[Federal Register Volume 68, Number 227 (Tuesday, November 25, 2003)]
[Notices]
[Pages 66131-66144]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-29107]


-----------------------------------------------------------------------

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, October 31, through November 13, 2003. The 
last biweekly notice was published on November 12, 2003 (68 FR 64133).

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.

[[Page 66132]]

However, should circumstances change during the notice period such that 
failure to act in a timely way would result, for example, in derating 
or shutdown of the facility, the Commission may issue the license 
amendment before the expiration of the 30-day notice period, provided 
that its final determination is that the amendment involves no 
significant hazards consideration. The final determination will 
consider all public and State comments received before action is taken. 
Should the Commission take this action, it will publish in the Federal 
Register a notice of issuance and provide for opportunity for a hearing 
after issuance. The Commission expects that the need to take this 
action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the 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 December 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,

[[Page 66133]]

supplemental petitions and/or requests for a hearing will not be 
entertained absent a determination by the Commission, the presiding 
officer or the Atomic Safety and Licensing Board that the petition and/
or request should be granted based upon a balancing of factors 
specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, 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].
    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: October 14, 2003.
    Description of amendments request: The proposed amendment would 
change the frequency of surveillance testing for some engineered safety 
features (ESF) components.
    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.
    Integrated testing of the ESF trains takes place while the unit 
is shut down. The equipment being tested is normally used to respond 
to an accident when the Unit is in Modes 1, 2, or 3. Changing the 
test Frequency to a longer period does not affect the scope of the 
testing or the methods used during the testing. Therefore, there is 
no increase in the probability of an accident previously evaluated 
caused by the testing itself.
    The components tested during the integrated ESF test are 
components needed to mitigate the consequences of an accident. 
Increasing the length of time between integrated tests increases the 
likelihood of undetected equipment failure. This creates a change in 
plant risk. This change in risk is analyzed and quantified using 
probabilistic risk assessment techniques. The risk analysis provides 
results that show the proposed increase in ESF component 
surveillance testing Frequency meets the guidance of Regulatory 
Guide 1.174, ``An Approach for Using Probabilistic Risk Assessment 
in Risk-Informed Decisions on Plant-Specific Changes to the 
Licensing Basis.'' The increase in risk is within the guidelines of 
the regulatory guidance. There is no significant change in the 
probability that the equipment will suffer an undetected failure in 
the increased time between Surveillance tests. Therefore, there is 
no significant increase in the consequences o[f] an accident 
previously evaluated.
    An additional change is proposed to delete a Surveillance 
Requirement because the signal tested in the Surveillance 
Requirement is no longer installed in the plant. This deletion has 
no impact on plant operations or the response of the plant in an 
accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Would not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed change would extend the Surveillance Frequency of 
the integrated ESF test. This change does not affect the scope of 
the testing or the methods used during the testing. Plant equipment 
will continue to operate as designed. Only the testing frequency is 
changed. Because there are no changes in the scope or method of 
testing and this proposed change does not affect the operation of 
the equipment in other circumstances, no new accident initiators 
have been introduced.
    An additional change is proposed to delete a Surveillance 
Requirement because the signal tested in the Surveillance 
Requirement is no longer installed in the plant. This deletion has 
no impact on plant operations or the response of the plant and 
therefore would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    Therefore, this proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Would not involve a significant reduction in [a] margin of 
safety.
    Surveillance testing is performed to evaluate the operability of 
equipment used to perform safety functions at the Unit. The 
components tested during the integrated ESF test are components 
needed to mitigate the consequences of an accident. Increasing the 
length of time between integrated tests increases the likelihood of 
undetected equipment failure. This creates a change in plant risk. 
This change in risk is analyzed and quantified using probabilistic 
risk assessment techniques. The risk analysis provides results that 
show the proposed increase in ESF component surveillance testing 
Frequency meets the guidance of Regulatory Guide 1.174. The increase 
in risk is within the guidelines of the regulatory guidance. There 
is no significant change in the probability that the equipment will 
suffer an undetected failure in the increased time between 
Surveillance tests. Since the function of Surveillance testing is to 
evaluate the operability of equipment, and the increased time 
between Surveillance tests has been evaluated and found to be 
acceptable under regulatory guidance, the proposed change would not 
involve a significant reduction in [a] margin of safety.
    An additional change is proposed to delete a Surveillance 
Requirement because the signal tested in the Surveillance 
Requirement is no longer installed in the plant. This deletion has 
no impact on plant operations or the response of the plant in an 
accident and does not impact the margin of safety.

    Therefore, this proposed change does not significantly reduce [a] 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Attorney for licensee: James M. Petro, Jr., Esquire, Counsel, 
Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor, 
Baltimore, MD 21202.
    NRC Section Chief: Richard J. Laufer.

    Consumers Energy Company, Docket No. 50-155, Big Rock Point Nuclear 
Plant, Charlevoix County, Michigan.
    Date of amendment requests: August 6, 2003.
    Description of amendment requests: The Big Rock Point Plant is in 
the 6th year of decommissioning. The reactor was defueled and certified 
as permanently shutdown by letter to the Nuclear Regulatory Commission 
dated September 22, 1997. As of March 26, 2003, all the spent fuel has 
been permanently removed from the plant's spent fuel pool and located 
to an Independent Spent Fuel Storage Installation (ISFSI). The spent 
fuel has been loaded into an NRC approved and licensed Spent Fuel Dry 
Storage System and will be temporarily stored at this installation 
until such time that a permanent repository is available. The 
requirements associated with the wet storage of the spent fuel as 
described in Defueled Technical Specifications are no longer applicable 
and are being revised.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?

[[Page 66134]]

