no 
substantive changes are proposed to the procedures which ensure the 
plant remains within analyzed limits or the procedures relied upon 
to respond to off-normal events. As such, no new failure modes are 
being introduced. The proposed change does not alter assumptions 
made in the safety analysis or licensing basis.
    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 change associated with modifying the surveillance 
requirements for the CSS does not affect the limiting conditions for 
operation used in the deterministic analysis to establish the margin 
of safety. The margin of safety is established through equipment 
design, operating parameters, and the setpoints at which automatic 
actions are initiated. None of these are adversely impacted by the 
proposed change. Sufficient equipment remains available to actuate 
upon demand for the purpose of mitigating a transient event.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: March 11, 2003.
    Description of amendment request: The proposed amendment will 
revise and relocate Surveillance Requirement (SR) 4.0.5 and SR 4.4.9 to 
the administrative section of the Technical Specifications (TS) under 
sections 6.5.8 and 6.5.7, respectively. The proposed amendment will 
also relocate TS 3.4.9, ``Reactor Coolant System Structural Integrity'' 
and its Bases to the Waterford Steam Electric Station, Unit 3 
(Waterford 3) Technical Requirements Manual (TRM). Additionally, the 
proposed amendment extends the Waterford 3 flywheel volumetric 
examination interval to ten years.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to relocate SR 4.0.5 to the administrative 
section of the TSs, including modifications to the wording to make 
it consistent with NUREG-1432, will not reduce the current testing 
and inspection requirements. The performance of a code (American 
Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel 
Code) inservice test is not an accident initiator. Verbally issuing 
relief to the ASME Code by the NRC (Nuclear Regulatory Commission) 
staff in lieu of written relief does not reduce assurance of the 
health and safety of the public since the NRC staff still reviews 
the basis for the relief on its technical merit and the NRC staff 
still obtains management approval prior to granting the relief.
    Inspections of the reactor coolant pump (RCP) flywheels are 
conducted to detect a flaw in the flywheel prior to it becoming a 
missile that could damage other portions of the facility. The 
fracture mechanics analyses conducted as part of NRC approved 
Topical Report SIR-94-080-A, Rev(ision) 1 shows that a 
conservatively sized pre-existing crack will not grow to a flaw size 
necessary to create flywheel missiles within the current or extended 
life of the facility thus the flywheel will remain intact and 
perform its function to reduce the rate of decay of coolant flow 
during a postulated loss of power to the RCP motor. This analysis 
conservatively assumes minimum material properties, maximum flywheel 
speed, location of the flaw in the highest stress area, and a number 
of startup and shutdown cycles higher than expected. Since a 
conservative flaw in the RCP flywheels will not grow to the 
allowable flaw size under large break LOCA (loss-of-coolant

[[Page 28852]]

accident) conditions over the life of the plant, reducing the 
inspection frequency of the flywheels will not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    The change to move the surveillance requirements for the RCP 
flywheels to the programs section of the technical specifications is 
administrative and has no impact on probability or consequences of 
an accident.
    The change to move TS 3.4.9 to the Waterford 3 TRM will have no 
adverse effect on plant operation or the availability or operation 
of any accident mitigation equipment. Changes to the TRM are 
controlled in accordance with 10 CFR 50.59. Therefore, moving TS 
3.4.9 to the Waterford 3 TRM will not adversely impact [as] an 
accident initiator and can not cause an accident.
    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 will not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. They do not alter the way any 
structure, system, or component functions and do not alter the 
manner in which the plant is operated. These changes do not 
introduce any new failure modes.
    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 testing and inspection requirements contained in TS 4.0.5 
are governed by 10 CFR 50.55a, ``Codes and Standards.'' The 10 CFR 
requirements to perform the ASME code testing and inspections will 
not be reduced by the proposed change. The inspections and tests 
will continue to be performed as they are currently. The proposed 
change has no impact on plant equipment operation.
    The fracture mechanics analysis conducted in support of 
extending the RCP flywheel volumetric examination interval from 
three years to ten years shows that significant conservatism has 
been used for calculating the allowable flaw size, critical flaw 
size, and crack growth rate in the RCP flywheels. These include 
minimum material properties, maximum flywheel accident speed, 
location of the flaw in the highest stress area, and a number of 
startup/shutdown cycles eight times greater than expected. Since a 
postulated flaw in a Waterford 3 flywheel will not grow to the 
allowable flaw size under normal operating conditions or to the 
critical flaw size under loss of coolant accident conditions over 
the life of the plant, reducing the examination requirements for the 
detection of such cracks over the life of the plant will not involve 
a significant reduction in the margin of safety. The proposed change 
has no impact on plant equipment operation.
    The change to move the surveillance requirements for the RCP 
flywheels to the programs section of the technical specifications is 
administrative and has no impact on plant operation.
    Relocation of TS 3.4.9 to the TRM does not imply any reduction 
in its importance in ensuring that the structural integrity and 
operational readiness of ASME Code Class 1, 2, and 3 components will 
be maintained at an acceptable level throughout the life of the 
plant. The proposed change has no impact on plant operation.
    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: N.S. Reynolds, Esquire, Winston & Strawn, 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: March 31, 2003.
    Description of amendment request: The proposed amendments would 
revise Appendix A, Technical Specifications (TS), of Facility Operating 
License Nos. NPF-11 and NPF-18. Specifically, the proposed change will 
modify TS 5.7, ``High Radiation Area,'' by incorporating the wording 
and requirements from NUREG-1434, ``Standard Technical Specifications 
General Electric Plants, BWR/6,'' Revision 2, dated June 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change will modify LaSalle County Station (LSCS) TS 
5.7, ``High Radiation Area,'' by incorporating into the TS the 
corresponding wording and requirements from NUREG-1434, ``Standard 
Technical Specifications General Electric Plants, BWR/6,'' Revision 
2, dated June 2001. TS 5.7 establishes the administrative controls 
on entry into high radiation areas. High radiation area 
administrative controls are not a precursor to accidents previously 
evaluated. Thus, the proposed change does not have any effect on the 
probability of an accident previously evaluated.
    The proposed change in administrative controls on entry into a 
radiation area does not affect the ability of LSCS to successfully 
respond to previously evaluated accidents and does not affect 
radiological assumptions used in the evaluations. Thus, the 
radiological consequences of any accident previously evaluated are 
not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change does not affect the control parameters 
governing unit operation or the response of plant equipment to 
transient conditions. The proposed change does not introduce any new 
equipment, modes of system operation or failure mechanisms.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change incorporates corresponding wording and 
requirements from NUREG-1434, into the LSCS TS. The LSCS evaluation 
of the proposed change concluded the following:
    [sbull] Both the proposed and current TS 5.7.1 describe the 
requirements for access into areas that have radiation levels that 
exceed 100 mrem/hr but are less than or equal to 1000 mrem/hr. The 
proposed and current TS 5.7.1 are considered to have equivalent 
level access controls as both contain the need to provide a 
barricade, conspicuously post the area and issue an RWP to control 
entrance to the area.
    [sbull] Proposed TS 5.7.2 and current TS 5.7.4 describe the 
requirements for access into areas that have radiation levels that 
exceed 1000 mrem/hr. Proposed TS 5.7.2 and current TS 5.7.4 are 
considered to have equivalent level access controls as both require 
these areas to be locked. For those areas where locking is not 
practical, proposed TS 5.7.2 and current TS 5.7.4 both require the 
area to be barricaded, conspicuously posted, and have an activated 
flashing light.
    [sbull] The proposed change includes the deletion of the use of 
computer controlled doors in current TS 5.7.2. This description is 
being removed as computer controlled doors are no longer utilized at 
LSCS. Rather, manual locking mechanisms are used on doors providing 
an equivalent level of control.
    [sbull] Current TS 5.7.4 also discusses ``high-high'' radiation 
areas. The term ``high-high'' radiation area is a legacy term that 
is being deleted from the proposed TS. This is an administrative 
change only to remove an outdated term.
    Therefore, LSCS has determined that the proposed change provides 
an equivalent level of protection as that currently provided.

