[Federal Register Volume 68, Number 92 (Tuesday, May 13, 2003)]
[Notices]
[Pages 25648-25664]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-11697]


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NUCLEAR REGULATORY COMMISSION


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

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 29, 2003 (68 FR 22744).

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

[[Page 25649]]

take this action, it will publish in the Federal Register a notice of 
issuance and provide for opportunity for a hearing after issuance. The 
Commission expects that the need to take this action will occur very 
infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By June 12, 2003, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714, which is 
available at the Commission's PDR, located at One White Flint North, 
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System's (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing 
or petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, Public File Area 01F21, 11555 
Rockville Pike (first floor), Rockville, Maryland, by the above date. 
Because of continuing disruptions in delivery of mail to United States 
Government offices, it is requested that petitions for leave to 
intervene and requests for hearing be transmitted to the Secretary of 
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the request for 
hearing and petition for leave to intervene should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and because of continuing disruptions in 
delivery of mail to United States Government offices, it is requested 
that copies be transmitted either by means of facsimile transmission to 
301-415-3725 or by e-mail to [email protected]. A copy of the 
request for hearing and petition for leave to intervene should also be 
sent to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North,

[[Page 25650]]

Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System's (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or 
by e-mail to [email protected].

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

    Date of amendment request: April 2, 2001, as supplemented by 
letters dated January 15, August 23, 2002, and March 28, 2003.
    Description of amendment request: The proposed amendment would add 
operational restrictions when the inclined fuel transfer system (IFTS) 
blind flange is removed during Modes 1, ``Power Operation,'' 2, 
``Startup,'' or 3, ``Hot Shutdown.'' The proposed changes would (1) 
include a limitation on the duration that the IFTS blind flange can be 
removed while primary containment integrity is required, (2) include a 
limitation on the duration that the IFTS blind flange can remain in the 
unbolted configuration, (3) specify the need to install the steam dryer 
pool to reactor cavity pool gate prior to opening the blind flange, and 
(4) provide the flexibility to remove the IFTS blind flange for other 
than maintenance and testing purposes only.
    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 evaluated?
    The proposed changes allow operation of the IFTS while primary 
containment operability is required. The proposed changes result in 
a change to the primary containment boundary. A loss of primary 
containment integrity is not an accident initiator. The proposed 
changes do not involve any modifications to plant systems or design 
parameters or conditions that contribute to the initiation of any 
accidents previously evaluated. Therefore, the proposed changes do 
not increase the probability of any accident previously evaluated.
    The proposed changes potentially affect the allowable leakage of 
the containment structure which is designed to mitigate the 
consequences of a loss-of-coolant accident (LOCA). The function of 
the primary containment is to maintain functional integrity during 
and following the peak transient pressures and temperatures that 
result from any LOCA. The primary containment is designed to limit 
fission product leakage following the design basis LOCA. Because the 
proposed changes do not alter the plant design, only the extent of 
the boundaries that provide primary containment isolation for the 
IFTS penetration, the proposed changes do not result in an increase 
in primary containment leakage. In addition, a time limit for IFTS 
blind flange removal of 40 days per cycle and a 12 hour limit for 
the unbolted configuration of the IFTS flange have been established 
as conservative measures to limit the associated risk to the 
containment boundary for all accident conditions. Once the blind 
flange is removed the IFTS transfer tube and its appurtenances 
become part of the primary containment boundary. As part of the 
primary containment boundary these subject components would be 