    No. The proposed change is an administrative change to update 
the facility's Operating License and Defueled Technical 
Specifications to reflect the permanent removal of the spent fuel 
from the Spent Fuel Pool. Requirements for safe storage and handling 
of irradiated fuel, definitions, design features and administrative 
controls that were applicable to the facility when spent fuel was 
stored in the spent fuel pool are no longer valid and are being 
removed to provide clarity to the licensing basis of the facility in 
its current configuration. The accidents previously evaluated in the 
Updated Final Hazards Safety Analysis are based on spent nuclear 
fuel being stored in the spent fuel pool. Since the spent fuel has 
been permanently removed from the spent fuel pool, the accidents 
previously analyzed are no longer credible. The spent fuel has been 
loaded into an NRC approved and licensed Spent Fuel Dry Storage 
System and will be temporarily stored at this installation until 
such time that a permanent repository is available. The spent fuel 
is now controlled by a different set of approved technical 
specifications issued and approved pursuant to 10 CFR part 72. 
Therefore, the proposed administrative change does not affect the 
consequences of any accident described and evaluated in the Updated 
Final Hazards Summary Report, and the accidents and transients 
associated with spent fuel stored in the facility's spent fuel pool 
are no longer applicable.
    Therefore, the proposed administrative change to the Operating 
License and Defueled Technical Specifications does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Will the proposed change create the possibility of a new or 
different type of accident from any accident previously evaluated?
    No. The spent fuel has been loaded into an NRC approved and 
licensed Spent Fuel Dry Storage System and will be temporarily 
stored at this installation until such time that a permanent 
repository is available. In accordance with 10 CFR part 72, 
``Licensing Requirements for the Independent Storage of Spent 
Nuclear Fuel and High-Level Radioactive Waste,'' credible accidents 
have been evaluated as part of the licensing and approval process 
for the Dry Fuel Storage System. The requirement to evaluate 
credible accidents has not changed.
    Therefore this proposed administrative change does not create 
the possibility of a new or different kind of accident previously 
evaluated.
    3. Will the proposed change involve a significant reduction in a 
margin of safety?
    The proposed activity is an administrative change to the 
Operating License and Defueled Technical Specifications to reflect 
the permanent removal of the spent fuel from the spent fuel pool and 
does not involve any significant reduction in any margin of safety 
that is usually associated with the design and performance of 
systems, structures and components. Requirements for safe storage 
and handling of irradiated fuel, definitions, design features and 
administrative controls that were applicable to the facility when 
spent fuel was stored in the spent fuel pool are no longer 
applicable and are being removed to provide clarity to the licensing 
basis of the facility in its current configuration.
    Therefore, the proposed administrative 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: David A. Mikelonis, Esquire, Consumers 
Energy Company, One Energy Plaza, Jackson, MI 49201-2276.
    NRC Section Chief: Claudia Craig.

    Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan.
    Date of amendment request: October 10, 2003.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 3.7.3, ``Control Room Emergency 
Filtration (CREF) System,'' Surveillance Requirement (SR) 3.7.3.6, to 
permit a one-time extension of SR 3.7.3.6 until startup from the next 
refueling outage (RF-10) to preclude a mid-cycle shutdown solely for 
the performance of this SR. SR 3.7.3.6 requires verifying that 
unfiltered inleakage from CREF system duct work outside the control 
room envelope that is at negative pressure during accident conditions 
is within limits. This SR is required to be performed every 36 months, 
and can be performed only when the CREF system is not required to be 
Operable (i.e., in MODES 4 or 5, with no operations with a potential 
for draining the reactor vessel and with no fuel movement of recently 
irradiated fuel in progress).
    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 allows a one-time extension of SR 3.7.3.6 
until startup from the next refueling outage (approximately 10 to 12 
months beyond its critical completion date). The Control Room 
Emergency Filtration (CREF) system provides a configuration for 
mitigating radiological consequences of accidents; however, it is 
not considered an initiator of any previously analyzed accident. 
Therefore, the proposed change 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. The current TS surveillance (SR 3.7.3.6) measures 
inleakage from four sections of CREF system duct work outside the 
Control Room Envelope (CRE) that are at negative pressure during 
accident conditions. Based on the results of previous surveillance 
testing, and the continued performance of SR 3.7.3.3 and 3.7.3.5 on 
their normal schedule, the delay in performing SR 3.7.3.6 by 
approximately 10 to 12 months will provide essentially the same 
degree of assurance that CRE integrity is being maintained as 
before. It is expected that CRE integrity will remain essentially 
unchanged from what it is today. Therefore, the proposed change does 
not significantly increase the radiological consequences of any 
previously analyzed accident.
    Based on the above, the proposed change does not significantly 
increase the probability or 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 accident previously 
evaluated.
    The proposed change to allow a one-time extension of SR 3.7.3.6 
until startup from the next refueling outage (approximately 10 to 12 
months beyond its critical completion date) does not alter the 
design or function of the system involved, nor does it introduce any 
new modes of plant or CREF system operation. Therefore, the proposed 
change does not create the potential for a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed change to allow a one-time extension of SR 3.7.3.6 
until startup from the next refueling outage (approximately 10 to 12 
months beyond its critical completion date) will not affect the 
radiological release from a design basis accident. Based on the 
results of previous surveillance testing and the continued 
performance of SR 3.7.3.3 and 3.7.3.5 on their normal schedule, the 
delay in performing SR 3.7.3.6 by approximately 10 to 12 months will 
provide essentially the same degree of assurance that CRE integrity 
is being maintained as existed before; and, the postulated dose to 
the control room occupants as a result of an accident will remain 
approximately the same. 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.

[[Page 66135]]

    NRC Section Chief: L. Raghavan.

    Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi.
    Date of amendment request: October 24, 2003.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.1.8, ``Scram Discharge Volume (SDV) 
Vent and Drain Valves,'' to allow a vent or drain line with one 
inoperable valve to be isolated instead of requiring the valve to be 
restored to Operable status within 7 days.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on February 24, 2003 (68 FR 8637), on possible 
amendments to revise the action for one or more SDV vent or drain lines 
with an inoperable valve, 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 April 15, 
2003 (68 FR 18294). The licensee affirmed the applicability of the 
model NSHC determination in its application dated October 24, 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:

    Criterion 1--The proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    A change is proposed to allow the affected SDV vent and drain 
line to be isolated when there are one or more SDV vent or drain 
lines with one valve inoperable instead o[f] requiring the valve to 
be restored to operable status within 7 days. With one SDV vent or 
drain valve inoperable in one or more lines, the isolation function 
would be maintained since the redundant valve in the affected line 
would perform its safety function of isolating the SDV. Following 
the completion of the required action, the isolation function is 
fulfilled since the associated line is isolated. The ability to vent 
and drain the SDVs is maintained and controlled through 
administrative controls. This requirement assures the reactor 
protection system is not adversely affected by the inoperable 
valves. With the safety functions of the valves being maintained, 
the probability or consequences of an accident previously evaluated 
are not significantly increased.
    Criterion 2--The proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any previously evaluated.
    Criterion 3--The proposed change does not involve a significant 
reduction in the margin of safety.
    The proposed change ensures that the safety functions of the SDV 
vent and drain valves are fulfilled. The isolation function is 
maintained by redundant valves and by the required action to isolate 
the affected line. The ability to vent and drain the SDVs is 
maintained through administrative controls. In addition, the reactor 
protection system will prevent filling of an SDV to the point that 
it has insufficient volume to accept a full scram. Maintaining the 
safety functions related to isolation of the SDV and insertion of 
control rods ensures that the proposed 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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.
    Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont.
    Date of amendment request: July 31, 2003, as supplemented on 
October 10, 2003.
    Description of amendment request: This amendment request 
incorporates a revision to the licensing basis of the Vermont Yankee 
Nuclear Power Station (VYNPS) that supports a full scope application on 
an Alternative Source Term (AST) methodology.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) Section 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration which is 
presented below:

    1. Will the proposed changes involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Adoption of the AST and those plant systems affected by 
implementation of the AST do not initiate DBAs [design basis 
accidents]. The proposed change does not affect the design or manner 
in which the facility is operated; rather, once the occurrence of an 
accident has been postulated, the new accident source term is an 
input to analyses that evaluate the radiological consequences. 
Therefore, the proposed change does not involve an increase in the 
probability of an accident previously evaluated.
    The structures, systems and components (SSCs) affected by the 
proposed change act as mitigators to the consequences of accidents. 
Based on the revised analyses, the proposed changes do revise 
certain performance requirements; however, the proposed changes do 
not involve a revision to the parameters or conditions that could 
contribute to the initiation of a design basis accident discussed in 
Chapter 14 of the Updated Final Safety Analysis Report.
    Because of the changed methodology, it is difficult to draw a 
quantitative comparison of before and after accident consequences 
due to the use of different dose calculations, conversion factors, 
source term, and other assumptions. However qualitatively, it can be 
shown that there is no significant increase in offsite doses, 
although there may be small variations in potential doses for 
postulated accidents. Plant-specific radiological analyses have been 
performed using the AST methodology. Based on the results of these 
analyses, it has been demonstrated that the dose consequences of the 
limiting events considered in the analyses meet the regulatory 
guidance provided for use with the AST, and the offsite doses are 
well within acceptable limits. This guidance is presented in 10 CFR 
50.67, Regulatory Guide 1.183, and Standard Review Plan (SRP) 
Section 15.0.1.
    Therefore, the proposed amendment does not result in a 
significant increase in the consequences or increase the probability 
of any previously evaluated accident.
    2. Will the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    Implementation of AST and the proposed changes does not alter or 
involve any design basis accident initiators. These changes do not 
affect the design function or mode of operations of SSCs in the 
facility prior to a postulated accident. Since SSCs are operated 
essentially no differently after the AST implementation, no new 
failure modes are created by this proposed change.
    Therefore, the proposed license amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Will the proposed changes involve a significant reduction in 
a margin of safety?
    The changes proposed are associated with a revision to the 
licensing basis for the VYNPS. Approval of the licensing basis 
change from the original source term to the alternative source term 
is requested by this application for a license amendment. The 
results of the accident analyses revised in support of the proposed 
change are subject to the acceptance criteria in 10 CFR 50.67. The 
analyzed events have been carefully selected, and the analyses 
supporting these changes have been performed using approved 
methodologies to ensure that analyzed events are bounding and safety 
margin has not been reduced. The dose consequences of these limiting 
events are within the acceptance criteria presented in 10 CFR 50.67, 
Regulatory Guide 1.183, and SRP 15.0.1. Thus, by meeting the 
applicable regulatory

[[Page 66136]]

limits for AST, there is no significant reduction in a margin of 
safety.
    Therefore, because the proposed changes continue to result in 
dose consequences within the applicable regulatory limits, the 
changes are considered to not result in 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: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.
    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: October 17, 2003.
    Description of amendment request: The proposed amendments revise 
the action requirements of Technical Specification (TS) 3/4 6.3, 
``Containment Isolation Valves [CIVs],'' to more clearly define action 
requirements for inoperable CIVs. The proposed changes to the action 
requirements also include: (1) Provisions for allowing the intermittent 
unisolation of penetration flow paths which have been isolated per 
action requirements under administrative control; (2) use of check 
valves as an isolation device; and (3) an increase in the allowed 
outage time to 72 hours for CIVs associated with closed systems inside 
containment. The proposed amendments also revise the TS surveillance 
requirements (SRs) for CIVs by replacing existing SRs with new SRs 
similar to those in NUREG-1431, Revision 2, ``Standard Technical 
Specifications for Westinghouse Plants.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not involve any changes to plant 
equipment, system design functions or a change in the methods 
governing normal plant operation. Therefore, the probability of a 
malfunction of a structure, system or component to perform its 
design function will not be increased.
    The proposed change modifies existing action requirements for 
inoperable containment isolation valves. Action requirements and 
their associated allowed outage times are not initiating conditions 
for any accident previously evaluated and the accident analyses do 
not assume that repaired equipment is out of service prior to the 
analyzed event. In addition, changes that are consistent with the 
ISTS [improved Standard Technical Specifications] have been 
previously evaluated and found not to adversely affect the safe 
operation of Westinghouse plants or the initiation of any accident 
previously evaluated. Based on the conclusions of the plant specific 
evaluation associated with the changes and the evaluation performed 
in developing the ISTS, the proposed revised action requirements do 
not result in operating conditions that will significantly increase 
the probability of initiating an analyzed event. The revised action 
requirements provide appropriate remedial actions to be taken in 
response to the degraded condition considering the operability 
status of the redundant systems of required features, and the 
capability of remaining features while minimizing the risk 
associated with continued operation. As a result, the consequences 
of any accident previously evaluated are not significantly 
increased.
    The proposed change also modifies and deletes some surveillance 
requirements. Surveillances are not initiators to any accident 
previously evaluated. Consequently, the probability of an accident 
previously evaluated is not significantly increased. The equipment 
specified in the Limiting Condition for Operation is still required 
to be operable and capable of performing the accident mitigation 
functions assumed in the accident analysis. This equipment will 
continue to be tested in a manner and at a frequency to give 
confidence that the equipment can perform its assumed safety 
function. The proposed changes are generally made to conform to the 
ISTS and have been evaluated to not be detrimental to plant safety. 
As a result, the proposed surveillance requirement changes do not 
significantly affect the consequences of any accident previously 
evaluated. 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 change does not involve any changes to plant 
equipment, system design functions or a change in the methods 
governing normal plant operation. The [technical] specification for 
containment isolation valves provide[s] controls for maintaining the 
containment pressure boundary. The revised action requirements and 
revised surveillance requirements are sufficient to ensure the 
containment isolation valves are capable of performing their 
accident mitigation functions. No new accident initiators are 
introduced by these changes. 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 revised action requirements do not involve a significant 
reduction in the margin of safety. The proposed actions for 
inoperable containment isolation valves minimize the risk of 
continued operation under the specified conditions, considering the 
operability status of the redundant containment isolation barriers, 
a reasonable time for repairs or replacement of the isolation 
feature, and the low probability of a design basis accident 
occurring during the repair period.
    The revised surveillance requirements do not involve a 
significant reduction in the margin of safety. The proposed 
surveillance requirements provide the required verifications for 
ensuring containment isolation valves operability. Containment 
isolation valve testing will continue to be performed in a manner 
and at a frequency necessary to give confidence that the equipment 
can perform its assumed safety function.
    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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Richard J. Laufer.
    FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-
Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio.
    Date of amendment request: December 17, 2001, as supplemented by 
letter dated June 4, 2002.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) Section 3/4.3.1, ``Reactor Coolant 
System Instrumentation,'' to delete an action involving either reducing 
core thermal power and the high neutron flux reactor trip setpoint or 
monitoring quadrant power tilt when a reactor protection system (RPS) 
channel is inoperable. Additionally, changes to the content and format 
of TS Tables 3.3-1 and 4.3-1 are proposed to enhance specification 
clarity.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided their analysis of