[[Page 28853]]

    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Based upon the above, Exelon Generation Company concludes that 
the proposed amendment presents 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 staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), 
Beaver County, Pennsylvania

    Date of amendment request: March 11, 2003.
    Description of amendment request: The proposed amendment revises 
the BVPS-1 and 2 Technical Specifications (TSs) to apply the 
Westinghouse best-estimate large break loss-of-coolant accident (LOCA) 
methodology to BVPS-1 and 2. The request is contingent upon Nuclear 
Regulatory Commission (NRC) approval of the licensee's amendment 
request for conversion of the BVPS-1 and 2 containments from sub-
atmospheric to atmospheric which had previously been requested by 
letter dated June 5, 2002, and which is currently under NRC staff 
review.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. No physical changes are required as a result of implementing 
best-estimate loss of cooling accident (LOCA) methodology and 
associated technical specification changes. The plant conditions 
assumed in the analysis are bounded by the design conditions for all 
equipment in the plant. Therefore, there will be no increase in the 
probability of a loss of cooling accident. The consequences of a LOCA 
are not being increased, since it is shown that the emergency core 
cooling system is designed so that its calculated cooling performance 
conforms to the criteria contained in 10 CFR 50.46, Paragraph b. No 
other accident is potentially affected by this change.
    Therefore, the proposed change does 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 previously evaluated?
    No. There are no physical changes being made to the plants. No new 
modes of plant operation are being introduced. The parameters assumed 
in the analysis are within the design limits of the existing plant 
equipment. All plant systems will perform as designed during the 
response to a potential accident.
    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?
    No. It has been shown that the methodology used in the analysis 
would more realistically describe the expected behavior of plant 
systems during a postulated loss of coolant accident. Uncertainties 
have been accounted for as required by 10 CFR 50.46. A sufficient 
number of loss of coolant accidents with different break sizes, 
different locations and other variations in properties are analyzed to 
provide assurance that the most severe postulated loss of coolant 
accidents are calculated. It has been shown by analysis that there is a 
high level of probability that all criteria contained in 10 CFR 50.46, 
Paragraph b are met.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: March 27, 2003.
    Description of amendment request: The proposed amendment would 
amend Unit 2 Technical Specification (TS) Table 3.3-4 and the P-11 
setpoint in the Engineered Safety Features Interlock Table as follows:
    1. Revise the low pressurizer pressure safety injection (SI) trip 
setpoint from its current value of greater than or equal to 1900 pounds 
per square inch gauge (psig), to greater than or equal to 1815 psig.
    2. Revise the low pressurizer pressure SI allowable value from 
greater than or equal to 1890 psig, to greater than or equal to 1805 
psig.
    3. Revise the P-11 setpoint from its current value of greater than 
or equal to 2010 psig, to greater than or equal to 1915 psig.
    4. Make format changes to the affected TS pages that improve 
appearance but do not affect any requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or the consequences of an accident previously 
evaluated?
    Response: No.
    [Indiana Michigan Power Company] (I&M) proposes changing the low 
pressurizer pressure SI trip setpoint, the low pressurizer pressure 
SI allowable value, the P-11 setpoint, and the format of the 
associated pages. Neither the change to the low pressurizer pressure 
SI trip setpoint value and the SI allowable value nor the change to 
the P-11 setpoint value alter any safety-related components or the 
means of accomplishing a safety-related function. The change in the 
values is supported by analyses that demonstrate that applicable 
acceptance criteria are met when SI is initiated at 1700 psig for a 
(loss-of-coolant accident) LOCA, a main steam system 
depressurization event, and a feedwater line break. Because the 
acceptance criteria are met, there is no significant increase in the 
consequences of an accident. The format changes are intended to 
improve readability and appearance, and do not alter any 
requirements. Thus, neither the probability of an accident nor the 
consequences of an accident are significantly increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.

[[Page 28854]]

