[Federal Register Volume 68, Number 82 (Tuesday, April 29, 2003)]
[Notices]
[Pages 22744-22759]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-10396]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from April 18, 2003, through May 1, 2003. The 
last biweekly notice was published on April 15, 2003, (68 FR 18269).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the 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 May 29, 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

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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.
    Non-timely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, 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].

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

    Date of amendment request: March 20, 2003.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) and the licensing basis in the 
Updated Safety Analysis Report (UFSAR) to support installation of a 
passive low-pressure injection (LPI) cross connect inside containment. 
The proposed changes to the TS would add requirements for the passive 
LPI cross connect and eliminate requirements associated with the 
capability to cross connect by manual operator action the trains 
outside containment. The proposed changes to the UFSAR would revise the 
licensing basis for a portion of the core flood and LPI/Decay Heat 
Removal (DHR) piping to allow the exclusion of dynamic effects 
associated with postulated pipe rupture of that piping by application 
of leak-before-break technology for Unit 1. The proposed changes to the 
UFSAR would also revise the licensing basis for selected portions of 
the LPI/DHR piping to adopt the design requirements of Standard Review 
Plan Section 3.6.2, Branch Technical Position MEB 3-1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards

[[Page 22746]]

consideration, which is presented below: Pursuant to 10 CFR 50.91, Duke 
Power Company (Duke) has made the determination that this amendment 
request involves a No Significant Hazards Consideration by applying the 
standards established by the NRC regulations in 10 CFR 50.92. This 
ensures that operation of the facility in accordance with the proposed 
amendment would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated: The proposed LAR 
[Licence Amendment Request] modifies the Technical Specifications 
[(TS)] to incorporate new TS requirements associated with the new Low 
Pressure Injection (LPI) System configuration and eliminate TS 
requirements associated with the old LPI configuration. The proposed 
LAR also modifies the licensing basis to adopt Standard Review Plan 
(SRP) 3.6.2 Branch Technical Position (BTP) MEB 3-1 requirements for 
selected portions of LPI piping and to credit Leak-Before-Break (LBB) 
to allow the dynamic effects associated with postulated pipe rupture of 
selected portions of the LPI/Core Flood (CF) piping to be excluded from 
the design basis. The proposed design allowances for these selected 
portions of piping continue to allow the LPI system design to meet GDC 
[General Design Criterion] 4 requirements related to environmental and 
dynamic effects. The proposed LAR will continue to ensure that ONS 
[Oconee Nuclear Station] can meet design basis requirements associated 
with the LPI safety function. The LPI System provides a means for 
delivering a large volume of borated water to the reactor core 
following postulated large pipe breaks in the Reactor Coolant System. 
The planned modification adds a passive crossover connection between 
the two LPI injection lines inside containment, along with necessary 
check valves and flow orifices that will eliminate the need for time-
critical operator actions to manually open the LPI discharge header 
outside containment. The new components will have the same pressure, 
seismic, and quality group qualifications as the existing components in 
the LPI system. The addition of the crossover line will enhance the 
ability of the control room operator to mitigate the consequences of 
specific events for which LPI is credited. Therefore, the proposed LAR 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The LPI system is also relied on to cool the reactor core during 
unit shutdown. Hydraulic analyses have demonstrated that adequate LPI 
flow is available for normal shutdown cooling with the new LPI piping 
configuration.
    2. Create the possibility of a new or different kind of accident 
from any kind of accident previously evaluated: The proposed LAR 
modifies the Technical Specification to incorporate new TS requirements 
associated with the new LPI System configuration and eliminate TS 
requirements associated with the old LPI System configuration. The 
proposed LAR also modifies the licensing basis to adopt MEB 3-1 
requirement for selected portions of LPI piping and to credit LBB to 
allow the dynamic effects associated with postulated pipe rupture of 
selected portions of the LPI/Core Flood (CF) piping to be excluded from 
the design basis. The proposed design allowances for these selected 
portions of piping continue to allow the LPI system design to meet GDC 
4 requirements related to environmental and dynamic effects. The LPI 
and Core Flood systems affected by implementing the proposed changes to 
the TS are not assumed to initiate design basis accidents. The systems 
affected by the changes are used to mitigate the consequences of an 
accident that has already occurred. The proposed TS and licensing basis 
changes do not affect the mitigating function of these systems. 
Consequently, these changes do not alter the nature of events 
postulated in the Safety Analysis Report nor do they introduce any 
unique precursor mechanisms. Therefore, the proposed amendment will not 
create the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed TS and licensing basis changes do not unfavorably 
affect any plant safety limits, set points, or design parameters. The 
changes also do not unfavorably affect the fuel, fuel cladding, RCS, or 
containment integrity. Therefore, the proposed TS and licensing basis 
changes, which adds TS requirements and adopts new design allowances 
associated with the passive LPI cross connect modification, do not 
involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: John A. Nakoski.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
    Date of amendment request: March 19, 2003.
    Description of amendment request: The proposed amendment deletes 
requirements from the technical specifications (TS) and other elements 
of the licensing bases to maintain a Post Accident Sampling System 
(PASS). Licensees were generally required to implement PASS upgrades as 
described in NUREG-0737, ``Clarification of TMI [Three Mile Island 
Nuclear Station] Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2 (TMI-2). Requirements 
related to PASS were imposed by Order for many facilities and were 
added to or included in the TS for nuclear power reactors currently 
licensed to operate. Lessons learned and improvements implemented over 
the last 20 years have shown that the information obtained from PASS 
can be readily obtained through other means or is of little use in the 
assessment and mitigation of accident conditions.
    The changes are based on NRC-approved Technical Specification Task 
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
413, ``Elimination of Requirements for a Post Accident Sampling System 
(PASS).'' The U.S. Nuclear Regulatory Commission (NRC) staff issued a 
notice of opportunity for comment in the Federal Register on December 
27, 2001 (66 FR 66949), on possible amendments concerning TSTF-413, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on March 20, 2002 (67 FR 13027). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated March 19, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an

[[Page 22747]]