[[Page 66137]]

the issue of no significant hazards consideration. The 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. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not result in an increase in the 
probability of an accident previously evaluated because no change is 
being made to any accident initiator. The proposed change does not 
result in an increase in the consequences of an accident previously 
evaluated because TS 3/4.2.4, ``Quadrant Power Tilt,'' continues to 
ensure the radial power distribution of the core is within the 
limits assumed in the accident analyses. In addition, compensatory 
actions will continue to be required should a single channel of RPS 
High Flux or Flux-'Flux-Flow become inoperable. 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 changes affect the TS requirements for the RPS 
instrumentation. The proposed changes do not change the RPS design 
function or result in the RPS being operated outside its design 
operating range. There are no new or different equipment failure 
modes introduced by the proposed changes. The proposed changes do 
not introduce any new or different accident initiators. Therefore, 
the proposed changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes affect the TS requirements for the RPS 
instrumentation. The capability of the RPS to perform its required 
functions is not adversely affected by the proposed changes. The 
proposed changes do not alter any initial conditions contributing to 
accident severity or consequences. There will be no changes to the 
plants' systems, structures, or components, nor in the manner in 
which they will be operated as a result of the proposed changes. 
Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.
    Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine.
    Date of amendment request: September 11, 2003.
    Description of amendment request: Revise the dose model for the 
containment activated concrete, rebar (hereafter referred to as 
activated concrete) and liner, by incorporating more realistic 
radionuclide release rates and to change the associated derived 
concentration guideline limit (DCGL) for activated concrete.
    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?
    Response: No.
    The requested license amendment does not authorize any plant 
activities beyond those allowed by 10 CFR Chapter I or beyond those 
considered in the DSAR. The bounding accident described in the 
Defueled Safety Analysis Report (DSAR) for potential airborne 
activity is the postulated resin cask drop accident in the Low Level 
Radioactive Waste Storage Building. This accident is expected to 
contain more potential airborne activity than can be released from 
other decommissioning events. The radionuclide distribution assumed 
for the spent resin cask has a greater inventory of transuranic 
radionuclides (the major dose contributor) than the distribution of 
plant derived radionuclides in the components involved in other 
decommissioning accidents. The other accidents considered in the 
DSAR include: (1) Explosion of liquid petroleum gas (LPG) leaked 
from a front end loader or forklift; (2) Explosion of oxyacetylene 
during segmenting of the reactor vessel shell; (3) Release of 
radioactivity from the RCS decontamination ion exchange resins; (4) 
Gross leak during in-situ decontamination; (5) Segmentation of RCS 
piping with unremoved contamination; (6) Fire involving contaminated 
clothing or combustible waste; (7) Loss of local airborne 
contamination control during blasting or jackhammer operations; (8) 
Temporary Loss of Services; (9) Dropping of Contaminated Concrete 
Rubble; (10) Natural phenomena; and (11) Transportation accidents. 
The probabilities and consequences for these accidents are estimated 
in the basis documentation for DSAR Section 7. No systems, 
structures, or components that could initiate or be required to 
mitigate the consequences of an accident are affected by the 
proposed change in any way not previously evaluated in the DSAR. 
Since Maine Yankee does not exceed the salient parameters associated 
with the plant referenced in the basis documentation in any material 
respects, it is concluded that these probabilities and consequences 
are not increased. Therefore, the proposed change to the Maine 
Yankee license does not involve any increase in the probability or 
consequences of any 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 requested license amendment does not authorize any plant 
activities that could precipitate or result in any accidents beyond 
those considered in the DSAR. The accidents previously evaluated in 
the DSAR are described above. These accidents are described in the 
basis documentation for DSAR Section 7. The proposed change does not 
affect plant systems, structures, or components in any way not 
previously evaluated in the DSAR. Since Maine Yankee does not exceed 
the salient parameters associated with the plant referenced in the 
basis documentation in any material respects, it is concluded that 
these accidents appropriately bound the kinds of accidents possible 
during decommissioning. Therefore, the proposed change to the Maine 
Yankee license would 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.
    The margin of safety defined in Maine Yankee's license basis for 
the consequences of decommissioning accidents has been established 
as the margin between the bounding decommissioning accident and the 
dose limits associated with the need for emergency plan offsite 
protection, namely the Environmental Protection Agency Protective 
Action Guidelines EPA-PAGs. As described above, the bounding 
decommissioning accident is the postulated resin cask drop accident 
in the Low Level Radioactive Waste Storage Building. Since the 
bounding decommissioning accident is expected to contain more 
potential airborne activity than can be released from other 
decommissioning events and since the radionuclide distribution 
assumed for the spent resin cask has more transuranics (the major 
dose contributor) than the distribution in the components involved 
in other decommissioning accidents, the margin of safety associated 
with the consequences of decommissioning accidents cannot be 
reduced. The margin of safety defined in the statements of 
consideration for the final rule on the Radiological Criteria for 
License Termination is described as the margin between the 100 mrem/
yr public dose limit established in 10 CFR 20.1301 for licensed 
operation and the 25 mrem/yr dose limit to the average member of the 
critical group at a site considered acceptable for unrestricted use. 
This margin of safety accounts for the potential effect of multiple 
sources of radiation exposure to the critical group. Since the 
license termination plan (LTP) was designed to comply with the 
radiological criteria for license termination for unrestricted use, 
the margin of safety cannot be reduced. Therefore, the proposed 
changes to the Maine Yankee license would not involve a significant 
reduction in any margin of safety.

[[Page 66138]]

Conclusion

    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.