    I&M proposes changing the low pressurizer pressure SI trip 
setpoint, the low pressurizer pressure SI allowable value, the P-11 
setpoint, and the format of the associated pages. Neither the change 
to the low pressurizer pressure SI trip setpoint value and the SI 
allowable value nor the change to the P-11 setpoint value involve 
changing the design function of any component, and a change in any 
of the values cannot initiate an accident. The format changes are 
intended to improve readability and appearance, and do not alter any 
requirements. Thus, no new accident initiators are introduced, and 
the possibility of a new or different kind of accident is not 
created.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    I&M proposes changing the low pressurizer pressure SI trip 
setpoint, the low pressurizer pressure SI allowable value, the P-11 
setpoint, and the format of the associate pages. The low pressurizer 
pressure instrument is credited for activating the engineered safety 
features in the event of a LOCA, a main steam system 
depressurization event, or a feedwater line break. The low 
pressurizer pressure SI trip setpoint value and the low pressurizer 
pressure SI allowable value have been selected to insure that the 
engineered safety features will be activated as assumed in the 
safety analysis. Present margins continue to be maintained because 
the applicable accident analyses criteria continue to be met. No 
margins of safety are associated with the P-11 setpoint value. The 
format changes are intended to improve readability and appearance, 
and do not alter any requirements. Thus, there is no 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 requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: April 24, 2003.
    Description of amendment request: Eliminate the requirement for 
continuous Control Room manning when fuel is stored in the fuel storage 
pool.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated? 
Involve a significant increase in the probability or consequence of 
an accident previously evaluated.
    Response: No.
    The Defueled Safety Analysis (DSAR) identifies five categories 
of events: spent fuel criticality accidents, a fuel handling 
accident, a spent fuel cask drop, spent fuel pool accidents, and a 
low level waste release incident. There are no active controls in 
the control room that are required to respond to these events. 
Actions to mitigate the consequences of these events are taken 
outside the control room. Emergency response is not adversely 
affected by this proposed change because the control room is still 
available to the emergency response team and communications 
capability and timeliness will not be affected. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The configuration, operations and accident response of the 
systems, structures or components that support safe storage of the 
spent fuel are unchanged by the proposed change to the technical 
specifications. Current site surveillance requirements ensure 
frequent and adequate monitoring of system and component 
functionality. Systems in the Spent Fuel Pool Island will continue 
to be operated in accordance with current design requirements and no 
new components or system interactions have been identified. No new 
accident scenarios, failure mechanisms or limiting single failures 
are introduced as a result of the proposed change. The proposed 
technical specification change does not have an adverse affect on 
any system related to safe storage of spent fuel. Therefore, the 
proposed technical specifications change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    All design basis accident acceptance criteria will continue to 
be met. The margin of safety relative to the cooling of the spent 
fuel is unaffected by the proposed changes as the spent fuel pool 
parameters will continue to be monitored at the same frequency as 
assumed in the accident analysis. The ability of the shift crew to 
respond to abnormal or accident conditions is unaffected by the 
proposed change since all controls are located in or near the fuel 
building and any necessary communications will be handled by the on-
shift staff and/or DERO. Therefore, it is concluded that the 
proposed TS change does not involve a significant reduction in the 
margin of safety
    Based on the above, Maine Yankee concludes that the proposed 
amendment presents 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 staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendment involves no significant hazards consideration.
    Attorney for licensee: Joe Fay, Esquire, Maine Yankee Atomic Power 
Company, 321 Old Ferry Road, Wiscasset, Maine 04578.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: October 8, 2002.
    Description of amendment request: The license amendment request 
proposes to change the title of (a) Shift Supervisor to Shift Manager, 
(b) ``Plant Manager'' to ``plant manager,'' (c) ``Vice President--
Nuclear'' to ``corporate officer with direct responsibility for the 
plant,'' (d) ``Radiological Manager'' to ``radiological manager,'' (e) 
``Operations Supervisor'' to ``operations supervisor'' and (f) ``Shift 
Radiological Protection/Chemistry Technician'' to ``radiation 
protection technician.'' This proposal includes an Updated Safety 
Analysis Report (USAR) reference correction resulting from the USAR 
Rebaseline Project and a correction to the title ``Shift Technical 
Advisor'' to ``Shift Technical Engineer'' in Technical Specification 
(TS) Section 5.3.1 so as to be consistent with the title used in TS 
Section 5.2.2.f. These changes do not eliminate any of the 
qualifications, responsibilities, or requirements for these positions.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The title of Shift Manager better conveys the appropriate level 
of responsibility and authority required of the position. The use of 
generic personnel titles and correction of the USAR reference are 
strictly administrative. The qualifications, training, duties and 
experience required of the individuals remain unchanged. The USAR 
section to be referenced is physically the same section that was 
referenced before the USAR renumbering. The requested changes do not

[[Page 28855]]

involve any change to the design basis of the plant or any 
structure, system, or component. Therefore, these changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    There will be no physical alterations to the plant 
configuration. No changes in operating mode or limits are proposed. 
The qualifications, training, duties and experience required of the 
individuals remain unchanged. The USAR section to be referenced is 
physically the same section that was referenced before the USAR 
renumbering. Therefore, these proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    The proposed change in titles and USAR reference are strictly 
administrative. The qualifications, training, duties and experience 
required of the individuals remain unchanged. The USAR section to be 
referenced is physically the same section that was referenced before 
the USAR renumbering. The proposed changes do not change any license 
condition or Technical Specifications safety limit or limiting 
condition for operation. The changes do not involve modification of 
the design or operation of any plant system involved with 
controlling the release of radioactivity to the environment. 
Therefore, these changes do not involve a significant reduction in a 
margin of safety.
    Based on the above, Nebraska Public Power District concludes 
that the proposed amendment presents 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 staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of amendment request: May 2, 2003.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) by replacing the existing 
Reactor Coolant System (RCS) pressure and temperature (P/T) limit 
curves for in-service leakage and hydrostatic testing, non-nuclear 
heatup and cooldown, and criticality (Figure 3.4.9-1, ``Pressure Versus 
Minimum Temperature Valid to Thirty-two Full Power Years, per Appendix 
G of 10 CFR 50'') with new, updated P/T limits curves. The replacement 
curves were generated using an NRC-approved methodology (General 
Electric Report NEDC-32983P) for determining the neutron fluence on the 
Reactor Pressure Vessel (RPV) and extends the RPV beltline region to 
encompass a new limiting component, the recirculation inlet nozzle. The 
change to Figure 3.4.9-1 would also delete the existing notation that 
states: ``(Interim Approval Until September 1, 2003).''
    The licensee's application for amendment dated May 2, 2003, 
supersedes and withdraws a previous application dated February 28, 
2003, for which the NRC has published a notice of consideration of 
issuance of amendment, proposed no significant hazards consideration 
determination, and opportunity for hearing in the Federal Register (68 
FR 12954, dated March 18, 2003).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The P/T limits are not derived from Design Basis Accident (DBA) 
analyses. They are prescribed by the (American Society of Mechanical 
Engineers Boiler and Pressure Vessel Code) ASME Code and 10 CFR 50 
Appendix G and H and associated guidance documents, such as 
Regulatory Guide 1.99, Rev. 2, as restrictions on normal operation 
to avoid encountering pressure, temperature, and temperature rate of 
change conditions that might cause undetected flaws to propagate and 
cause non-ductile failure of the reactor coolant pressure boundary. 
Thus, they ensure that an accident precursor is not likely. Hence, 
they are included in the TS as satisfying Criterion 2 of 10 CFR 
50.36(c)(2)(ii). The revision of the numerical value of these 
limits, i.e., new curves, using an NRC-approved methodology, does 
not change the existing regulatory requirements, upon which the 
curves are based. Thus, this revision will not increase the 
probability of any accident previously evaluated.
    The proposed change does not alter the design assumptions, 
conditions, or configuration of the facility or the manner in which 
the facility is operated or maintained. The proposed changes will 
not affect any other System, Structure or Component (SSC) designed 
for the mitigation of previously analyzed events. The proposed 
change does not affect the source term, containment isolation, or 
radiological release assumptions used in evaluating the radiological 
consequences of any accident previously evaluated. Thus, the 
proposed revision of the existing numerical values with the updated 
figure for the RCS P/T limits, which are based upon an NRC-approved 
methodology for calculating the neutron fluence on the RPV and new 
limiting component, will not increase the consequences of any 
previously evaluated accident.
    (2) The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the processes governing normal plant 
operation. The proposed changes are consistent with the safety 
analysis assumptions and current plant operating practice. (Nuclear 
Management Company, LLC) NMC is only requesting to revise the 
existing numerical values and update the TS figure for the RCS P/T 
limits based upon an NRC-approved methodology for calculating the 
neutron fluence on the RPV, and to reflect a new limiting component. 
The curves continue to be based upon ASME Code Case N-640, which has 
been previously approved for use at the [Duane Arnold Energy Center] 
DAEC.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    (3) The proposed amendment will not involve a significant 
reduction in a margin of safety.
    The proposed changes do not alter the manner in which Safety 
Limits, Limiting Safety System Settings or Limiting Conditions for 
Operation are determined. The setpoints at which protective actions 
are initiated are not altered by the proposed changes. Sufficient 
equipment remains available to actuate upon demand for the purpose 
of mitigating an analyzed event. NMC is only requesting to revise 
the existing numerical values and update the TS figure for the RCS 
P/T limits based upon an NRC-approved methodology for calculating 
the neutron fluence, NEDC-32983P-A. The new curves also reflect the 
addition of a new limiting component, the recirculation inlet nozzle 
(N2). No other changes to the Limiting Conditions for Operation or 
any Surveillance Requirements of Technical Specification 3.4.9 are 
proposed.
    10 CFR 50, Appendix G specifies fracture toughness requirements 
to provide adequate margins of safety during operation over the 
service lifetime. The values of adjusted reference temperature and 
upper shelf energy are expected to remain within the limits of 
Regulatory Guide 1.99, Revision 2 and Appendix G of 10 CFR 50 for at 
least 32 effective full power years (EFPY) of operation. The safety 
analysis supporting this change continues to satisfy the ASME Code, 
including ASME Code Case N-640, and 10 CFR 50, Appendices G and H 
requirements and associated guidance documents, such as Regulatory 
Guide 1.99, Rev. 2. Thus, the