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.
    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result of 
the TMI-2 accident. The specific intent of the PASS was to provide a 
system that has the capability to obtain and analyze samples of plant 
fluids containing potentially high levels of radioactivity, without 
exceeding plant personnel radiation exposure limits. Analytical results 
of these samples would be used largely for verification purposes in 
aiding the plant staff in assessing the extent of core damage and 
subsequent offsite radiological dose projections. The system was not 
intended to and does not serve a function for preventing accidents and 
its elimination would not affect the probability of accidents 
previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual benefit 
to post accident mitigation. Past experience has indicated that there 
exists in-plant instrumentation and methodologies available in lieu of 
a PASS for collecting and assimilating information needed to assess 
core damage following an accident. Furthermore, the implementation of 
Severe Accident Management Guidance (SAMG) emphasizes accident 
management strategies based on in-plant instruments. These strategies 
provide guidance to the plant staff for mitigation and recovery from a 
severe accident. Based on current severe accident management strategies 
and guidelines, it is determined that the PASS provides little benefit 
to the plant staff in coping with an accident.
    The regulatory requirements for the PASS can be eliminated without 
degrading the plant emergency response. The emergency response, in this 
sense, refers to the methodologies used in ascertaining the condition 
of the reactor core, mitigating the consequences of an accident, 
assessing and projecting offsite releases of radioactivity, and 
establishing protective action recommendations to be communicated to 
offsite authorities. The elimination of the PASS will not prevent an 
accident management strategy that meets the initial intent of the post-
TMI-2 accident guidance through the use of the SAMGs, the emergency 
plan (EP), the emergency operating procedures (EOP), and site survey 
monitoring that support modification of emergency plan protective 
action recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any accident 
previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident from Any Previously Evaluated
    The elimination of PASS related requirements will not result in any 
failure mode not previously analyzed. The PASS was intended to allow 
for verification of the extent of reactor core damage and also to 
provide an input to offsite dose projection calculations. The PASS is 
not considered an accident precursor, nor does its existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radioisotopes within the 
containment building.
    Therefore, this change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety
    The elimination of the PASS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a neutral 
impact to the margin of safety. Methodologies that are not reliant on 
PASS are designed to provide rapid assessment of current reactor core 
conditions and the direction of degradation while effectively 
responding to the event in order to mitigate the consequences of the 
accident. The use of a PASS is redundant and does not provide quick 
recognition of core events or rapid response to events in progress. The 
intent of the requirements established as a result of the TMI-2 
accident can be adequately met without reliance on a PASS.
    Therefore, this change does not involve a significant reduction in 
the margin of safety.
    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    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: March 26, 2003.
    Description of amendment request: The amendment request proposes to 
adopt the Boiling Water Reactor Vessel and Internals Project integrated 
surveillance program (BWRVIP ISP) as the basis for demonstrating 
compliance with the requirements of Appendix H to Title 10 of the Code 
of Federal Regulations Part 50 (10 CFR 50), ``Reactor Vessel Material 
Surveillance Program Requirements'' and delete Technical Specification 
(TS) 4.6.A.5. The licensee also proposes to update the pressure-
temperature (P-T) curves through the end of the current operating 
license by revising TS Figures 3.6.1, 3.6.2, and 3.6.3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff's review is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Brittle fracture of the reactor pressure vessel (RPV) is not a 
postulated or evaluated design basis accident. No evaluations of other 
postulated accidents are affected by this proposed change. Because the 
applicable regulatory requirements continue to be met, the change does 
not significantly increase the probability of any accident previously 
evaluated.
    Also, the change will not alter any assumptions previously made in 
evaluating the radiological consequences of accidents.
    Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Create the possibility for a new or different kind of accident 
from any previously evaluated.
    The proposed change does not involve a modification of the design 
of plant structures, systems, or components. The change will not impact 
the manner in which the plant is operated and will not degrade the 
reliability of structures, systems, or components important to safety 
as equipment protection features will not be deleted or modified, 
equipment redundancy or independence will not

[[Page 22748]]

be reduced, supporting system performance will not be affected, and no 
severe testing of equipment will be imposed. No new failure modes or 
mechanisms will be introduced as a result of this proposed change.
    Therefore, the changes to the material surveillance program and 
pressure-temperature limits that compose this proposed change do not 
create the possibility of a new or different kind of accident than 
those previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. There is no change or impact on any safety 
analysis assumption or in any other parameter affecting the course of 
an accident analysis supporting the Bases of any Technical 
Specification. The proposed change does not involve any increase in 
calculated off-site dose consequences.
    Therefore, this change does not involve a significant reduction in 
a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire
    Date of amendment request: February 3, 2003.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) 3/4.7.1.4, ``Turbine Cycle--
Specific Activity,'' and its associated bases. With the exception of TS 
4.0.4, wording similar to that presented in the improved Standard 
Technical Specifications will be adopted. The amendment request 
proposes an exception to the requirements of TS 4.0.4 when entering 
MODE 4, along with conditions for when the surveillance requirement 
must be satisfied in MODE 4. Additionally there are editorial changes 
to the TS Index reflecting the proposed revision.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes, in part, modify the modes of applicability by 
stating that TS 4.0.4 is not applicable for Mode 4 entry. For the 
surveillance requirement, the change specifies the conditions in Mode 4 
that are necessary to obtain a representative sample from the steam 
generators. Analyzed events are assumed to be initiated by the failure 
of plant structures, systems or components. The level of specific 
activity contained in the reactor coolant is germane to the 
consequences of an accident and is not related in any way to the 
probability of failure of a plant structure, system or component which 
would result in the occurrence of an unanalyzed event. Because the 
probability of failure of plant equipment is not affected, there is no 
impact on the probability of occurrence of a previously analyzed 
accident.
    The consequences of a previously analyzed event are dependent on 
the initial conditions assumed for the analysis, and the availability 
and successful functioning of the equipment assumed to operate in 
response to the analyzed event. The proposed changes do not alter the 
initial conditions assumed in the analysis of interest. The plant 
parameters assumed for the analyses are maintained within assumed 
limits through compliance with the Technical Specifications and plant 
procedures. Additionally, the proposed changes do not impose any new 
safety analyses limits. Any deviation from the allowable activity 
limits will require the plant to be placed in a condition where the 
specification does not apply. Therefore, the proposed changes do not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The proposed changes do not involve a physical alteration of the 
plant. No new equipment is being introduced, and installed equipment is 
not being operated in a new or different manner. There is no change 
being made to the parameters within which the plant is operated, or to 
the setpoints at which protective or mitigative actions are initiated. 
No alteration in the procedures that ensure the plant remains within 
analyzed limits is being proposed, and no change is being made to the 
procedures relied upon to respond to an off-normal event. As such, no 
new failure modes are being introduced. These changes have no physical 
effect on any plant equipment. Therefore, the changes do not create the 
possibility of a new of different kind of accident from any previously 
evaluated.
    3. The proposed changes do not involve a significant reduction in a 
margin of safety.
    The margin of safety is established through equipment design, 
limitations on operating parameters, and the setpoints at which 
automatic actions are initiated. No equipment design features are 
impacted by these changes, no operating parameters are revised, and no 
changes are proposed to the actuation setpoints. The limit on secondary 
coolant Dose Equivalent Iodine remains at the current value of 0.1 
microcuries per gram. Therefore, the proposed changes do not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis, and based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. S. Ross, Florida Power & Light Company, 
PO Box 14000, Juno Beach, FL 33408-0420.
    NRC Section Chief: James W. Clifford.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station, Unit 1 (NMP1), Oswego County, New York
    Date of amendment request: October 7, 2002, as supplemented on 
March 24, 2003.
    Description of amendment request: The licensee's October 7, 2002, 
application proposed to add Specification 4.0.3 to address missed 
surveillances to the Technical Specifications (TSs). This new 
specification specifies an initial 24-hour delay period for performing 
a missed surveillance prescribed by Specification 3.0.3. Specification 
4.0.3 will also require: ``A risk evaluation shall be performed for any 
surveillance delayed greater than 24 hours and the risk impact shall be 
managed.'' In addition, the licensee proposed to add wording to each of 
the following existing specifications such that the new Specification 
4.0.3 would apply to them: Specification 6.16, 6.17, 6.18, and 6.19. On 
November 12, 2002, the Nuclear Regulatory Commission (NRC) staff 
published a proposed no significant hazards consideration determination 
and opportunity for a