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

    Date of amendment request: September 26, 2003.
    Description of amendment request: The proposed amendments would 
modify TS 5.6.5.b to add a reference to a Nuclear Regulatory Commission 
(NRC) letter that would approve the use of a new master curve 
methodology for Unit 2. The NRC staff is currently reviewing an 
associated exemption request by NMC to use this new methodology. The 
requested exemption would allow the use of the master curve methodology 
described in Babcock & Wilcox Report BAW-2308, Revision 1, ``Initial 
RTNDT [reference nil-ductility temperature] of Linde 80 Weld 
Materials,'' for determining the adjusted RTNDT of the Unit 
2 reactor vessel limiting circumferential weld metal. This method is 
used for the pressurized thermal shock screening evaluation. The 
proposed amendments would also make editorial changes to TS 5.6.5.b.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Operation of PBNP in accordance with the proposed amendments 
does not result in a significant increase in the probability or 
consequences of any accident previously evaluated.
    The proposed change references the NRC safety evaluation 
[currently under NRC staff review] accepting the new Master Curve 
Methodology used in the evaluation of the revised P/T [pressure/
temperature] limits and LTOP [low-temperature overpressure 
protection] setpoints. Implementation of revisions to Topical 
Reports would still be reviewed in accordance with 10 CFR 50.59 and, 
where required, receive NRC review and approval. The proposed change 
does not adversely affect accident initiators or precursors nor 
alter the design assumptions, conditions, or configuration of the 
facility or the manner in which the plant is operated and 
maintained. The proposed change does not alter or prevent the 
ability of structures, systems, and components (SSCs) from 
performing their intended function to mitigate the consequences of 
an initiating event within the assumed acceptance limits. The 
proposed change does not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
Further, the proposed change does not increase the types or amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures. The proposed change is consistent with safety 
analysis assumptions and resultant consequences. Therefore, it is 
concluded that this change does not increase the probability of 
occurrence of an accident previously evaluated.
    2. Operation of PBNP in accordance with the proposed amendments 
does not result in a new or different kind of accident from any 
accident previously evaluated.
    The proposed change references the NRC safety evaluation 
[currently under NRC staff review] accepting the new Master Curve 
Methodology used in the evaluation of the revised P/T limits and 
LTOP setpoints. Implementation of revisions to Topical Reports would 
still be reviewed in accordance with 10 CFR 50.59 and, where 
required, receive NRC review and approval. The change does not 
involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operation. In addition, the changes 
do not impose any new or different requirements or eliminate any 
existing requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Operation of PBNP in accordance with the proposed amendments 
does not result in a significant reduction in a margin of safety.
    The proposed change references the NRC safety evaluation 
[currently under NRC staff review] accepting the new Master Curve 
Methodology used in the evaluation of the revised P/T limits and 
LTOP setpoints. Implementation of revisions to Topical Reports would 
still be reviewed in accordance with 10 CFR 50.59 and, where 
required, receive NRC review and approval. The proposed change does 
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.

    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: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Section Chief: L. Raghavan.

    Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo 
County, California.
    Date of amendment requests: September 12, 2003.
    Description of amendment requests: The proposed license amendments 
would revise Technical Specification (TS) 3.3.1, ``Reactor Trip System 
(RTS) Instrumentation,'' and TS 3.3.2, ``Engineered Safety Feature 
Actuation System (ESFAS) Instrumentation,'' to change the current steam 
generator (SG) narrow range (NR) water level-low low setpoints from 
greater than or equal to 7.0 percent allowable value and 7.2 percent 
nominal value, to greater than or equal to 14.8 percent allowable value 
and 15.0 percent nominal value.
    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 protection system performance will remain within the bounds 
of the previously performed accident analyses since there are no 
hardware changes and the actuation logic changes are conservative. 
The design of the steam generator (SG) water level sensing equipment 
and the coincidence logic will be unaffected. The only physical 
change to the reactor trip system (RTS) and the engineered safety 
feature actuation system (ESFAS) instrumentation is the increased 
actuation setpoints. These changes have already been implemented in 
the plant through the design change process. These changes are in 
the conservative direction, i.e., a trip actuation signal will be 
generated sooner for an event that challenges the ability of the SGs 
to provide a heat sink for the reactor. In all other regards, the 
design of the RTS and ESFAS instrumentation will be unaffected. 
These protection systems will continue to function in a manner 
consistent with the plant design basis.

[[Page 66139]]

    The probability and consequences of accidents previously 
evaluated in the Final Safety Analysis Report Update (FSARU) are not 
adversely affected because changes to the RPS and ESFAS trip 
setpoints assure a conservative response of the affected trip 
functions, consistent with the safety analyses and licensing basis.
    The proposed changes will not affect the probability of any 
accident initiators. There will be no degradation in the performance 
of, or an increase in the number of challenges imposed on, safety-
related equipment assumed to function during an accident. There will 
be no change to normal plant operating parameters or accident 
mitigation performance.
    The proposed changes will not alter any assumptions or change 
any mitigation actions in the radiological consequence evaluations 
in the FSARU.
    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.
    The proposed changes do not change any hardware or the design 
functions of any structures, systems or components involved, other 
than to revise the SG narrow range (NR) water level-low low 
setpoints; changes that have already been implemented. The proposed 
changes will not affect the normal method of plant operation or 
change any operating parameters. No new accidents, accident 
initiators, or failure mechanisms are created by the proposed 
changes.
    Therefore, the proposed changes do not create the possibility of 
a new or different accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The SG NR water level-low low setpoints specified in the 
Technical Specifications have already been increased in the 
conservative direction. The safety analysis limits assumed in the 
transient and accident analyses remain unchanged. None of the 
acceptance criteria for any accident analysis are changed.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

    Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo 
County, California.
    Date of amendment requests: October 22, 2003.
    Description of amendment requests: The proposed license amendments 
would revise Surveillance Requirement 3.6.3.7 of Technical 
Specification (TS) 3.6.3, ``Containment Isolation Valves,'' by 
extending the leakage rate testing frequency of the containment purge 
supply and exhaust and vacuum/pressure relief valves, all with 
resilient seals, from 184 days to 24 months. The amendments would also 
delete the requirement to leakage rate test the containment vacuum/
pressure relief valves within 92 days after opening.
    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.
    Operability and leakage control effectiveness of the containment 
purge supply and exhaust and containment vacuum/pressure relief 
isolation valves have no effect on whether an accident occurs. 
Consequently, increasing the interval between surveillances of 
isolation valve leak rate does not involve any significant increase 
in the probability of an accident previously evaluated. The 
consequences of a unisolated reactor containment building at the 
time of a fuel-handling accident or loss of coolant accident (LOCA) 
are the release of radionuclides to the environment. Offsite 
exposures due to containment leakage during a LOCA and fuel-handling 
accident have been evaluated in Final Safety Analysis Report Update 
(FSARU) sections 15.5.17.3 and 15.5.22, respectively. For a LOCA, 
the Diablo Canyon Power Plant (DCPP) analyses assume containment 
leakage of 0.1 percent of the containment volume per day for the 
first 24 hours and 0.05 percent per day for the rest of the duration 
of the accident. Calculated radiological exposures from the LOCA are 
listed in FSARU Chapter 15, Table 15.5-75 and are within the 10 CFR 
part 100 limits. The good performance history of these valves, along 
with the very low total containment leakage rate, are reasonable 
bases that there should not be any significant increase in the 
consequences of [an] accident previously evaluated. For the fuel-
handling accident inside containment, DCPP analyses do not credit 
these valves to provide a containment isolation function. It was 
assumed that activity released from the containment refueling pool 
is transported to the environment over a short time period through 
the open equipment hatch. Calculated radiological exposures from the 
fuel-handling accident inside containment are listed in FSARU 
Chapter 15, Table 15.5-50 and are also within the 10 CFR part 100 
limits. In summary, increasing the interval between leakage rate 
surveillances of these isolation valves will not involve any 
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 changes do not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed change will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. The functions of the 
containment purge and containment vacuum/pressure relief systems are 
not altered by this change. Therefore, the proposed change does not 
create the possibility of a new or different accident from any 
accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    This proposed change only increases the interval between 
surveillance tests of the containment purge supply and exhaust, and 
containment vacuum/pressure relief valves. These valves have a good 
performance history and should be able to perform their intended 
containment isolation function reliably when called upon. In FSARU 
Chapter 15, two offsite exposure scenarios are applicable to the 
containment isolation function. These scenarios are LOCA containment 
leakage and fuel-handling accident inside containment. For LOCA 
containment leakage, the DCPP analyses assume containment leakage of 
0.1 percent of the containment volume per day for the first 24 hours 
and 0.05 percent per day for the remainder of the accident. 
Calculated radiological exposures from a LOCA are listed in FSARU 
Chapter 15, Table 15.5-75 and meet the 10 CFR part 100 limits. For 
the fuel-handling accident inside containment, the DCPP analyses do 
not credit these valves to provide a containment isolation function. 
The analyses assume that activity released from the containment 
refueling pool is transported to the environment over a short time 
period through the open equipment hatch. Calculated radiological 
exposures from the fuel-handling accident inside containment are 
listed in FSARU Chapter 15, Table 15.5-50 and also meet the 10 CFR 
part 100 limits. If in the unlikely event that these valves exceed 
their leakage rate limits due to the extension of the surveillance 
interval, the consequences will be consistent with the containment 
leakage assumed in the accident analyses. Therefore, the extension 
of leakage rate test interval will have an insignificant 
radiological consequence, and the proposed change will not involve 
any 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