[[Page 28856]]

proposed changes will not significantly reduce any margin of safety 
that currently exists.
    Based upon the above, NMC has determined that the proposed 
amendment will not involve a significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based upon 
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: Jonathan Rogoff, Esquire, General Counsel, 
NMC, LLC, 700 First St., Hudson, WI 54016.
    NRC Section Chief: L. Raghavan.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: March 13, 2003.
    Description of amendment request: The proposed amendment deletes 
requirements from the Technical Specifications (TSs) and other elements 
of the licensing bases to maintain a Post Accident Sampling System 
(PASS). Licensees were generally required to implement PASS upgrades as 
described in NUREG-0737, ``Clarification of TMI [Three Mile Island] 
Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
PASS were imposed by Order for many facilities and were added to, or 
included in, the TSs for nuclear power reactors currently licensed to 
operate. Lessons learned and improvements implemented over the last 20 
years have shown that the information obtained from PASS can be readily 
obtained through other means, or is of little use in the assessment and 
mitigation of accident conditions.
    The changes are based on NRC-approved Technical Specification Task 
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
413, ``Elimination of Requirements for a Post Accident Sampling System 
(PASS).'' The NRC staff issued a notice of opportunity for comment in 
the Federal Register on December 27, 2001 (66 FR 66949), on possible 
amendments concerning TSTF-413, including a model Safety Evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on March 20, 2002 (67 FR 13027). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated March 13, 2003.
    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. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post-accident situations and were put into place as a result of 
the TMI-2 accident. The specific intent of the PASS was to provide a 
system that has the capability to obtain and analyze samples of plant 
fluids containing potentially high levels of radioactivity, without 
exceeding plant personnel radiation exposure limits. Analytical results 
of these samples would be used largely for verification purposes in 
aiding the plant staff in assessing the extent of core damage and 
subsequent offsite radiological dose projections. The system was not 
intended to, and does not, serve a function for preventing accidents 
and its elimination would not affect the probability of accidents 
previously evaluated.
    In the 20 years since the TMI-2 accident, and the consequential 
promulgation of post-accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual benefit 
to post-accident mitigation. Past experience has indicated that there 
exists in-plant instrumentation and methodologies available in lieu of 
a PASS for collecting and assimilating information needed to assess 
core damage following an accident. Furthermore, the implementation of 
Severe Accident Management Guidance (SAMG) emphasizes accident 
management strategies based on in-plant instruments. These strategies 
provide guidance to the plant staff for mitigation and recovery from a 
severe accident. Based on current severe accident management strategies 
and guidelines, it is determined that the PASS provides little benefit 
to the plant staff in coping with an accident.
    The regulatory requirements for the PASS can be eliminated without 
degrading the plant emergency response. The emergency response, in this 
sense, refers to the methodologies used in ascertaining the condition 
of the reactor core, mitigating the consequences of an accident, 
assessing and projecting offsite releases of radioactivity, and 
establishing protective action recommendations to be communicated to 
offsite authorities. The elimination of the PASS will not prevent an 
accident management strategy that meets the initial intent of the post-
TMI-2 accident guidance through the use of the SAMGs, the emergency 
plan (EP), the emergency operating procedures (EOP), and site survey 
monitoring that support modification of emergency plan protective 
action recommendations (PARs).
    Therefore, the elimination of the PASS requirements from TSs (and 
other elements of the licensing bases) does not involve a significant 
increase in the consequences of any accident previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The elimination of PASS-related requirements will not result in any 
failure mode not previously analyzed. The PASS was intended to allow 
for verification of the extent of reactor core damage, and also to 
provide an input to offsite dose projection calculations. The PASS is 
not considered an accident precursor, nor does its existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post-accident confinement of radioisotopes within the 
containment building.
    Therefore, this change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction in 
the margin of safety.
    The elimination of the PASS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of, and recovery from, reactor accidents, results in a 
neutral impact to the margin of safety. Methodologies that are not 
reliant on PASS are designed to provide rapid assessment of current 
reactor core conditions and the direction of degradation while 
effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and does 
not provide quick recognition of core events or rapid response to 
events in progress. The intent of the requirements established as