[[Page 22749]]

hearing (67 FR 68739) for the October 7, 2002, application.
    As a result of the NRC staff comments, the licensee supplemented 
the application by a letter dated March 24, 2003. The supplement adds 
new requirements related to the use and application of the surveillance 
requirements (SRs) currently included in the TSs.
    These new explicit SR applicability requirements would supersede 
the more general current requirements. The proposed new requirements 
reflect the current practices at NMP1, and as such, do not change any 
existing method of plant operation.
    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 for the March 24, 2003, supplement, 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.
    Adoption of new administrative requirements related to the proper 
use of the surveillance requirements currently included in the NMP1 TSs 
do not affect any accident initiator, and as such, will have no effect 
on the probability of an accident. The proposed changes do not involve 
physical changes to the plant or introduce any new modes of operation. 
Accordingly, continued assurance is provided that the process 
variables, structures, systems, and components are maintained such that 
there will be no degradation of any fission product barrier which could 
increase the radiological consequences of an accident. Therefore, the 
proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Adoption of new administrative requirements related to the proper 
use of the surveillance requirements currently included in the NMP1 TSs 
will have no adverse effect on the design or assumed accident 
performance of any structure, system, or component, or introduce any 
new modes of system operation or failure modes. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes add new administrative requirements related to 
the proper use of the surveillance requirements currently included in 
the NMP1 TSs. The addition of requirements will make application of the 
surveillance requirements more restrictive than currently required by 
the TSs. Accordingly, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
supplement of March 24, 2003, involves no significant hazards 
consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard J. Laufer.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin
    Date of amendment request: March 27, 2003.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Surveillance Requirement 3.1.4.1, ``Rod 
Group Alignment Limits, to change the allowable alignment limits of 
individual rods in Mode 1 when greater than 85-percent power.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant increase in 
the probability or consequences of any accident previously evaluated.
    This proposed change does not cause an increase in the 
probabilities of any accidents previously evaluated because the change 
will not cause an increase in the probability of any initiating events 
for accidents previously evaluated.
    The consequences of the accidents previously evaluated in the PBNP 
[Point Beach Nuclear Plant] Final Safety Analysis Report (FSAR) are 
determined by the results of analyses that are based on initial 
conditions of the plant, the type of accident, transient response of 
the plant, and the operation and failure of equipment and systems.
    Based on the analyses documented in WCAP-15432, Revision 2 
[``Conditional Extension of the Rod Misalignment Technical 
Specification for Point Beach Units 1 and 2, (proprietary)'' dated 
April 2001], all pertinent licensing-basis acceptance criteria have 
been met and the margin of safety, as defined in the Technical 
Specification Bases, is not significantly reduced in any of the Point 
Beach licensing basis accident analyses due to the subject change. 
Therefore, the probability of an accident previously evaluated has not 
significantly increased. Because design limitations continue to be met 
and the integrity of the reactor coolant system pressure boundary is 
not challenged, the assumptions employed in the calculation of the 
offsite radiological doses remain valid. Neither rod position 
indication nor the limits on allowed rod position deviation is an 
accident initiator or precursor. Therefore, the consequences of an 
accident previously evaluated will not be significantly increased.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a new or different kind of 
accident from any accident previously evaluated.
    The changes described in the proposed amendment are supported by 
the analyses provided in the submittal [the March 27, 2003, 
application]. The evaluation of the effects of the proposed changes 
indicates that all design standards and applicable safety criteria 
limits are met. These changes therefore do not cause the initiation of 
any new or different accident nor create any new failure mechanisms.
    Equipment important to safety will continue to operate as designed. 
The proposed change does not result in any event previously deemed 
incredible being made credible. The change does not result in more 
adverse conditions or result in any increase in the challenges to 
safety systems. Therefore, operation of the Point Beach Nuclear Plant 
in accordance with the proposed amendment will not create the 
possibility of a new or different type of accident from any accident 
previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant reduction in a 
margin of safety.
    Based on the analyses documented in WCAP-15432, Revision 2, all 
pertinent licensing-basis acceptance criteria have been met and the 
margin of safety, as defined in the Technical Specification

[[Page 22750]]

Bases, is not significantly reduced in any of the Point Beach licensing 
basis accident analyses based on the subject changes to safety analyses 
input parameter values. There are no new or significant changes to the 
initial conditions contributing to accident severity or consequences. 
Since the analyses in the accompanying submittals [March 27, 2003, 
application and WCAP-15432] demonstrate that all applicable acceptance 
criteria continue to be met, the subject operating conditions will not 
involve a significant reduction in a margin of safety at Point Beach.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota
    Date of amendment request: March 25, 2003.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.2.1 and TS 3.2.3 for 
implementation of relaxed axial offset control of the reactor cores, 
relocate selected operating parameters from TS 2.0 and TS 3.3.1 to the 
Core Operating Limits Report (COLR), revise the Pressurizer Pressure-
Low Allowable Value, and revise the appropriate references in TS 5.6.5 
to the NRC-approved methodologies which support relocation of operating 
parameters to the COLR.
    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:

Group 1--Implementation of Relaxed Axial Offset Control

    A. TS 3.2.1, Heat Flux Hot Channel Factor--FQ(Z) and 
Bases: Modification of Required Actions and Completion Time if 
FWQ(Z) is not within its limit and update Bases.
    B. TS 3.2.3, Axial Flux Difference (AFD) and Bases: Modification 
of Limiting Conditions for Operation, Actions and Surveillance 
Requirements and revision of the Bases.
    This license amendment request proposes to revise the Technical 
Specifications to implement the relaxed axial offset control 
methodology to address the heat flux hot channel factor and axial 
flux difference limits.
    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This license amendment request proposes to revise the Technical 
Specifications to implement the relaxed axial offset control 
methodology to address the heat flux hot channel factor and axial 
flux difference limits. The revised Technical Specifications and 
parameter changes associated with relaxed axial offset control 
assure that the limiting safety analysis inputs (such as, heat flux 
hot channel factor and axial flux difference limits) are not 
exceeded. The bounding power distribution transient factor values, 
W(Z), and the axial flux difference limits that are documented in 
the Core Operating Limits Report will be determined by NRC approved 
analytical methods and will be validated as part of the cycle 
specific reload evaluation process.
    Heat flux hot channel factors and axial flux difference limits 
are not assumed accident initiators. Therefore, the relaxed axial 
offset control related Technical Specification changes do not 
involve a significant increase in the probability of an accident.
    Likewise, operation of the plant within the proposed Technical 
Specification controls and limits assures that safety analysis 
assumptions are met, thus, if an accident were to occur, the 
consequences would continue to be bounded by the accident analyses. 
Therefore, the relaxed axial offset control related technical 
specification changes do not involve a significant increase in the 
consequences of an accident.
    The relaxed axial offset control related technical specification 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    This proposed change does not involve a physical alteration of 
the plant; that is, no new or different type of equipment will be 
installed. This proposed change does not introduce any new mode of 
plant operation or change the methods governing normal plant 
operation. No new failure mode has been created and no new equipment 
performance burdens are imposed. Therefore the possibility of a new 
or different kind of accident from those previously analyzed has not 
been created.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    This license amendment request proposes to revise the Technical 
Specifications to implement the relaxed axial offset control 
methodology to address the heat flux hot channel factor and axial 
flux difference limits. The supporting Technical Specification 
limits are defined by NRC approved analytical methods which are 
performed to conservatively bound the operating conditions defined 
by the Technical Specifications and to demonstrate meeting the 
regulatory acceptance limits. The heat flux hot channel factor 
licensed safety margins are maintained. The heat flux hot channel 
factor conforms to plant design bases and limits actual plant 
operation within analyzed and licensed boundaries. The relaxed axial 
offset control methodology has been demonstrated to ensure that core 
heat flux hot channel factors will remain below accident analysis 
limits. The margin of safety provided by the analyses in accordance 
with the acceptance limits is maintained and not reduced. Thus, the 
implementation of relaxed axial offset control at Prairie Island 
does not involve a significant reduction in a margin of safety.

Group 2--Relocation of Technical Specifications Safety Limits Figure 
and Overtemperature Delta-T and Overpower Delta-T Parameter Values to 
the Core Operating Limits Report, and Miscellaneous Administrative 
Changes

    A. TS 2.1.1, ``Reactor Core SLs [Safety Limits]'' and Bases: 
Relocate the safety limits Figure to the Core Operating Limits 
Report, update TS 2.1.1 and Bases.
    B. TS 3.3.1, Table 3.3.1-1 (Pages 2, 7 and 8), ``Reactor Trip 
System Instrumentation'', Overpower Delta-T Trip Function, and 
Overtemperature Delta-T and Overpower Delta-T parameter values: 
Delete SR [Surveillance Requirement] 3.3.1.3, SR 3.3.1.6, and remove 
f(DI) from Overpower Delta-T Trip Function, relocate overtemperature 
delta-T and overpower delta-T parameter values and revise the Bases.
    C. TS 5.6.5, Core Operating Limits Report (COLR): Additions to 
document Technical Specifications with limits in the Core Operating 
Limits Report and the analytical methods used to determine the 
values for relocated safety limits and overtemperature delta-T and 
overpower delta-T parameters and miscellaneous administrative 
changes.
    This license amendment request proposes to relocate the safety 
limits and overtemperature delta-T and overpower delta-T parameter 
values to the Core Operating Limits Report. Relocation of these 
limits and parameter values to the Core Operating Limits Report 
allows them to be changed under licensee controls. This license 
amendment also proposes to include, in the Technical Specifications 
administrative controls section, the appropriate references to the 
NRC approved methodologies which will be used to determine the 
safety limits and overtemperature delta-T and overpower delta-T 
parameter values. These changes are acceptable because the values 
used to operate the Prairie Island plant will be determined using 
NRC approved methods and these changes are consistent with the 
guidance of the industry standard Technical Specifications, NUREG-
1431, Revision 2, ``Standard Technical Specifications Westinghouse 
Plants''. This license amendment request also proposes to delete 
references to an NRC Safety Evaluation and make some editorial 
corrections in the Technical Specifications administrative controls 
section. These changes are

[[Page 22751]]

acceptable since they are administrative and do not affect plant 
operation.
    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This license amendment request proposes to relocate the safety 
limits and overtemperature delta-T and overpower delta-T parameter 
values to the Core Operating Limits Report and to include, in the 
Technical Specifications administrative controls section, the 
appropriate references to the NRC approved methodologies which 
support determination of these limits and parameter values. The 
safety limits and overtemperature delta-T and overpower delta-T 
parameter values that are documented in the Core Operating Limits 
Report will be determined by NRC approved analytical methods and 
will be validated as part of the cycle specific reload evaluation 
process.
    Safety limits are not assumed accident initiators. Thus 
relocation of the safety limits does not involve a significant 
increase in the probability of an accident. Overtemperature delta-T 
and overpower delta-T parameter values are inputs to the reactor 
trip system which is provided to mitigate the consequences of an 
accident. The reactor trip system is not an accident initiator and 
therefore, changes to input values do not increase the probability 
of an accident.
    Safety limits define bounding values within which plant 
operation will not initiate an accident condition. Safety limits 
relocated to the Core Operating Limits Report and determined by use 
of NRC approved methodologies will continue to determine the safe 
limits of plant operation, thus this change does not involve a 
significant increase in the consequences of an accident. The reactor 
trip system, with inputs from the overtemperature delta-T and 
overpower delta-T trip functions, mitigates the consequences of 
accidents.
    The overtemperature delta-T and overpower delta-T trip parameter 
values are determined to assure that the design limit departure from 
nucleate boiling ratio is met and fuel integrity is maintained. 
Overtemperature delta-T and overpower delta-T trip parameters 
relocated to the Core Operating Limits Report and values determined 
by use of NRC approved methodologies will continue to determine the 
inputs for these trip functions which mitigate the design basis 
accident consequences, thus this change does not involve a 
significant increase in the consequences of an accident.
    Addition of references to NRC approved methodologies in the 
Technical Specifications administrative controls section is an 
administrative change which does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed miscellaneous administrative changes in the 
Technical Specifications administrative controls section do not 
affect plant operation and therefore do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    As discussed above, these proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The malfunction of safety related equipment, assumed to be 
operable in the accident analyses, would not be impacted as a result 
of the proposed technical specification changes. No new failure mode 
has been created and no new equipment performance burdens are 
imposed. Therefore the possibility of a new or different kind of 
accident from those previously analyzed has not been created. The 
proposed administrative changes do not create the possibility of a 
new or different kind of accident from those previously analyzed.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    This license amendment request proposes to relocate the safety 
limits and overtemperature delta-T and overpower delta-T parameter 
values to the Core Operating Limits Report and to include, in the 
Technical Specifications administrative controls section, the 
appropriate references to the NRC approved methodologies which 
support determination of these limits and parameter values. This 
proposed change also allows these relocated limits and parameter 
values to be changed under licensee controls. Safety limits in the 
Core Operating Limits Report will be determined by use of NRC 
approved methodologies and will continue to determine the safe 
limits of plant operation. Overtemperature delta-T and overpower 
delta-T trip parameter values in the Core Operating Limits Report 
will be determined by use of NRC approved methodologies and will 
continue to determine the inputs for these trip functions which 
mitigate design basis accidents. The Safety Limits licensed safety 
margins are maintained. The Safety Limits conform to plant design 
bases and limit actual plant operation within analyzed and licensed 
boundaries. The methodology described in WCAP-8745, along with the 
low pressurizer pressure allowable value, ensures that the 
overtemperature delta-T and overpower delta-T trips will protect 
against fuel centerline melting and departure from nucleate boiling 
during Condition II events. Thus, these changes do not involve a 
significant reduction in the margin of safety.
    This license amendment request proposes to delete references to 
an NRC Safety Evaluation and make some editorial corrections in the 
Technical Specifications administrative controls section. These 
changes are administrative and thus do not involve a significant 
reduction in the margin of safety.