[[Page 66140]]

amendment requests involve no significant hazards consideration.
    Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

    Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo 
County, California.
    Date of amendment requests: October 22, 2003.
    Description of amendment requests: The proposed license amendments 
would revise Technical Specifications (TS) Section 5.5.9, ``Steam 
Generator (SG) Tube Surveillance Program,'' and TS Section 5.6.10, 
``Steam Generator (SG) Tube Inspection Report,'' to allow use of leak 
limiting Alloy 800 sleeves to repair degraded SG tubes as an 
alternative to plugging the SG tubes. The proposed amendments would 
also remove an unnecessary reporting requirement contained in TS Table 
5.5.9-2, ``Steam Generator (SG) Tube Inspection.''
    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 leak limiting Alloy 800 sleeves are designed using the 
applicable American Society of Mechanical Engineers (ASME) Boiler 
and Pressure Vessel Code and, therefore, meet the design objectives 
of the original steam generator (SG) tubing. The applied stresses 
and fatigue usage for the sleeves are bounded by the limits 
established in the ASME Code. Mechanical testing has shown that the 
structural strength of sleeves under normal, upset, emergency, and 
faulted conditions provides margin to the acceptance limits. These 
acceptance limits bound the most limiting (three times normal 
operating pressure differential) burst margin recommended by NRC 
Regulatory Guide 1.121, ``Bases for Plugging Degraded PWR Steam 
Generator Tubes.'' Burst testing of sleeve-tube assemblies has 
confirmed the analytical results and demonstrated that no 
unacceptable levels of primary-to-secondary leakage are expected 
during any plant condition.
    The leak limiting Alloy 800 sleeve depth-based structural limit 
is determined using NRC guidance and the pressure stress equation of 
ASME Code, Section III with additional margin added to account for 
the configuration of long axial cracks. A sleeved tube will be 
plugged on detection of an imperfection in the sleeve or in the 
pressure boundary portion of the original tube wall in the leak 
limiting sleeve/tube assembly.
    Evaluation of the repaired SG tube testing and analysis 
indicates no detrimental effects on the leak limiting Alloy 800 
sleeve or sleeved tube assembly from reactor system flow, primary or 
secondary coolant chemistries, thermal conditions or transients, or 
pressure conditions as may be experienced at Diablo Canyon Power 
Plant (DCPP) Units 1 and 2. Corrosion testing and historical 
performance of sleeve-tube assemblies indicates no evidence of 
sleeve or tube corrosion considered detrimental under anticipated 
service conditions.
    The implementation of the proposed change has no significant 
effect on either the configuration of the plant or the manner in 
which it is operated. The consequences of a hypothetical failure of 
the leak limi[ti]ng Alloy 800 sleeve-tube assembly is bounded by the 
current SG tube rupture (SGTR) analysis described in the DCPP Final 
Safety Analysis Report Update. Due to the slight reduction in the 
inside diameter caused by the sleeve wall thickness, primary coolant 
release rates through the parent tube would be slightly less than 
assumed for the SGTR analysis and therefore, would result in lower 
total primary fluid mass release to the secondary system. A main 
steam line break or feedwater line break will not cause a SGTR since 
the sleeves are analyzed for a maximum accident differential 
pressure greater than that predicted in the DCPP safety analysis. 
The sleeve-tube assembly leakage during plant operation would be 
minimal and is well within the Technical Specification (TS) leakage 
limits.
    The proposed change to TS 5.5.9 Table 5.5.9-2, ``Steam Generator 
(SG) Tube Inspection,'' to delete the requirement to notify the NRC 
pursuant to 10 CFR 50.72(b)(2) if the first sample inspection or the 
second sample inspection results in a C-3 classification, is an 
administrative change only and does not affect plant equipment or 
accident analyses.
    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 leak limiting Alloy 800 sleeves are designed using the 
applicable ASME Code as guidance, and therefore meet the objectives 
of the original SG tubing. As a result, the functions of the SG will 
not be significantly affected by the installation of the proposed 
sleeve. The proposed sleeves do not interact with any other plant 
systems. Any accident as a result of potential tube or sleeve 
degradation in the repaired portion of the tube is bounded by the 
existing SGTR accident analysis. The continued integrity of the 
installed sleeve-tube assembly is periodically verified by the TS 
requirements and a sleeved tube will be plugged on detection of an 
imperfection in the sleeve or in the pressure boundary portion of 
the original tube wall in the leak limiting sleeve/tube assembly.
    Implementation of the proposed change has no significant effect 
on either the configuration of the plant, or the manner in which it 
is operated. The proposed change to delete the requirement to notify 
the NRC pursuant to 10 CFR 50.72(b)(2) from TS 5.5.9 Table 5.5.9-2 
is an administrative change only and does not affect plant equipment 
or accident analyses.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The repair of degraded SG tubes with leak limiting Alloy 800 
sleeves restores the structural integrity of the degraded tube under 
normal operating and postulated accident conditions and thereby 
maintains current core cooling margin as opposed to plugging the 
tube and taking it out of service. The design safety factors 
utilized for the sleeves are consistent with the safety factors in 
the ASME Boiler and Pressure Vessel Code used in the original SG 
design. The sleeve and portions of the installed sleeve-tube 
assembly that represent the reactor coolant pressure boundary will 
be monitored and a sleeved tube will be plugged on detection of an 
imperfection in the sleeve or in the pressure boundary portion of 
the original tube wall in the leak limiting sleeve/tube assembly. 
Use of the previously identified design criteria and design 
verification testing assures that the margin to safety is not 
significantly different from the original SG tubes.
    The proposed change to delete the requirement to notify the NRC 
pursuant to 10 CFR 50.72(b)(2) from TS 5.5.9 Table 5.5.9-2 is an 
administrative change only, does not affect plant equipment or 
accident analyses, does not relax any safety system settings, and 
does not relax the bases for any limiting conditions for operations.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

    STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas.
    Date of amendment request: November 4, 2003.
    Description of amendment request: The proposed amendments would 
revise the South Texas Project, Units 1 and 2 Technical Specifications 
for the Remote Shutdown System to reflect requirements consistent with 
those in NUREG-1431, ``Standard Technical Specifications--Westinghouse 
Plants.''