exposed to LOCA pressures. While these components have not been 
fabricated or installed to meet the acceptance criteria for a 
containment penetration, they have been built to withstand the 
rigors of a commercial nuclear application. This includes, but is 
not limited to, consideration of adequate seismic support, inertial 
forces imparted to the fuel, appropriate cooling and shielding for 
the spent nuclear fuel, integrity of the fluid system pressure 
boundary, and a safety analysis, including a failure modes and 
effects evaluation which assumes that credible events and credible 
combinations of events have been considered and mitigated against by 
either a fail safe design or redundancy. They are judged to be an 
acceptable barrier to prevent the uncontrolled release of post-
accident fission products for the purposes of this amendment 
request.
    Further, it has been shown that the largest potential leakage 
pathway, the IFTS transfer tube itself, would remain sealed by the 
depth of water required by the proposed [technical specification] TS 
change to be maintained in the fuel building fuel transfer pool. The 
transfer tube drain line constitutes the other possible leakage 
pathway, and will be required to be capable of being isolated via 
administrative control of the manual isolation valve in the drain 
line. Additionally, due to the physical relationships of the 
buildings and components involved, any leakage from either of these 
pathways is fully contained within the boundaries of the secondary 
containment and would be filtered by the Standby Gas Treatment 
System prior to release to the environment.
    Leakage from the containment upper pool through the open IFTS 
transfer tube could potentially result in the excessive loss of 
water from the volume intended to provide post-LOCA makeup water to 
the suppression pool. The upper pool dump volume is maintained by 
requiring the installation of the steam dryer pool to reactor cavity 
pool gate with the seal inflated and a backup air supply provided. 
Maintaining the upper pool dump volume ensures proper suppression 
pool level can be achieved following a LOCA which provides for long-
term steam condensation.
    Based on the above, the proposed changes do not increase the 
consequences of an accident previously evaluated.
    In summary, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not involve a change to the plant design 
or operation except for when IFTS is operated. As a result, the 
proposed changes do not affect any of the parameters or conditions 
that could contribute to the initiation of any accidents. No new 
accident modes or equipment failure modes are created by these 
changes. Extending the primary containment boundary to include 
portions of the IFTS has no influence on, nor does it contribute to 
the possibility of a new or different kind of accident or 
malfunction from those previously evaluated. Furthermore, operation 
of IFTS is unrelated to the operation of the reactor. There is no 
mishap in the process that can lead or contribute to the possibility 
of losing any coolant in the reactor or introducing the chance for 
positive or negative reactivity or other accidents different from 
and not bounded by those previously evaluated. Therefore, these 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes only affect the extent of a portion of the 
primary containment boundary. The time that the IFTS is in the 
seismically indeterminate configuration with the flange unbolted 
will be limited to 12 hours per operating cycle. The time the IFTS 
blind flange will be removed will be limited to 40 days per 
operating cycle. These restrictions will limit the risk from the 
potential leakage through the primary containment boundary. Having 
IFTS in operation does not affect the reliability of equipment used 
for core cooling. In addition, precautions will be taken to 
administratively control the IFTS transfer tube drain path so that 
the proposed change will not increase the probability that an 
increase in leakage from the primary containment to the secondary 
containment could occur. Precautions will also be taken to ensure 
that the steam dryer pool to reactor cavity pool gate is installed 
prior to removing the IFTS flange when primary containment is 
required to be operable. Installation of this gate will ensure that 
an adequate containment upper pool dump volume is maintained to 
support post-LOCA suppression pool makeup water volume requirements.
    The margin of safety that has the potential of being impacted by 
the proposed changes involve the offsite dose consequences of 
postulated accidents which are directly related to containment 
leakage rate. The containment isolation system is designed to limit 
leakage to La which is defined by the

[[Page 25651]]