Group 3--Revision of Pressurizer Pressure-Low reactor trip Allowable 
Value

    TS 3.3.1, Table 3.3.1-1 (Page 2), ``Reactor Trip System 
Instrumentation'', Function 8.a, Pressurizer Pressure-Low: Increase 
Pressurizer Pressure-Low Allowable Value.
    This license amendment request proposes to increase the 
Allowable Value defined in Table 3.3.1-1 for the Pressurizer 
Pressure-Low reactor trip.
    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This license amendment request proposes to increase the 
Allowable Value defined in Table 3.3.1-1 for the Pressurizer 
Pressure-Low reactor trip. Pressurizer Pressure-Low reactor trip is 
an input to the reactor trip system which is provided to mitigate 
the consequences of an accident. The reactor trip system is not an 
accident initiator and therefore, changes to the Pressurizer 
Pressure-Low Allowable Value do not involve an increase in the 
probability of an accident.
    The Pressurizer Pressure-Low Allowable Value is being increased 
which is a conservative change. The increase in the Pressurizer 
Pressure-Low reactor trip Allowable Value will assure that the 
overtemperature delta-T and overpower delta-T reactor trip 
functions, with values determined in accordance with NRC approved 
methodologies, provide protection against fuel centerline melting 
and departure from nucleate boiling for overpower and 
overtemperature events. Therefore, this change does not involve an 
increase in the consequences of an accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    This proposed change does not involve a physical alteration of 
the plant; that is, no new or different type of equipment will be 
installed. This proposed change does not introduce any new mode of 
plant operation or change the methods governing normal plant 
operation. No new failure mode has been created and no new equipment 
performance burdens are imposed. Therefore the possibility of a new 
or different kind of accident from those previously analyzed has not 
been created.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    This license amendment request proposes to increase the 
Allowable Value defined in Table 3.3.1-1 for the Pressurizer 
Pressure-Low reactor trip. The Allowable Value is determined in 
accordance with an NRC accepted setpoint methodology with input from 
NRC approved analytical methods. These determinations are performed 
to conservatively bound the operating conditions defined by the 
Technical Specifications and to demonstrate meeting the regulatory 
acceptance limits.
    Performance of analyses and evaluations for the cycle specific 
reload evaluation process will confirm that the operating envelope 
defined by the Technical Specifications continues to be bounded by 
the analytical basis and in no case exceeds the acceptance limits. 
The proposed Pressurizer Pressure-Low Allowable Value along with the 
overtemperature delta-T and overpower delta-T trips will protect 
against fuel centerline melting and departure from nucleate boiling 
during Condition II events. The proposed Allowable Value conforms to 
plant design bases and limits actual plant

[[Page 22752]]

operation within analyzed and licensed boundaries. The margin of 
safety provided by the proposed Pressurizer Pressure-Low Allowable 
Value is maintained and not reduced. Thus, the increase in the 
Pressurizer Pressure-Low reactor trip Allowable Value 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: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: March 3, 2003.
    Description of amendment request: The proposed amendments would 
delete requirements from the technical specifications (TS) and other 
elements of the licensing bases to maintain a Post-Accident Sampling 
System (PASS). Licensees were generally required to implement PASS 
upgrades as described in NUREG-0737, ``Clarification of TMI [Three Mile 
Island] Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI Unit 2. Requirements related to 
PASS were imposed by Order for many facilities and were added to or 
included in the TSs for nuclear power reactors currently licensed to 
operate. Lessons learned and improvements implemented over the last 20 
years have shown that the information obtained from PASS can be readily 
obtained through other means or is of little use in the assessment and 
mitigation of accident conditions.
    The proposed changes are based on NRC-approved Technical 
Specification Task Force (TSTF) Standard Technical Specification Change 
Traveler, TSTF-413, ``Elimination of Requirements for a Post-Accident 
Sampling System (PASS).'' The NRC staff issued a notice of opportunity 
for comment in the Federal Register on December 27, 2001 (66 FR 66949), 
on possible amendments concerning TSTF-413, including a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination, using the consolidated line item improvement process. 
The NRC staff subsequently issued a notice of availability of the 
models for referencing in license amendment applications in the Federal 
Register on March 20, 2002 (67 FR 13027). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated March 3, 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

    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident from any Previously Evaluated

    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radioisotopes 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety
    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: December 23, 2002.
    Description of amendment request: The amendment would change the 
Hope

[[Page 22753]]