[[Page 28857]]

a result of the TMI-2 accident can be adequately met without reliance 
on a PASS.
    Therefore, this change does not involve a significant reduction in 
the margin of safety.
    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: March 21, 2003.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications (TS) Section 5.5.1, ``Offsite Dose 
Calculation Manual (ODCM),'' to remove reference to the Plant 
Operations Review Committee review and acceptance of licensee initiated 
changes to the ODCM.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change for TS section 5.5.1.b removes the reference 
to the Plant Operations Review Committee review and acceptance of 
licensee initiated changes to the ODCM. This change is an 
administrative change and does not change plant design or responses.
    The proposed change does not involve changing any structure, 
system, or component, or affect reactor operations. It is not an 
initiator of an accident and does not change any existing safety 
analysis previously analyzed in the UFSAR. As such, the proposed 
change does not involve a significant increase in the probability of 
an accident previously evaluated. Since the proposed change does not 
alter the plant design, it does not involve a significant increase 
in the 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?
    The proposed change for TS section 5.5.1.b removes the reference 
to the Plant Operations Review Committee review and acceptance of 
licensee initiated changes to the ODCM. This change is an 
administrative change and does not change plant design or responses.
    The proposed change will not alter any plant design basis or 
postulated accident. In addition, the proposed change does not 
impact any plant systems or components.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed change for TS section 5.5.1.b removes the reference 
to the Plant Operations Review Committee review and acceptance of 
licensee initiated changes to the ODCM. This change is an 
administrative change and does not change plant design or responses. 
The proposed change does not impact accident offsite dose, 
containment pressure or temperature, emergency core cooling system 
setpoints, reactor protection system settings or any other parameter 
that could affect a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: April 11, 2003 (TS-424).
    Description of amendment request: The proposed amendments would 
reduce the number of Emergency Core Cooling System subsystems that are 
available in response to certain design basis loss-of-coolant accident 
(LOCA) scenarios because of TVA's planned restart of Unit 1. The 
licensee stated that the reduced number has been analyzed and is 
consistent with the current approved LOCA analysis methodology. The 
amendments are needed to eliminate the potential for overloading a 
shutdown board or a diesel generator when both Units 1 and 2 are in-
service. The reduction requires a change to the Updated Final Safety 
Analysis Report.
    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. Do the proposed amendments and Technical Specification 
changes involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. The proposed amendments revise the actual number of 
Emergency Core Cooling System (ECCS) subsystems that are available 
in response to certain design basis Loss of Coolant Accident (LOCA) 
scenarios. The associated modifications also result in a revision to 
the number of required channels for the Low Pressure Coolant 
Injection (LPCI) pump start time delay relay function specified in 
Technical Specifications. The proposed amendments and Technical 
Specification changes do not affect any accident precursors. 
Therefore, the probability of an evaluated accident is not 
increased.
    The reduction in the number of ECCS subsystems that are actually 
available in response to the bounding LOCA case (A recirculation 
suction line break with an assumed battery failure) will now be the 
same as the number of ECCS subsystems evaluated in the current BFN 
SAFER/GESTR-LOCA analysis. The ECCS performance for the bounding 
LOCA case has previously been evaluated using the approved SAFER/
GESTR-LOCA application methodology and is described in Updated Final 
Safety Analysis Report (UFSAR) Sections 6.5 and 14.6.3. The revision 
to the number of required channels for the LPCI pump start time 
delay relay function does not affect the LOCA analysis. The 
requirements of 10 CFR 50.46 and Appendix K are met. Therefore, the 
proposed amendments and Technical Specification changes will not 
significantly increase the consequences of an accident previously 
evaluated.
    2. Do the proposed amendments and Technical Specification 
changes create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    No. The proposed amendments revise the number of ECCS subsystems 
that are actually available in response to certain design basis LOCA 
scenarios. The proposed Technical Specification changes revise the 
number of required channels for the LPCI pump start time delay relay 
function. The proposed amendments and Technical Specification 
changes do not introduce new equipment, which could create a new or 
different kind of accident.
    No new external threats, release pathways, or equipment failure 
modes are created. Therefore, the implementation of the proposed 
amendments and Technical Specification changes will not create a 
possibility for an accident of a new or different type than those 
previously evaluated.
    3. Do the proposed amendments and Technical Specification 
changes involve a significant reduction in a margin of safety?
    No. The proposed amendments and Technical Specification changes 
revise the number of ECCS subsystems that are actually available in 
response to certain design basis LOCA scenarios. The reduction in 
the number of ECCS subsystems that are actually available in 
response to the bounding LOCA case (A recirculation suction line 
break with an assumed battery failure) will now be the same as the 
number of ECCS subsystems evaluated in the current BFN SAFER/GESTR-
LOCA analysis. The ECCS performance for the bounding LOCA case has 
previously been evaluated using the approved SAFER/GESTR-LOCA 
application methodology. The revision to the number of required 
channels for the LPCI pump start time delay relay function does not 
affect the LOCA analysis. The requirements of 10 CFR 50.46 and 
Appendix K are met. Therefore,

[[Page 28858]]