[Clinton Power Station] CPS TS to be 0.65% of primary containment 
air weight per day at the design basis LOCA maximum peak containment 
pressure (i.e., Pa). The limitation on containment 
leakage rate is designed to ensure that total leakage volume will 
not exceed the volume assumed in the accident analyses at 
Pa. The margin of safety for the offsite dose 
consequences of postulated accidents directly related to the 
containment leakage rate is maintained by meeting the La 
acceptance criteria during operation. The La value is not 
being modified by this proposed TS change. The IFTS will continue to 
provide an acceptable barrier to prevent unacceptable containment 
leakage during a LOCA, and therefore these changes will not create a 
situation causing the containment leakage rate acceptance criteria 
to be violated.
    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: Edward J. Cullen, Deputy General Counsel 
Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendments request: March 28, 2003.
    Description of amendments request: The amendment would remove the 
post-accident hydrogen monitoring and control requirements from the 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 Technical 
Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed change to the Technical Specifications has been 
evaluated against the standards in 10 CFR 50.92. The proposed 
amendment revises Technical Specification 3.3.10, Post-Accident 
Monitoring Instrumentation, and Technical Specification Table 
3.3.10-1, Post-Accident Monitoring Instrumentation to delete 
references to the containment hydrogen analyzers. Additionally, the 
proposed amendment will delete Technical Specification 3.6.7, 
Hydrogen Recombiners. The proposed change has been determined to not 
involve a significant hazards consideration, in that operation of 
the facility in accordance with the proposed amendments:
    1. Would not involve a significant increase in the probability 
or consequences of any accident previously evaluated.
    Components used in the control of hydrogen in the Containment 
(consisting of hydrogen recombiners, a hydrogen vent, and hydrogen 
detectors) are not considered accident initiators. Therefore, this 
change does not increase the probability of an accident previously 
evaluated.
    The purpose of the Hydrogen Control System is to ensure that 
hydrogen concentration is maintained below 4.0 volume percent so 
that Containment integrity is not challenged following a design 
basis loss-of-coolant accident (LOCA). The Calvert Cliffs Nuclear 
Power Plant Individual Plant Examination analyzed the probability of 
Containment failure under a variety of conditions. This proposed 
amendment does not alter the conclusions or assumptions of the 
Individual Plant Examination. The Calvert Cliffs Nuclear Power Plant 
Containment provides a safety margin against hydrogen burn following 
a design basis accident, such that the Containment will not fail 
even without hydrogen control equipment. Therefore, this change does 
not increase the consequences of accidents previously evaluated.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Would not create the possibility of a new or different [kind] 
of accident from any accident previously evaluated.
    The proposed change does not change the configuration of the 
plant beyond the Hydrogen Control System. Hydrogen generation 
following a design basis LOCA has been evaluated. Deletion of the 
Hydrogen Control System from the plant design basis and Technical 
Specifications does not alter the generation of hydrogen post-LOCA.
    Therefore, this change does not create the possibility of a new 
or different [kind] of accident from any accident previously 
evaluated.
    3. Would not involve a significant reduction in [a] margin of 
safety.
    The margin of safety in this case is the ability of Containment 
to withstand a pressure increase caused by the deflagration of 
hydrogen in the Containment. Industry experience and experimentation 
has shown that large, dry, well-ventilated Containments such as 
those at Calvert Cliffs can withstand pressures generated by 
ignition of hydrogen resulting from a LOCA. The Calvert Cliffs 
Nuclear Power Plant Containment provides a safety margin against 
hydrogen burn following a design basis accident, such that the 
Containment will not fail even without hydrogen control equipment.
    Therefore, this change does not significantly reduce [a] margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Richard J. Laufer.
    Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, 
Unit No. 1, Pope County, Arkansas
    Date of amendment request: April 2, 2003.
    Description of amendment request: The proposed amendment would 
change the spent fuel pool loading restrictions by redefining the 
regions, inserting Metamic[reg] poison panels in a portion of the spent 
fuel pool, and increasing the minimum boron concentration.
    Basis for no significant hazards consideration determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The three fuel handling accidents described below can be 
postulated to increase reactivity. However, for these accident 
conditions, the double contingency principle of ANS [American 
Nuclear Society] N16.1-1975 is applied. This states that it is 
unnecessary to assume two unlikely, independent, concurrent events 
to ensure protection against a criticality accident. Thus, for 
accident conditions, the presence of soluble boron in the storage 
pool water can be assumed as a realistic initial condition since its 
absence would be a second unlikely event.
    Three types of drop accidents have been considered: a vertical 
drop accident, a horizontal drop accident, and an inadvertent drop 
of an assembly between the outside periphery of the rack and the 
pool wall:
    [sbull] A vertical drop directly upon a cell will cause damage 
to the racks in the active fuel region. The current 1600 ppm soluble 
boron concentration TS limit will ensure that Keff does 
not exceed 0.95.
    [sbull] A fuel assembly dropped on top of the rack horizontally 
will not deform the rack structure such that criticality assumptions 
are invalidated. The rack structure is such that an assembly 
positioned horizontally on top of the rack results in a separation 
distance from the upper end of the active fuel region of the stored 
assemblies. This distance is sufficient to preclude interaction 
between the dropped assembly and the stored fuel.
    [sbull] An inadvertent drop of an assembly between the outside 
periphery of the rack and the pool wall is bounded by the worst case 
fuel misplacement accident condition.
    The fuel assembly misplacement accident was considered for all 
storage configurations. An assembly with high reactivity is assumed