[[Page 66141]]

The proposed changes would increase the allowed outage time for 
inoperable Remote Shutdown System components to a time that is more 
consistent with their safety significance. It would also relocate the 
description of the required components to the Bases where it will be 
directly controlled by the licensee.
    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.
    Because the proposed changes do not involve potential accident 
initiators, there is no significant increase in the probability of 
an accident previously evaluated. There is no proposed change to the 
design basis or configuration of the plant and the extension of the 
allowed outage time of the Remote Shutdown System functions does not 
have a significant effect on safety. Consequently there is no 
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?
    Response: No.
    The proposed changes do not affect how the plant is operated or 
involve any physical changes to the plant. Therefore there is no 
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.
    Except for extending the allowed outage time for Remote Shutdown 
System function from 7 days to 30 days, the proposed changes are 
essentially administrative. The evaluation of the extension of the 
allowed outage time demonstrated that there was no 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 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: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 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 e-mail to [email protected].
    Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan.
    Date of application for amendment: June 24, 2003.
    Brief description of amendment: The amendment revises Technical 
Specification 3.1.8, ``Scram Discharge Volume (SDV) Vent and Drain 
Valves,'' to allow a vent or drain line with one inoperable valve to be 
isolated instead of requiring the valve to be restored to Operable 
status within 7 days.
    Date of issuance: October 30, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 157.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 19, 2003 (68 FR 
49815).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 30, 2003.
    No significant hazards consideration comments received: No.

    Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina.
    Date of application of amendments: July 10, 2003.
    Brief description of amendments: The amendments revised the 
Technical Specifications to remove requirements that are no longer 
applicable because the implementation of the automatic feedwater 
isolation system modification has been completed on all three Oconee 
units.
    Date of Issuance: November 5, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 336, 336, & 337.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 19, 2003 (68 FR 
49816). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 5, 2003.
    No significant hazards consideration comments received: No.
    Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station,
    Unit 1, Claiborne County, Mississippi.
    Date of application for amendment: April 3, 2003.
    Brief description of amendment: The changes revise the Updated 
Final Safety Analysis Report to change the Reactor Vessel Material 
Surveillance Program. The change reflects participation in the Boiling 
Water Reactor Vessel and Internals Project Integrated Surveillance 
Program.
    Date of issuance: November 4, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.

[[Page 66142]]

    Amendment No: 160.
    Facility Operating License No. NPF-29: The amendment revises the 
Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: May 13, 2003 (68 FR 
25653).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 4, 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: October 23, 2001, as 
supplemented on March 29 and December 17, 2002, and June 12, 2003.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 5.5.10, ``Ventilation Filter Testing Program,'' to 
adopt the requirements of the American Society for Testing and 
Materials Standard D3803-1989, ``Standard Test Method for Nuclear-Grade 
Activated Carbon.'' The TS revisions are in response to Nuclear 
Regulatory Commission (NRC) Generic Letter (GL) 99-02, ``Laboratory 
Testing of Nuclear-Grade Activated Charcoal.'' The amendment revises 
the TSs: (1) To provide a control room ventilation system (CRVS) methyl 
iodide removal efficiency of greater than or equal to 95.5% and remove 
the notation that there is a 1-inch charcoal bed depth; (2) to allow 
for the continued use of the existing CRVS through Refueling Outage 13, 
in order to design, fabricate, and install a 2-inch charcoal filter 
bed; and (3) to add a note in the TS requiring a demonstration of 
charcoal efficiency of 93% when changing the charcoal in the existing 
CRVS bed prior to any fuel movement in the upcoming Refueling Outage 12 
and every 6 months thereafter until the new beds are installed. The NRC 
had previously published a notice of consideration on December 12, 2001 
(66 FR 64292) regarding a similar proposal from the licensee in 
response to GL 99-02. However, in response to a request for additional 
information from the NRC dated March 29, 2002, the licensee revised its 
application and withdrew the prior request to change the maximum CRVS 
differential pressure in TS 5.5.10.d.
    Date of issuance: October 30, 2003.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment No.: 219.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 18, 2003 (68 FR 
12951).
    The March 29 and December 17, 2002, and June 12, 2003, letters 
provided clarifying information that did not enlarge the scope of the 
amendment request or change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 30, 2003.
    No significant hazards consideration comments received: No.

    Entergy Operations, Inc., Docket No. 50-382, Waterford Steam 
Electric Station, Unit 3, St. Charles Parish, Louisiana.
    Date of amendment request: December 16, 2002, as supplemented by 
letters dated July 30, and September 29, 2003.
    Brief description of amendment: The amendment adds Combustion 
Engineering topical report CEN-372-P-A, May 1990, ``Fuel Rod Maximum 
Allowable Gas Pressure,'' to the list of topical reports in Technical 
Specification 6.9.1.11.1, used to determine the Waterford Steam 
Electric Sation, Unit 3 core operating limits. In addition, the 
amendment approves the deletion of applicable dates and revision 
numbers for CEN-372-P-A and other topical reports listed in TS 
6.9.1.11.1.
    Date of issuance: October 31, 2003.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 191.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5673). The July 30, and September 29, 2003, supplemental letters 
provided clarifying information that did not change the scope of the 
original Federal Register notice or the original no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 31, 2003.
    No significant hazards consideration comments received: No.
    Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, 
LaSalle County Station, Units 1 and 2, LaSalle County, Illinois.
    Date of application for amendments: March 31, 2003.
    Brief description of amendments: The amendments revise Appendix A, 
Technical Specifications (TS), of Facility Operating License Nos. NPF-
11 and NPF-18. Specifically, the changes 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. The revision also 
includes administrative changes regarding access control and 
terminology for high radiation areas.
    Date of issuance: October 31, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 161/147.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28852).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 31, 2003.
    No significant hazards consideration comments received: No.

    Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353.
    Limerick Generating Station, Units 1 and 2, Montgomery County, 
Pennsylvania.
    Date of application for amendments: December 20, 2002, as 
supplemented May 30, 2003.
    Brief description of amendments: The amendments removed the current 
facility reactor material specimen surveillance schedule from the 
Technical Specifications for Limerick Generating Station, Units 1 and 2 
(LGS-1 and 2). The licensee also revised the Updated Final Safety 
Analysis Report (UFSAR) for LGS-1 and 2 to reflect implementation of 
the Boiling Water Reactor Vessel and Internals Project reactor pressure 
vessel integrated surveillance program as the basis for demonstrating 
the compliance with the requirements of Appendix H, ``Reactor Vessel 
Material Surveillance Program Requirements,'' to title 10 of the Code 
of Federal Regulations, Part 50.
    Date of issuance: November 4, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 167 and 130.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments 
revised the Technical Specifications and authorized changes to the 
UFSAR for LGS-1 and 2.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5669). The supplement dated May 30, 2003, provided additional 
information that

[[Page 66143]]

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 
Evaluationdated November 4, 2003.
    No significant hazards consideration comments received: No.

    Exelon Generation Company, LLC, and PSEG Nuclear LLC,

    Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station,
    Units 2 and 3, (PBAPS-2 and 3) York County and Lancaster County, 
Pennsylvania.
    Date of application for amendments: December 20, 2002, as 
supplemented May 30, 2003.
    Brief description of amendments: The amendments revised the Updated 
Final Safety Analysis Report (UFSAR) for Peach Bottom Atomic Power 
Station, Units 2 and 3, by allowing implementation of the Boiling Water 
Reactor Vessel and Internals Project reactor pressure vessel integrated 
surveillance program as the basis for demonstrating the compliance with 
the requirements of Appendix H, ``Reactor Vessel Material Surveillance 
Program Requirements,'' to Title 10 of the Code of Federal Regulations, 
Part 50.
    Date of issuance: November 4, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 249 and 253.
    Renewed Facility Operating License Nos. DPR-44 and DPR-56: The 
amendments authorized changes to the UFSAR for PBAPS-2 and 3.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5669). The supplement dated May 30, 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 November 4, 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: March 27, 2003, as supplemented 
August 15, 2003.
    Brief description of amendment: The amendment lowers the trip 
setpoint and allowable value contained in Technical Specification (TS) 
Table 3.3-4 for the pressurizer pressure low safety injection signal. 
The amendment also lowers the value for the P-11 setpoint in TS Table 
3.3-3. These changes increase the margin between the low pressurizer 
pressure safety injection actuation setpoint and the minimum 
pressurizer pressure that occurs immediately following a reactor trip.
    Date of issuance: November 12, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 263.
    Facility Operating License No. DPR-74: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28853).
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 12, 2003.
    No significant hazards consideration comments received: No.

    Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska.
    Date of amendment request: December 31, 2002, as supplemented by 
letter dated July 24, 2003.
    Brief description of amendment: The amendment revises the Updated 
Safety Analysis Report (USAR) reflecting a change of the reactor vessel 
material surveillance program to incorporate the Boiling Water Reactor 
Vessel and Internals Project Integrated Surveillance Program into the 
licensing basis.
    Date of issuance: October 31, 2003.
    Effective date: As of the date of issuance. The amendment shall be 
implemented within 30 days of issuance and the USAR changes shall be 
implemented in the next periodic update to the USAR in accordance with 
10 CFR 50.71(e).
    Amendment No.: 201.
    Facility Operating License No. DPR-46: Amendment revised the USAR.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5678).
    The July 24, 2003, supplemental letter provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register on February 4, 2003 (68 FR 5678).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 31, 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, as supplemented by 
letter dated August 1, 2003.
    Brief description of amendment: The amendment authorizes revisions 
to the Updated Safety Analysis Report (USAR) to incorporate the NRC 
approval of the GOTHIC 7.0 computer program for performing containment 
analyses.
    Date of issuance: November 5, 2003.
    Effective date: November 5, 2003, and shall be implemented within 
30 days of the date of issuance. The implementation of the amendment 
includes the incorporation into the USAR the changes discussed above, 
as described in the licensee's application dated January 27, 2003, and 
supplement dated August 1, 2003, and evaluated in the staff's Safety 
Evaluation attached to the amendment.
    Amendment No.: 222.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the USAR.
    Date of initial notice in Federal Register: March 18, 2003 (68 FR 
12956).
    The August 1, 2003, supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noticed, and did not change the staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 5, 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, as supplemented by 
letter dated October 14, 2003.
    Brief description of amendment: The amendment deletes Technical 
Specification (TS) 2.3(2)i and the corresponding Bases that allows the 
performance of the surveillance test of Table 3-2, Item 20 
(Recirculation Actuation Logic Channel Functional Test) under 
administrative controls, while components in excess of those allowed by 
Conditions a, b, d, and e of TS 2.3(2) are inoperable, provided they 
are returned to operable status within one hour. This allowance was 
granted in Amendment No. 206 issued April 19, 2002, and only applied 
until the end of Cycle 21.

[[Page 66144]]

    Date of issuance: November 10, 2003.
    Effective date: November 10, 2003, and shall be implemented within 
60 days from the date of issuance.
    Amendment No.: 223.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 18, 2003 (68 FR 
12955).
    The October 14, 2003, supplemental letter provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 10, 2003.
    No significant hazards consideration comments received: No.

    PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna 
Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania.
    Date of application for amendments: May 6, 2003, as supplemented by 
letters dated August 12 and September 18, 2003.
    Brief description of amendments: These amendments deleted Technical 
Specification (TS) 3.3.1.3, ``Oscillation Power Range Monitor (OPRM) 
Instrumentation,'' and revised TS 3.4.1, ``Recirculation Loops 
Operating,'' to formally extend the currently implemented requirements, 
which define appropriately conservative restrictions to plant operation 
and operator response to thermal hydraulic instability events. In 
addition, the amendments revise TS 3.4.1 to refer to the power flow map 
in the core operating limits report and include a reference in TS 
5.6.5.
    Date of issuance: October 29, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 215 and 190.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 24, 2003 (68 FR 
37582).
    The supplemental letters dated August 12 and September 18, 2003, 
provided clarifying information that did not change the scope of the 
amendment as described in the initial notice of the proposed action 
published in the Federal Register notice (68 FR 37582, June 24, 2003), 
or the U.S. Nuclear Regulatory Commission staff's proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 29, 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: July 25, 2003.
    Description of amendment request: The amendments revised Technical 
Specification 3.1.8, ``Scram Discharge Volume (SDV) Vent and Drain 
Valves,'' to allow a vent or drain line with one inoperable valve to be 
isolated instead of requiring the valve to be restored to operable 
status within 7 days.
    Date of issuance: November 3, 2003.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment Nos.: 248, 285, and 243.
    Facility Operating License Nos. DPR-33, DPR-52, and DPR-68. 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 18, 2003 (68 
FR 54753).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation.
    No significant hazards consideration comments received: No.


    Dated at Rockville, Maryland, this 17th day of November, 2003.

    For the Nuclear Regulatory Commission.
Eric Leeds,
Deputy Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 03-29107 Filed 11-24-03; 8:45 am]
BILLING CODE 7590-01-P