the proposed license amendments and Technical Specification changes 
do not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: April 15, 2003 (TS 409).
    Description of amendment request: The proposed amendments are 
applicable to BFN Units 1, 2, and 3. They would revise Technical 
Specification (TS) Limiting Condition for Operation 3.7.3, Control Room 
Emergency Ventilation (CREV) System, and its associated TS Bases to 
provide specific conditions and required actions that address a 
degraded main control room boundary. The proposed changes are 
consistent with the TS Task Force Traveler 287, Revision 5.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    2. No. The proposed TS change involves the CREV system, which 
provides a radiological controlled environment from which the plant 
can be operated following a design basis accident (DBA). The CREV 
system is not assumed to be the initiator of any analyzed accident 
and cannot not [sic] affect the probability of accidents.
    The proposed change allows the main control room boundary to be 
opened intermittently under administrative control, and allows 24 
hours to restore the main control room boundary to Operable status 
before requiring the plant to perform an orderly shutdown. The 24-
hour Completion Time is reasonable based on the low probability of a 
DBA occurring during this time period and TVA's commitment to 
implement, via administrative controls, appropriate compensatory 
measures consistent with the intent of 10 CFR part 50, Appendix A, 
General Design Criteria (GDC) 19. These compensatory measures 
minimize the consequences of an open main control room boundary and 
assure that CREV system can continue to perform its function. As 
such, these changes will not affect the function or operation of any 
other systems, structures, or components.
    Therefore, the proposed TS change does not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed change allows the main control room boundary to 
be opened intermittently under administrative control and allows 24 
hours to restore the main control room boundary to Operable status 
before requiring the plant to perform an orderly shutdown. The 24-
hour Completion Time is reasonable based on the low probability of a 
DBA occurring during this time period and TVA's commitment to 
implement, via administrative controls, appropriate compensatory 
measures consistent with the intent of 10 CFR part 50, Appendix A, 
GDC 19. These compensatory measures minimize the consequences of an 
open main control room boundary and assure that the CREV system can 
continue to perform its function. As such, these changes will not 
affect the function or operation of any other systems, structures, 
or components.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed change allows the main control room boundary to 
be opened intermittently under administrative control and allows 24 
hours to restore the main control room boundary to Operable status 
before requiring the plant to perform an orderly shutdown. The 24-
hour Completion Time is reasonable based on the low probability of a 
DBA occurring during this time period and TVA's commitment to 
implement, via administrative controls, appropriate compensatory 
measures consistent with the intent of 10 CFR part 50, Appendix A, 
GDC 19. These compensatory measures minimize the consequences of an 
open main control room boundary and assure that the CREV system can 
continue to perform its function such that compliance with GDC 19 is 
maintained.
    Therefore, the proposed TS change does not involve a reduction 
in the margin of safety.
    Based on the above, TVA concludes that the proposed amendment 
presents 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 staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: April 14, 2003 (TS 425).
    Description of amendment request: The proposed amendments would 
revise two Technical Specification (TS) Limiting Conditions for 
Operation 3.3.4.1, ``End Of Cycle Recirculation Pump Trip (EOC-RPT) 
Instrumentation,'' and 3.7.5, ``Main Turbine Bypass System,'' to 
reference additional core limits adjustment factors for linear heat 
generation rate for equipment out-of-service conditions. Also, Section 
b of TS 5.6.5, ``Core Operating Limits Report (COLR),'' would be 
revised to add references to the Framatome Advanced Nuclear Power 
(FANP) analytical methods that will be used in the upcoming fuel cycles 
to determine core operating limits. The above TS changes are needed to 
support a transition to the use of FANP fuel, and FANP core design and 
analysis services.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. Core operating limits are established to support 
requirements, which in turn ensure that fuel design limits are not 
exceeded during any conditions of operating transients or accidents. 
The methods used to determine the limits for each operating cycle 
are based on methods previously found acceptable by the NRC and are 
required to be listed in COLR TS Section 5.6.5.b. Accordingly, a 
change to TS Section 5.6.5.b is requested to include FANP methods in 
the list of NRC-approved methods applicable to BFN. This TS change 
also adds provisions that ensure core thermal limits adjustment 
factors are applied for equipment out-of-service conditions 
associated with the use of FANP methods for transient analyses. The 
application of these NRC-approved methods will continue to ensure 
that acceptable operating limits are established and applied for 
protection of fuel cladding integrity during transient and 
accidents.
    The requested TS changes do not involve any plant modifications 
or operational changes that could affect system reliability,

[[Page 28859]]

performance, or possibility of operator error. The requested changes 
do not affect any postulated accident precursors, do not affect any 
accident mitigation systems, and do not introduce any new accident 
initiation mechanisms.
    Therefore, the proposed TS change does not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The core operating limits and required limits adjustments 
for equipment out-of-service conditions will continue to be 
determined using methodologies that have been approved by the NRC. 
The limits derived from approved methodologies will provide adequate 
margins of safety. The proposed changes do not involve any new modes 
of operation, any changes to setpoints, or any plant modifications, 
and do not result in any new precursors to an accident.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety.
    No. The core operating limits and required limits adjustments 
for equipment out-of-service will continue to be determined using 
methodologies that have been approved by the NRC. On this basis, the 
implementation of the changes does not involve a significant 
reduction in margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: May 1, 2003.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.8.7, ``Inverters--Operating.'' 
The TS as currently written requires two inverters for each of the four 
instrument channels. The revision changes the requirement to one 
inverter for each of the four channels. The amendment is the initial 
phase of a project that will replace the vital inverters to achieve 
improvements in the reliability of the 120V AC Vital Instrument Power 
System.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed revisions to WBN's Vital AC Power System do not 
alter the safety functions of the Vital Inverters or the Unit 1 and 
Unit 2 120V AC Vital Instrument Power Boards. The initial conditions 
for the Design Basis Accidents (DBAs) defined in the WBN Updated 
Final Safety Analysis Report (UFSAR) assume the Engineered Safety 
Feature (ESF) systems are operable. The vital inverters are designed 
to provide the required capacity, capability, redundancy, and 
reliability to ensure the availability of necessary power to vital 
instrumentation so that the fuel, reactor coolant system, and 
containment design limits are not exceeded. Adding the Unit 2 loads 
to the Unit 1 inverters does not alter the accident analyses as long 
as the Unit 1 inverters are capable of handling the additional loads 
and channel separation is maintained. Design calculations document 
that the Unit 1 inverters have adequate capacity to support the 
addition of the Unit 2 loads and no changes are proposed that will 
impact the separation of the Vital AC Power System. In addition, the 
redundant capabilities of the Vital AC System as currently described 
in the UFSAR are not impacted by the proposed amendment.
    The inverters and the associated 120V AC Vital Instrument Power 
Boards are utilized to support instrumentation that monitor critical 
plant parameters to aid in the detection of accidents and to support 
the mitigation of accidents, but are not considered to be an 
initiator of design basis accidents. Based on this and the preceding 
information, the proposed amendment does 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?
    No. When implemented, the proposed TS amendment will allow the 
Unit 2 Vital Instrument Power Boards to receive their 
uninterruptible power supply (UPS) power from the Unit 1 inverters 
instead of the Unit 2 inverters. Calculations have verified that the 
additional load will not affect the ability of the Unit 1 inverters 
to perform their intended safety functions. In addition, the 
inverters and the 120V AC Vital Instrument Power Boards are not 
considered to be an initiator of a design basis accident. These 
components provide power to instrumentation that supports the 
identification and mitigation of accidents as well as system control 
functions during normal plant operations. The functions of the 
inverters are not altered by the proposed TS change and will not 
create the possibility of a new or different accident. Further, the 
addition of the Unit 2 loads to the Unit 1 inverters is the 
principal change to the inverter system and this change is bounded 
by previously evaluated accident analyses. Therefore, the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The plant setpoints and limits that are utilized to ensure 
safe operation and detect accident conditions are not impacted by 
the proposed TS amendment. The inverters and the 120V Vital 
Instrument Power Boards will continue to provide reliable power to 
safety-related instrumentation for the identification and mitigation 
of accidents and to support plant operation. Therefore, the margin 
of safety is not reduced.
    Based on the above, TVA concludes that the proposed amendment 
presents 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.
    In conclusion, based on the considerations discussed above, (1) 
There is reasonable assurance that the health and safety of the 
public will not be endangered by operation in the proposed manner, 
(2) such activities will be conducted in compliance with the 
Commission's regulations, and (3) the issuance of the amendment will 
not be inimical to the common defense and security or to the health 
and safety of the public.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) 
are satisfied. Therefore, the NRC staff proposes to determine that 
the amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley 
Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 
37902.
    NRC Section Chief: Allen G. Howe.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these

[[Page 28860]]

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, Public File Area 01F21, 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 email to [email protected].