[[Page 25652]]

to be placed in a storage location which requires restricted storage 
based on initial U-235 loading, cooling time, and burnup. The 
presence of boron in the pool water assumed in the analysis has been 
shown to offset the worst case reactivity effect of a misplaced fuel 
assembly for any configuration. This boron requirement is less than 
the 1600 ppm currently required by the ANO-1 TS. Thus, a five 
percent subcriticality margin can be easily met for postulated 
accidents, since any reactivity increase will be much less than the 
negative worth of the dissolved boron.
    For fuel storage applications, water is usually present. An 
``optimum moderation'' accident is not a concern in spent fuel pool 
storage racks because the rack design prevents the preferential 
reduction of water density between the cells of a rack (e.g., 
boiling between cells). An ``optimum moderation'' accident in the 
new fuel pit was previously evaluated and the conclusions of that 
evaluation have not changed as a result of the fuel enrichment.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes will define a portion of the current Region 
2 as Region 3. The new region will contain Metamic[reg] poison panel 
inserts and will allow unrestricted storage of fuel assemblies with 
various enrichments and burnup. To support the proposed change, new 
criticality analyses have been performed. The analyses resulted in 
new loading restrictions in Region 1 and Region 2. The presence of 
boron in the pool water assumed in the analysis is less than the 
1600 ppm currently required by the ANO-1 TSs.
    Thus, a five percent subcriticality margin can be easily met for 
postulated accidents, since any reactivity increase will be much 
less than the negative worth of the dissolved boron.
    No new or different types of fuel assembly drop scenarios are 
created by the proposed change. During the installation of the 
Metamic[reg] panels, the possible drop of a panel is bounded by the 
current fuel assembly drop analysis. No new or different fuel 
assembly misplacement accidents will be created. Administrative 
controls currently exist to assist in assuring fuel misplacement 
does not occur.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    With the presence of a nominal boron concentration, the SFP 
storage racks will be designed to assure that fuel assemblies of 
less than or equal to five weight percent U-235 enrichment when 
loaded in accordance with the proposed loading restrictions will be 
maintained within a subcritical array with a five percent 
subcritical margin (95% probability at the 95% confidence level). 
This has been verified by criticality analyses.
    Credit for soluble boron in the SFP water is permitted under 
accident conditions. The proposed modification that will allow 
insertion of Metamic[reg] poison panels does not result in the 
potential of any new misplacement scenarios. Criticality analyses 
have been performed to determine the required boron concentration 
that would ensure the maximum Keff does not exceed 0.95. 
The ANO-1 TS for the minimum SFP boron concentration is greater than 
that required to ensure Keff does not exceed 0.95. 
Therefore, the margin of safety currently defined by taking credit 
for soluble boron will be maintained.
    The structural analysis of the spent fuel racks, along with the 
evaluation of the SFP structure, showed that the integrity of these 
structures will be maintained with the addition of the poison 
inserts. The structural requirements were shown to be satisfied, so 
the safety margins were maintained.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: 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] 
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 U.S. Nuclear Regulatory Commission (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 19, 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

[[Page 25653]]