Creek Generating Station (HCGS) reactor vessel material surveillance 
program required by Appendix H to Title 10 of the Code of Federal 
Regulations (10 CFR) part 50. This change would incorporate the Boiling 
Water Reactor Vessel and Internals Project Integrated Surveillance 
Program (ISP) into the HCGS licensing basis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously analyzed?
    Response: No.
    The proposed change implements an integrated surveillance 
program that has been evaluated by the NRC staff as meeting the 
requirements of paragraph Ill.C of Appendix H to 10 CFR 50. 
Consequently, the proposed change does not significantly increase 
the probability of any accident previously evaluated. The proposed 
change provides the same assurance of RPV [reactor pressure vessel] 
integrity. As a result, the consequences of any accident previously 
evaluated are not significantly increased.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability of occurrence or 
consequences of an accident previously analyzed.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously analyzed?
    Response: No.
    The proposed change revises the HCGS licensing basis to reflect 
participation in the ISP. The proposed change does not involve a 
modification of the design of plant structures, systems or 
components (SSC). Also, the proposed change will not degrade the 
reliability of SSCs important to safety since protective features 
will not be deleted or modified. The proposed change will not impact 
the manner in which the plant is normally operated. The proposed 
change maintains an equivalent level of RPV material surveillance 
and does not introduce any new accident initiators. Therefore, this 
proposed amendment does not create the possibility of a new or 
different kind of accident from any previously analyzed.
    3. Does the change involve a significant reduction in the margin 
of safety?
    Response: No.
    The proposed change has been evaluated as providing an 
acceptable alternative to the plant-specific RPV material 
surveillance program that meets the requirements of the regulations 
for RPV material surveillance. Therefore, these changes do not 
involve a significant reduction in margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, PO Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of amendment request: September 20, 2002, as revised on 
February 14, 2003. This notice supercedes a previous notice (67 FR 
75884) published on December 10, 2002, which was based on the 
licensee's application dated September 20, 2002.
    Description of amendment request: The proposed amendment will: (1) 
Add a new limiting condition for operation (LCO) for spent fuel pool 
(SFP) boron concentration; (2) relocate requirements from Technical 
Specification (TS) Section 5.0, ``Design Features,'' to a new LCO in TS 
Section 3/4.7; and (3) revise existing TS 3/4.9.1 for refueling 
operations by relocating requirements for boron concentration to the 
Core Operating Limits Report (COLR) described in TS 6.9.1.9. The 
licensee also proposed related changes to the TS Bases. By letter dated 
February 14, 2003, PSEG revised its request, including lowering the 
minimum SFP boron concentration from 2300 parts per million (ppm) to 
800 ppm.
    Therefore, this notice supercedes a previous notice published on 
December 10, 2002, to reflect this change.
    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 a 
revised analysis of the issue of no significant hazards consideration. 
The NRC staff has reviewed the licensee's analysis against the 
standards of 10 CFR 50.92(c). The NRC staff's review is presented 
below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The licensee proposed to change the Salem Nuclear Generating 
Station (Salem) TSs by: (1) adding a new LCO for SFP boron 
concentration; (2) relocating requirements from TS Section 5.0, 
``Design Features,'' to a new LCO in TS Section 3/4.7; and (3) revising 
existing TS 3/4.9.1 for refueling operations by relocating requirements 
for boron concentration to the COLR. These changes are consistent with 
applicable LCOs in NUREG-1431, Revision 2, ``Improved Standard 
Technical Specifications, Westinghouse Plants,'' and will continue to 
provide administrative controls to ensure that a proper boron 
concentration is maintained in accordance with Salem's accident 
analyses. Because there are no changes to any of the input assumptions 
associated with postulated accidents involving refueling operations and 
the SFP, the proposed amendment does not involve a significant increase 
in the probability of occurrence or consequences of an accident 
previously analyzed.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously analyzed?
    Adding new LCOs for boron concentration in the SFP and relocating 
boron concentration requirements to the COLR will not change the 
conduct of operations in the SFP, refueling cavity and fuel transfer 
tube at Salem. Therefore, because plant operations will not change, the 
proposed amendment does not create the possibility of a new or 
different kind of accident from any previously analyzed.
    3. Does the proposed change involve a significant reduction in the 
margin of safety?
    Refueling operations and SFP boron concentration limits will be 
based on approved methodologies and accident analyses that are 
unchanged as a result of the proposed TS amendments. Therefore, because 
existing margins of safety will be maintained, the proposed change does 
not involve a significant reduction in the margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, PO Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment requests: February 28, 2003.
    Description of amendment requests: The proposed license amendments 
would revise Action A of Technical Specification (TS) 3.5.2, ``ECCS--
Operating,'' to change the completion time for restoring centrifugal 
charging pump (CCP) 1-1 to operable status during Diablo Canyon Power 
Plant (DCPP) Unit 1 Cycle 12, from 72 hours to 7 days. The 72-hour 
allowed completion time is not sufficient to

[[Page 22754]]

accomplish such emergent repairs on an inoperable CCP. This license 
amendment request also removes a similar one-time change for DCPP Unit 
2 CCP 2-1 which has expired.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The emergency core cooling system (ECCS) and the centrifugal 
charging pumps (CCPs) are designed to respond to mitigate the 
consequences of an accident. They are not an accident initiator, and 
as such cannot increase the probability of an accident.
    The loss of both CCPs, due to an inoperable CCP 1-1 and a single 
failure of CCP 1-2, could increase the consequences of an accident. 
A probabilistic risk assessment was performed to evaluate the 
increased consequences. The worst case risk increment due to the 
increased completion time for CCP 1-1 and the maximum allowed 
results in only a small quantitative impact on plant risk.
    Allowing 7 days to complete the seal replacement and post-
maintenance testing of CCP 1-1 is acceptable since the ECCS system 
remains capable of performing its intended function of providing at 
least the minimum flow assumed in the accident analyses. During the 
extended maintenance and test period, appropriate compensatory 
measures will be implemented to restrict high risk activity. The 
consequences of accidents, which rely on the ECCS system, will not 
be significantly affected.
    Therefore, the proposed changes will not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no new failure modes or mechanisms created due to 
plant operation for an extended period to perform repairs and post-
maintenance testing of CCP 1-1. Extended operation with an 
inoperable CCP does not involve any modification in the operational 
limits or physical design of the systems. There are no new accident 
precursors generated due to the extended allowed completion time.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Plant operation for seven days with an inoperable CCP 1-1 does 
not adversely affect the margin of safety. During the extended 
allowable completion time the ECCS system maintains the ability to 
perform its safety function of providing at least the minimum flow 
assumed in the accident analyses. During the extended maintenance 
and test period, appropriate compensatory measures will be 
implemented to restrict high-risk activity.
    Therefore, the change does not involve a significant reduction 
in a margin of safety as defined in the basis for any Technical 
Specification.