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

    Date of application for amendment: August 1, 2002, as supplemented 
on October 18, 2002, and April 17, 2003.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.7.1.1, ``Plant Systems: Turbine Cycle Safety 
Valves,'' to reflect results of a reanalysis of overpressurization 
events to allow plant operation, at corresponding reduced power levels, 
with up to four main steam safety valves in each main steam line 
inoperable.
    Date of issuance: May 7, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 275.
    Facility Operating License No. DPR-65: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: September 17, 2002 (67 
FR 58638). The supplements dated October 18, 2002, and April 17, 2003, 
provided additional information which 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 May 7, 2003.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: September 30, 2002, as 
supplemented by letters dated October 17, 2002 and April 2, 2003.
    Brief description of amendments: The amendments revise the 
Technical Specification to: (1) Modify the Surveillance Requirement to 
be consistent with the design of the reactor building access openings, 
(2) modify the frequency of the Surveillance Requirement for visual 
inspections for the exposed interior and exterior surface of the 
reactor building, and (3) modify the administrative controls for the 
containment leakage rate testing program.
    Date of issuance: May 8, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 212/193.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68733). The supplement dated October 17, 2002, and April 12, 2003, 
provided clarifying information that did not change the scope of the 
September 30, 2002, 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 May 8, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 3, 2002, as 
supplemented by letters dated November 27, 2002, and April 17, 2003.
    Brief description of amendment: The amendment allows the addition 
of depleted uranium to the fuel assembly composition described in 
Technical Specification (TS) 4.2.1. The amendment also revises TS 
5.6.5.b to incorporate the references to the analytical methods to be 
used to determine core operating limits and removes those references 
that will no longer be used. The amendment also allows the format for 
those document references to be revised as described in the staff-
approved Industry/TSTF Standard Technical Specification Change 
Traveler, TSTF-363, ``Revise Topical Report References in ITS 5.6.5, 
COLR.''
    Date of issuance: May 12, 2003.
    Effective date: May 12, 2003, and shall be implemented within 30 
days from the date of issuance.
    Amendment No.: 185.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 15, 2002 (67 FR 
63693). The November 27, 2002, and April 17, 2003, supplemental letters 
provided additional clarifying information, did not change 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 May 12, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 26, 2002, as supplemented 
on March 12, 2003.
    Brief description of amendment: The amendment revises Technical 
Specification 5.6.5.b, ``Core Operating Limits Report (COLR),'' to 
incorporate the reference to Westinghouse Topical Report WCAP-12945-P-
A, ``Code Qualification Document for Best Estimate Loss-of-Coolant 
Analysis,'' dated March 1998. The amendment allows the use of the 
analytical methodology to determine the core operating limits.
    Date of issuance: May 6, 2003.
    Effective date: May 6, 2003.
    Amendment No.: 217.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 6, 2002 (68 FR 
50952). The March 12 letter provided clarifying information that did 
not expand the scope of the Federal Register notice or change the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is

[[Page 28861]]

contained in a Safety Evaluation dated May 6, 2003.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of application for amendment: July 5, 2002, as supplemented on 
September 27, November 6, November 21, and December 30, 2002; February 
4, February 10, March 17, and April 14, 2003.
    Brief description of amendment: The amendment increases the 
licensed power level by 1.5%, from 1998 MWt to 2028 MWt, based on the 
installation of ultrasonic flow measurement instrumentation resulting 
in improved feedwater flow measurement accuracy. The amendment changes 
the Operating License (OL) and Technical Specifications (TSs) to 
reflect the increase in licensed power level.
    Date of issuance: May 9, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 201.
    Facility Operating License No. DPR-35: Amendment revised the TSs 
and OL.
    Date of initial notice in Federal Register: September 3, 2002 (67 
FR 56322). The supplements dated September 27, November 6, November 21, 
and December 30, 2002; February 4, February 10, March 17, and April 14, 
2003, provided additional information that clarified the application, 
and did not expand the scope of the application or 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 May 9, 2003.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of application for amendments: February 27, 2003, as 
supplemented April 7, 2003.
    Brief description of amendments: The amendments revise the 
Technical Specifications by adding a surveillance requirement to 
perform a quarterly trip unit calibration of the reactor protection 
system scram discharge volume water level--high differential pressure 
switches.
    Date of issuance: May 6, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 214/208.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 1, 2003 (68 FR 
15760). The supplement dated April 7, 2003, 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 amendments is contained in a 
Safety Evaluation dated May 6, 2003.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County, 
Pennsylvania

    Date of application for amendments: January 16, 2002, as 
supplemented October 17, 2002.
    Brief description of amendments: These amendments revised portions 
of the current Technical Specifications, Section 6.0, ``Administrative 
Controls,'' to conform with improved Technical Specifications. The 
conversion is based upon: NUREG-1431, ``Standard Technical 
Specifications for Westinghouse Plants,'' Revision 2, dated April 2001; 
``Final Policy Statement on Technical Specification Improvements for 
Nuclear Power Reactors'' (Final Policy Statement), published on July 
22, 1993 (58 FR 39132); and Title 10 of the Code of Federal Regulations 
(10 CFR), Sec.  50.36, ``Technical Specifications,'' as amended July 
19, 1995.
    Date of issuance: May 15, 2003.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 255 and 136.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 29, 2002 (67 FR 
66010). The supplement dated October 17, 2002, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the Nuclear 
Regulatory Commission (NRC) staff's original proposed no significant 
hazards consideration determination as published in the Federal 
Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 15, 2003.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit 2, Beaver County, Pennsylvania

    Date of application for amendment: May 31, 2002, as supplemented 
September 11, 2002, January 30, and February 21, 2003.
    Brief description of amendment: The amendment revised the Technical 
Specification Design Feature 5.3.1, Criticality, such that the new fuel 
(fresh fuel) racks enrichment limit specified in Section 5.3.1.2.a was 
increased from 4.85 weight percent to a 5.00 weight percent limit.
    Date of issuance: May 15, 2003.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment No: 135.
    Facility Operating License No. NPF-73. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 17, 2002 (67 
FR 58645). The September 11, 2002, January 30, and February 21, 2003, 
letters provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 15, 2003.
    No significant hazards consideration comments received: No.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: March 22, 2002, as supplemented May 13, 
June 24, July 29, and December 20, 2002.
    Description of amendment request: The amendment revises Technical 
Specifications (TSs) Surveillance Requirement (SR) 4.0.3 to extend the 
delay period, before entering a Limiting Condition for Operation, 
following a missed surveillance. The delay period is extended from the 
current limit of ``... up to 24 hours'' to ``...up to 24 hours or up to 
the limit of the specified surveillance interval, whichever is 
greater.'' In addition, the following requirement is added to SR 4.0.3: 
``A risk evaluation shall be performed for any Surveillance delayed 
greater than 24 hours and the risk impact shall be managed.'' The 
amendment also adds a requirement for a TS Bases Control Program to the 
administrative controls section of TSs and makes administrative changes 
to SRs 4.0.1 and 4.0.3 to be consistent with NUREG-1431, Revision