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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: April 3, 2003.
    Description of amendment request: The proposed changes will revise 
the Updated Final Safety Analysis Report to change the Reactor Vessel 
Material Surveillance Program. The change will reflect participation in 
the Boiling Water Reactor Vessel and Internals Project (BWRVIP) 
Integrated Surveillance Program (ISP).
    Basis for proposed no significant hazards consideration 
determination: As required by Section 50.91(a) of Title 10 of the Code 
of Federal Regulations (10 CFR ), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Changes in the fracture toughness properties of reactor vessel 
beltline materials, resulting from the neutron irradiation and the 
thermal environment, are monitored by a surveillance program in 
compliance with the requirements of 10CFR50, Appendix H. The 
proposed change implements an integrated surveillance program that 
has been evaluated by the NRC [U.S. Nuclear Regulatory Commission] 
staff as meeting the requirements of paragraph III.C of Appendix H 
to 10 CFR 50. The BWRVIP's ISP surveillance material selection 
process adequately ensures that materials in the program effectively 
provide meaningful information to monitor changes in fracture 
toughness for GGNS [Grand Gulf Nuclear Station, Unit 1, or Grand 
Gulf] RPV [Reactor Pressure Vessel] materials. In addition, the ISP 
program requires participants to acquire and evaluate relevant ISP 
test data from the program which may affect RPV integrity 
evaluations in a timely manner. One advantage of participating in 
the BWRVIP ISP is that surveillance test data applicable to the 
Grand Gulf RPV will be available sooner than under the current plant 
specific program.
    The proposed change will not affect current RPV performance and 
will not cause the RPV or interfacing systems to be operated outside 
of their design or testing limits. The proposed change will not 
alter any assumptions previously made in evaluating the radiological 
consequences of accidents.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The change does not affect the design, function, reliability, or 
operation of any plant structure, system or component. The purpose 
of the reactor vessel material surveillance program is to monitor 
neutron embrittlement and thermal environment effects in order to 
predict the behavioral characteristics of materials of pressure 
retaining components of the reactor coolant pressure boundary and to 
ensure that reactor vessel fracture toughness and integrity 
requirements are not violated. The ISP is an approved alternate 
monitoring program that meets the regulatory requirements in 
Appendix H to 10 CFR 50. As an acceptable alternate monitoring 
program, the ISP cannot create a new failure mode involving the 
possibility of a new or different kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from that previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The reactor material surveillance program required by 10 CFR 50, 
Appendix H, is designed to ensure that adequate margins of safety 
are provided for the reactor coolant pressure boundary during any 
condition of normal operation, including anticipated operational 
occurrences and hydrostatic tests. Monitoring changes in the 
fracture toughness of reactor vessel materials ensures that material 
changes due to radiation embrittlement are adequately considered for 
safe reactor operations. Paragraph lll.C of Appendix H to 10 CFR 50 
delineates the regulatory requirements for an ISP. The BWRVIP ISP 
meets these requirements and has been approved by the NRC.
    One of the uses of the material surveillance data obtained 
through the proposed ISP is to ensure the reactor coolant system P/T 
[Pressure/Temperature] limits established by the Technical 
Specifications are conservative. The material surveillance data 
obtained through the proposed Integrated Surveillance Program will 
provide new information that will be evaluated to ensure that the P/
T limits are conservative. In addition, a neutron fluence 
calculation methodology which has been approved by the NRC staff and 
is consistent with the attributes identified in U.S. Nuclear 
Regulatory Commission Regulatory Guide 1.190, ``Calculational and 
Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,'' 
will be used for the determination of reactor vessel and 
surveillance capsule neutron fluence values to ensure quality of the 
method and compatibility between ISP results.

[[Page 25654]]