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

Previously Published Notices of Consideration of Issuance of Amendments 
to Facility Operating Licenses, Proposed no Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Florida Power & Light Company, et al. (FPL's), Docket Nos. 50-335, and 
50-389, St. Lucie Plant, Unit No. 1, and Unit No. 2, St. Lucie County, 
Florida

    Date of amendment request: October 23, 2002.
    Description of amendment request: The proposed license amendments 
would revise the Technical Specifications to include the design of a 
new cask pit spent fuel storage rack for each unit to increase the 
allowable spent fuel wet storage capacity at both units and include the 
description of Boral TM as the neutron absorbing material 
used in the new cask pit storage racks. The proposal would also revise 
the spent fuel pool thermal-hydraulic analyses for core offload times 
and include a change in FPL's commitments regarding the Unit 2 spent 
fuel cooling system design basis described in the Updated Final Safety 
Analysis Report.
    Date of publication of individual notice in the Federal Register: 
January 28, 2003 (68 FR 4244), as corrected March 31, 2003 (64 FR 
15487).
    Expiration date of individual notice: February 27, 2003.

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

    Date of application for amendments: February 14, 2003.
    Description of amendments request: Revise the Updated Final 
Analysis Report to change the methodology using a through-bolted 
connection frame that is different than the original design and 
construction of the steam generator roof compartment.
    Date of publication of individual notice in the Federal Register: 
March 14, 2003 (68 FR 12382).
    Expiration date of individual notice: April 14, 2003.

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

    Date of application for amendments: March 18, 2003.
    Description of amendments request: Revise the Updated Final 
Analysis Report to provide an alternative methodology using a Bar-Lock 
mechanical splice in lieu of the Cadweld splice used in the original 
design and construction of the concrete shield building dome.
    Date of publication of individual notice in the Federal Register: 
March 17, 2003 (68 FR 12718).
    Expiration date of individual notice: April 16, 2003.

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

[[Page 22755]]

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].

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

    Date of application for amendment: November 27, 2002.
    Brief description of amendment: The amendment deleted Section 6.17, 
``Post Accident Sampling,'' and thereby eliminating the requirements to 
have and maintain the subject system. The subject requirements were 
imposed by a July 7, 1981, Nuclear Regulatory Commission Confirmatory 
Order.
    Date of Issuance: April 4, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days.
    Amendment No.: 237.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 21, 2003 (68 FR 
2798).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated April 4, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 11, 2002, as supplemented 
January 22, 2003.
    Brief description of amendments: The amendment changes Technical 
Specifications 3.7.11 related to the operation of the spent fuel pool 
exhaust ventilation system during the movement of irradiated fuel 
assemblies.
    Date of issuance: April 7, 2003.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 234, 257.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 15, 2002 (67 FR 
63689).
    The January 22, 2003, supplemental letter provided clarifying 
information that did not enlarge the scope of the amendments as noticed 
in the original Federal Register notice or change the initial proposed 
no significant hazards consideration determination. The Commission's 
related evaluation of these amendments is contained in a Safety 
Evaluation dated April 7, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 17 and August 6, 2002.
    Brief description of amendments: These amendments permit operation 
of Calvert Cliffs Unit 2 with a core containing up to eight lead fuel 
assemblies with fuel rods clad with an advanced zirconium-based alloy.
    Date of issuance: April 14, 2003.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 258 and 235.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 17, 2002 (67 
FR 58637).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated April 14, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 10, 2002, as 
supplemented on November 22, 2002, and January 28, 2003. The October 
10, 2002, application replaced the original application dated December 
12, 2001.
    Brief description of amendment: This amendment changes Technical 
Specification (TS) Tables 3.2.A, 3.2.B, 4.2.A, and 4.2.B. The proposed 
changes affect various instrument trip level settings and decrease 
calibration frequencies for a variety of instruments. The proposed 
changes identify that the Reactor Water Cleanup (RWCU) system requires 
one channel in each of the two trip systems for each location. The 
proposed changes also clarify the titles of certain trip systems, move 
note numbers to their proper location, and correct a mis-referenced 
figure in a table note. Appropriate Bases pages were also changed to 
reflect the TS changes.
    Date of issuance: April 17, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 198.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 18, 2003, (68 
FR 7815).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 17, 2003.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

    Date of application for amendments: September 27, 2002.
    Brief description of amendments: The amendments revise Appendix B, 
``Environmental Protection Plan (Non-Radiological),'' of the licenses 
to remove a parenthetical reference to a superseded section of 10 CFR 
part 51.
    Date of issuance: April 4, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 132/132.
    Facility Operating License Nos. NPF-37 and NPF-66: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 29, 2002 (67 FR 
66009).

[[Page 22756]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 4, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 4, 2002, as supplemented by 
letter dated February 19, 2003.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) Surveillance Requirement (SR) 4.0.3 to extend the 
delay period, before entering a Limiting Condition for Operation, 
following a missed surveillance to ``* * * up to 24 hours or up to the 
limit of the specified frequency, whichever is greater.'' In addition, 
the amendment adds requirements to SR 4.0.3 to perform a risk 
evaluation for any Surveillance delayed greater than 24 hours and 
manage the risk impact, and specifies actions to be taken when a 
delayed surveillance is not performed or not met. The amendment is 
consistent with TS Task Force traveler TSTF-358, which has been 
approved by the Nuclear Regulatory Commission for incorporation into 
standard technical specifications in NUREG-1430. The TS Bases will be 
revised under the licensee's existing TS Bases control program to be 
consistent with the bases for TSTF-358.
    Date of issuance: April 11, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 254.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 7, 2003 (68 FR 
804).
    The supplemental information contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 11, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 10, 2002.
    Brief description of amendment: This amendment revises Surveillance 
Requirement (SR) 3.0.3 to extend the delay period, before entering a 
Limiting Condition for Operation, following a missed surveillance. The 
delay period is extended from the current limit of ``* * * up to 24 
hours or up to the limit of the specified Frequency, whichever is 
less'' to ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is greater.'' In addition, the following 
requirement is added to SR 3.0.3: ``A risk evaluation shall be 
performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    Date of issuance: April 17, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 125.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 18, 2003 (68 FR 
12954).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 17, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 14, 2002, as supplemented 
by letter dated January 20, 2003.
    Brief description of amendment: This amendment revised technical 
specification (TS) 5.5.12, ``Primary Containment Leakage Rate Testing 
Program,'' to allow a one-time exception to Nuclear Energy Institute 
94-01, ``Industry Guidance for Implementing Performance-Based Option of 
10 CFR part 50 Appendix J,'' that extends the test interval of the 
containment integrated leak rate test from 10 to 15 years.
    Date of issuance: April 8, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 126.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5676).
    The supplemental information contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 8, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 15, 2002, as 
supplemented December 13, 2002.
    Brief description of amendments: The amendments revise Technical 
Specifications Section 6.8.4.h, Containment Leakage Rate Testing 
Program, to allow a one-time 5-year extension to the current 10-year 
test interval for the containment integrated leak rate test (ILRT). The 
changes were submitted on a risk-informed basis as described in 
Regulatory Guide 1.174, An Approach for Using Probabilistic Risk 
Assessment in Risk-Informed Decisions on Plant-Specific Changes to the 
Licensing Basis. The risk-informed analysis supporting the changes 
indicates that the increase in risk from extending the ILRT test 
interval from 10 to 15 years is insignificant.
    Date of Issuance: April 10, 2003.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 187 & 130.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 17, 2002 (67 
FR 58647).
    The supplement dated December 13, 2002, provided clarifying 
information that did not change the scope of the August 15, 2002, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 10, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 28, 2002, as supplemented 
December 18, 2002, January 18, 2003, and February 25, 2003.
    Brief description of amendment: The amendment relaxes certain 
Technical Specifications (TSs) requirements for containment isolation 
and removes references to the Filtration Recirculation and Ventilation 
System charcoal filters.
    Date of issuance: April 15, 2003.