[[Page 28862]]

2, ``Standard Technical Specifications Westinghouse Plants.''
    Date of issuance: May 15, 2003.
    Effective date: As of its date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 87.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 21, 2003 (68 FR 
2804).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 15, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 15, 2002, as 
supplemented February 24 and April 25, 2003.
    Brief description of amendment: The amendment increases the 
licensed reactor core power level by 1.66 percent from 3411 megawatts 
thermal (MWt) to 3468 MWt. The power level increase is considered a 
measurement uncertainty recapture power uprate.
    Date of issuance: May 2, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 259.
    Facility Operating License No. DPR-74: Amendment revises the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register: January 21, 2003 (68 FR 
2805)
    The February 24 and April 25, 2003, supplemental letters provided 
additional clarifying information that was within the scope of the 
original application and did not change the Nuclear Regulatory 
Commission staff's initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 2, 2003.
    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: January 27, 2003.
    Brief description of amendment: The amendment makes administrative 
and editorial changes to the Fort Calhoun Station Technical 
Specifications 1.3 Basis (1); 2.7 (1)a; 2.7 (1)b; 2.7 (1)d; 2.7 (1)i; 
2.7 Basis; 3.0.2; Table 3-5, Item 11; and 3.5(3)ii. The changes are 
primarily editorial and are typographical changes or corrections.
    Date of issuance: May 8, 2003.
    Effective date: May 8, 2003, and shall be implemented within 60 
days from the date of issuance.
    Amendment No.: 218.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 18, 2003 (68 FR 
12955).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 8, 2003.
    No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket No. 50-206, San 
Onofre Nuclear Generating Station, Unit 1, San Diego County, California

    Date of application for amendment: March 11, 2003.
    Brief description of amendments: The amendment application requests 
a revision to the Unit 1 defueled Technical Specifications 
administrative controls section to propose changes in organizational 
responsibilities. Specifically, the proposed changes identify that the 
Vice President, Engineering & Technical Services will be responsible 
for decommissioning activities. Additionally, the Station Manager will 
be designated as having approval authority for activities within the 
Station Manager's organization.
    Date of issuance: May 15, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: Unit 1-161.
    Facility Operating License No. DPR-13: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 15, 2003 (68 FR 
18285).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 15, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 19, 2003.
    Description of amendment request: The amendments deleted Technical 
Specification 5.5.3, ``Post Accident Sampling'' and, thereby, 
eliminated the requirements to have and maintain the post accident 
sampling system.
    Date of issuance: May 9, 2003.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment Nos.: 245, 282, 240.
    Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 18, 2003 (68 FR 
12957).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 9, 2003.
    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: July 25, 2002, as supplemented by 
letters dated February 5 and February 11, 2003.
    Brief description of amendments: The amendments change the Comanche 
Peak Steam Electric Station, Units 1 and 2, Facility Operating Licenses 
as follows: The license conditions related to Decommissioning Trusts, 
specified in Sections 2.C.(4)(a), 2.C.(4)(b), 2.C.(4)(d), 2.C(4)(e), 
and 2.C.(6), are deleted and Section 2.E, which requires reporting any 
violations of the requirements contained in Section 2.C of the 
licenses, is deleted. Additionally, Technical Specification Table 5.5-
2, ``Steam Generator Tube Inspection,'' Table 5.5-3, ``Steam Generator 
Repaired Tube Inspection for Unit 1 Only,'' and TS 5.6.10c, ``Steam 
Generator Tube Inspection Report,'' are revised to delete the 
requirement to notify the NRC pursuant to Sec.  50.72(b)(2), 
``Immediate notification requirements for operating nuclear power 
reactors,'' of Title 10 of the Code of Federal Regulations (10 CFR) if 
the steam generator tube inspection results are in a Category C-3 
classification.
    Date of issuance: May 15, 2003.
    Effective date: December 24, 2003, and shall be implemented within 
60 days from that date.
    Amendment Nos.: 103/103.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 3, 2002 (67 
FR 56329).
    The February 5, 2003, supplement was the subject of a second no 
significant hazards consideration determination (68 FR 10282, published 
March 4, 2003). The February 11, 2003, supplement provided clarifying 
information that did not change the scope of the original Federal 
Register notice or the original no significant hazards consideration 
determination.

[[Page 28863]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 15, 2003.
    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, Public File Area 01F21, 11555 
Rockville Pike (first floor), Rockville, Maryland. Publicly available 
records will be accessible from the Agencywide Documents Access and 
Management System's (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 email to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By June 26, 2003, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.714, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, and 
electronically on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are problems in accessing the 
document, contact the PDR Reference staff at 1-800-397-4209, 301-415-
4737, or by e-mail to [email protected]. 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

[[Page 28864]]

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 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, Public File Area 01F21, 11555 
Rockville Pike (first floor), Rockville, Maryland, by the above date. 
Because of the continuing disruptions in delivery of mail to United 
States Government offices, it is requested that petitions for leave to 
intervene and requests for hearing be transmitted to the Secretary of 
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the petition for 
leave to intervene and request for hearing should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and because of continuing disruptions in 
delivery of mail to United States Government offices, it is requested 
that copies be transmitted either by means of facsimile transmission to 
301-415-3725 or by e-mail to [email protected]. A copy of the 
request for hearing and petition for leave to intervene should also be 
sent 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).

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of application for amendments: April 25, 2003.

    Brief description of amendments: The amendments modify Technical 
Specification surveillance requirements to provide an alternative means 
of testing the Unit 2 main steam power operated relief valves, 
including those that provide the automatic depressurization system and 
low set relief functions.
    Date of issuance: May 8, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 215/209.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. Quad-City Times, dated May 5, 2003. The 
notice provided an opportunity to submit comments on the Commission's 
proposed NSHC determination. No comments have been received.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a Safety Evaluation dated May 8, 2003.

    Dated at Rockville, Maryland, this 19th day of May 2003.

    For the Nuclear Regulatory Commission.
William H. Ruland,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 03-12973 Filed 5-23-03; 8:45 am]
BILLING CODE 7590-01-U