    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

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

    Date of application for amendment request: February 14, 2003.
    Description of amendment request: The proposed amendments would 
relax the Technical Specifications (TSs) surveillance requirement (SR) 
for reactor instrumentation line excess flow check valves (EFCVs). 
Currently, TSs require testing of each reactor instrumentation line 
EFCV on a 24-month frequency. The proposed TS SR would require that a 
representative sample of reactor instrumentation line EFCVs be tested 
every 24 months, such that each EFCV will be tested nominally once 
every 10 years.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The current Technical Specification (TS) Surveillance 
Requirement (SR) frequency requires each reactor instrumentation 
line excess flow check valve (EFCV) to be tested every 24 months. 
The EFCVs at Dresden Nuclear Power Station (DNPS) and Quad Cities 
Nuclear Power Station (QCNPS) are designed to remain open during 
normal operation, but will close automatically in the event of an 
instrument line break downstream of the valve. The proposed change 
allows a reduced number of reactor instrumentation line EFCVs to be 
tested every 24 months. Industry operating experience demonstrates a 
high level of reliability for these EFCVs. A failure of an EFCV to 
isolate cannot initiate previously evaluated accidents (i.e., a 
break in a reactor coolant pressure boundary (RCPB) instrument line 
outside containment). Therefore, there is no increase in the 
probability of an accident as a result of this proposed change.
    The postulated break of an instrument line connected to the RCPB 
is discussed and evaluated in the Updated Final Safety Analysis 
Reports (UFSARs) for DNPS and QCNPS. The integrity and functional 
performance of the secondary containment and standby gas treatment 
system are not impaired by this event, and the calculated potential 
offsite exposures are below the guidelines of 10 CFR 100, ``Reactor 
Site Criteria.'' The NRC approved General Electric Nuclear Energy 
Licensing Topical Report, NEDO-32977-A, ``Excess Flow Check Valve 
Testing Relaxation,'' discusses through operating experience that 
there is a high degree of reliability with the EFCVs and that there 
are little radiological consequences resulting from an EFCV failure. 
The radiological consequences for an instrument line break do not 
credit the EFCVs for isolating the break. Therefore, the 
consequences of an instrument line break are not impacted by the 
proposed level of testing. Based on the above, the proposed TS 
change does not involve a significant increase in the consequences 
of an accident previously evaluated.
    In summary, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change allows a reduced number of reactor 
instrumentation line EFCVs to be tested every 24 months. No other 
changes in requirements are being proposed. Industry operating 
experience as documented in NEDO-32977-A, provides supporting 
evidence that the reduced testing will not affect the high 
reliability of these valves. The potential failure of an EFCV to 
isolate as a result of the proposed reduction in testing is bounded 
by the evaluation of an instrument line break described in the 
UFSARs for DNPS and QCNPS. The proposed changes do not physically 
alter the plant and will not alter the operation of structures, 
systems, and components as described in the UFSARs. Therefore, a new 
or different kind of accident from any accident previously evaluated 
will not be created.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The consequences of an unisolable rupture of a RCPB instrument 
line outside containment has been previously evaluated in the UFSARs 
for DNPS and QCNPS. That evaluation assumed a continuous discharge 
of reactor coolant for the duration of the detection and cooldown 
sequence (i.e., no credit was assumed for isolating the break by the 
associated EFCV in the ruptured instrument line). Since a continuous 
discharge was assumed in this evaluation, any potential failure of 
the associated EFCV to isolate postulated by the reduced testing 
frequency is bounded. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Vice President, 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: March 31, 2003.
    Description of amendment request: The proposed amendments would 
revise Appendix A, Technical Specifications (TS), of Facility Operating 
License Nos. NPF-11 and NPF-18. Specifically, the proposed change will 
increase the upper limit associated with TS Table 3.3.5.1-1, 
``Emergency Core Cooling System Instrumentation,'' Function 3.e, ``HPCS 
System Flow Rate--Low (Bypass),'' Allowable Value from less than or 
equal to (<=) 1704 gallons per minute (gpm) to <= 2194 gpm. The 
proposed change increases the Allowable Value band to account for 
instrumentation deadband, as-left setting tolerances and setpoint 
drift, and resolves historical difficulties during calibration. The 
current Allowable Value was initially provided in the LaSalle County 
Station TS during conversion to Improved Technical Specifications (ITS) 
format. This value was based on vendor supplied data and believed at 
the time to adequately account for these parameters. The upper 
Allowable Value limit is being increased based on historical 
performance data for the High Pressure Core Spray (HPCS) system flow 
switches. The increase in the allowed bypass flow rate does not affect 
the capability of the HPCS system in performing its intended safety 
function.
    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 probability or consequences of an accident previously evaluated.
    The proposed change to LaSalle County Station Technical 
Specifications (TS) Table 3.3.5.1-1, ``Emergency Core Cooling System

[[Page 25655]]