[[Page 22757]]

    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 146.
    Facility Operating License No. NPF-57: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: February 18, 2003 (68 
FR 7818).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 15, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 9, 2002, as supplemented 
November 22, 2002, and December 6, 2002.
    Brief description of amendment: The amendment grants, on a one-time 
basis, an extension of the Type A Integrated Leak Rate Test interval 
from 10 years to 15 years.
    Date of issuance: April 16, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 147.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 18, 2003 (68 
FR 7819).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 16, 2003.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of application for amendments: October 23, 2002.
    Brief description of amendments: The amendments revise the Salem 
Nuclear Generating Station, Unit Nos. 1 and 2, Technical Specification 
(TS) 6.12, ``High Radiation Area'' to be consistent with the Standard 
TSs for Westinghouse Plants (NUREG-1431, Revision 2) by updating the 
current reference to Title 10 of the Code of Federal Regulations (10 
CFR), Section 20.203 with the corresponding reference to 10 CFR 
20.1601.
    Date of issuance: April 10, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment Nos.: 255 and 236.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 4, 2002 (68 FR 
5681).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 10, 2003.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of application for amendments: July 25, 2002, as supplemented 
October 21, 2002.
    Brief description of amendments: The amendments revise Technical 
Specifications (TSs) Surveillance Requirement (SR) 4.0.3 to extend the 
delay period, before entering a Limiting Condition for Operation, 
following a missed surveillance. The delay period is extended from the 
current limit of up to 24 hours to ``* * * up to 24 hours or up to the 
limit of the specified frequency, whichever is greater.'' In addition, 
the following requirement is added to SR 4.0.3: ``A risk evaluation 
shall be performed for any surveillance delayed greater than 24 hours 
and the risk impact shall be managed.'' The amendments also add a 
requirement for a TS Bases Control Program to the administrative 
controls section of TSs and makes administrative changes to SRs 4.0.1 
and 4.0.3 to be consistent with NUREG-1431, Revision 2, ``Standard 
Technical Specifications Westinghouse Plants.''
    Date of issuance: April 16, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 256 and 237.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the TSs.
    Date of initial notice in Federal Register: February 18, 2003 (68 
FR 7820).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 16, 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: August 1, 2002.
    Description of amendments request: The amendments revised the 
Updated Safety Analysis Report (UFSAR) to eliminate consideration of a 
pressure regulator downscale failure as an abnormal operational 
transient.
    Date of issuance: April 4, 2003.
    Effective date: As of the date of issuance, to be incorporated into 
the UFSAR at the time of its next update.
    Amendment Nos.: 244, 281 and 239.
    Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: 
Amendments revised the UFSAR.
    Date of initial notice in Federal Register: October 15, 2002 (67 FR 
63697).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 4, 2003.
    No significant hazards consideration comments received: No.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a

[[Page 22758]]

nuclear power plant or in prevention of either resumption of operation 
or of increase in power output up to the plant's licensed power level, 
the Commission may not have had an opportunity to provide for public 
comment on its no significant hazards consideration determination. In 
such case, the license amendment has been issued without opportunity 
for comment. If there has been some time for public comment but less 
than 30 days, the Commission may provide an opportunity for public 
comment. If comments have been requested, it is so stated. In either 
event, the State has been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, located at One White Flint North, Public File Area 01F21, 11555 
Rockville Pike (first floor), Rockville, Maryland. Publicly available 
records will be accessible from the Agencywide Documents Assess and 
Management System's (ADAMS) Public Electronic Reading Room on the 
Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. 
If you do not have access to ADAMS or if there are problems in 
accessing the documents located in ADAMS, contact the NRC Public 
Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or 
by e-mail to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By May 16, 2003, the licensee 
may file a request for a hearing with respect to issuance of the 
amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.714, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, and 
electronically on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are problems in accessing the 
document, contact the PDR Reference staff at 1-800-397-4209, 301-415-
4737, or by e-mail to [email protected]. If a request for a hearing or 
petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, Public File Area 01F21, 11555 
Rockville Pike (first floor), Rockville, Maryland,

[[Page 22759]]

by the above date. Because of the continuing disruptions in delivery of 
mail to United States Government offices, it is requested that 
petitions for leave to intervene and requests for hearing be 
transmitted to the Secretary of the Commission either by means of 
facsimile transmission to (301) 415-1101 or by e-mail to 
[email protected]. A copy of the petition for leave to intervene 
and request for hearing should also be sent to the Office of the 
General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and because of continuing disruptions in delivery of mail 
to United States Government offices, it is requested that copies be 
transmitted either by means of facsimile transmission to (301) 415-3725 
or by e-mail to [email protected]. A copy of the request for 
hearing and petition for leave to intervene should also be sent to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

    Date of amendment request: April 14, 2003, as supplemented by 
letter dated April 15, 2003.
    Description of amendment request: The amendment revises Limiting 
Condition for Operation (LCO) 3.7.5, ``Control Building Chiller (CBC) 
System,'' Required Action A.1 to add a provision that temporarily 
removes the restrictions of LCO 3.0.4 until May 16, 2003. This 
amendment allows entry into LCO 3.7.5 with an inoperable CBC subsystem.
    Date of issuance: April 16, 2003.
    Effective date: As of the date of issuance and shall be implemented 
immediately.
    Amendment No.: 250.
    Facility Operating License No. DPR-49: Amendment revises the 
technical specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, state consultation, and final NSHC 
determination are contained in a safety evaluation dated April 16, 
2003.
    Attorney for licensee: Mr. Alvin Gutterman, Morgan Lewis, 1111 
Pennsylvania Avenue NW., Washington, DC 20004
    NRC Section Chief: L. Raghavan.

    Dated at Rockville, Maryland, this 21st day of April, 2003.

    For the Nuclear Regulatory Commission
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 03-10396 Filed 4-28-03; 8:45 am]
BILLING CODE 7590-01-P