Instrumentation,'' Function 3.e, ``HPCS System Flow Rate--Low 
(Bypass),'' request an increase in the Allowable Value from less 
than or equal to <= 1704 gpm to <= 2194 gpm. The operation of High 
Pressure Core Spray (HPCS) System Flow Rate--Low (Bypass) function 
is not a precursor to any accident previously evaluated. Thus, the 
proposed change does not have any effect on the probability of an 
accident previously evaluated.
    The LaSalle County Station Emergency Core Cooling Systems (ECCS) 
are designed, in conjunction with the primary and secondary 
containments, to limit the release of radioactive material to the 
environment following a loss of coolant accident (LOCA). The ECCS 
uses two independent methods, flooding and spraying, to cool the 
reactor core following a LOCA. The HPCS is one of the core spray 
systems. The evaluation of the proposed change concluded that the 
HPCS will operate as assumed in accidents previously evaluated. 
Thus, the radiological consequences of any accident previously 
evaluated are not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not affect the control parameters 
governing unit operation and does not introduce any new equipment, 
modes of system operation or failure mechanisms. Calculations have 
been performed which evaluated the performance of the HPCS system 
without the closure of the minimum flow bypass valve. The 
calculations determined that the Unit 1 and Unit 2 HPCS pump 
capacity with the minimum flow bypass valve open will support HPCS 
System injection flow into the reactor pressure vessel (RPV) over 
the full range of RPV pressures above the requirements for HPCS in 
the Loss of Coolant Accident (LOCA) analyses.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The HPCS System Flow Rate--Low (Bypass) Function is one of the 
inputs to the logic that controls the opening and closing of the 
minimum flow bypass valve. The current Allowable Values for this 
function are greater than or equal to (=)1380 gpm and 
<=1704 gpm. The lower Allowable Value limit (i.e., 1380 gpm) ensures 
that the minimum flow bypass valve opens when pump flow is too low 
for adequate cooling of the pump while the pump is operating. This 
limit is not affected by the proposed change.
    The upper Allowable Value limit (i.e., 1704 gpm) ensures that 
the minimum flow bypass valve automatically closes to allow maximum 
flow to the RPV spray sparger. The proposed change increases the 
value to <= 2194 gpm. LaSalle County Station has evaluated the 
effect of this change and concluded the following:
    [sbull] The proposed change to increase the upper Allowable 
Value limit from <= 1704 gpm to <= 2194 gpm will provide further 
assurance that the minimum flow bypass valve remains full open until 
the HPCS pump flow to the RPV spray sparger is sufficient to prevent 
overheating of the pump, and
    [sbull] The upper Allowable Value ensures that the HPCS minimum 
flow bypass valve closes to allow maximum flow to the RPV spray 
sparger. The proposed change will delay the initiation of valve 
closure from <= 1704 gpm to <= 2194 gpm. The calculations determined 
that the Unit 1 and Unit 2 HPCS pump capacity with the minimum flow 
bypass valve open will support HPCS system injection flow into the 
RPV over the full range of RPV pressures above the requirements for 
HPCS in the Loss of Coolant Accident (LOCA) analysis up to the 
maximum assumed injection flow of 5400 gpm. The margin to the flow 
requirements of the LOCA analysis varies from approximately 200 gpm 
at very low RPV pressures to greater than 1000 gpm at higher RPV 
pressures. Since the HPCS system injection flow requirement to the 
RPV spray sparger assumed in the LOCA analysis is met with the 
minimum flow bypass valve open, the LOCA analysis results are not 
adversely affected by increasing the value of flow when the minimum 
flow bypass valve starts to close. Although the calculations show 
that closure of the HPCS minimum flow bypass valve is not necessary 
to meet the HPCS system injection flow requirements assumed in the 
LOCA analyses, LaSalle County Station has chosen to retain the upper 
Allowable Value in the TS to provide additional margin to the 
assumed injection flow of the analyses.
    Thus, increasing the TS upper Allowable Value limit for the HPCS 
System Flow Rate--Low (Bypass) Function from <= 1704 gpm to <= 2194 
gpm will not affect the capability of the HPCS system in performing 
its intended safety function.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Based upon the above, Exelon Generation Company concludes that 
the proposed amendment presents no significant hazards consideration 
under the standards set forth in 10 CFR 50.92(c), and, accordingly, 
a finding of ``no significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: March 19, 2003.
    Description of amendment request: The proposed amendment would 
delete requirements from the Technical Specifications (TSs) and other 
elements of the licensing bases related to the post-accident sampling 
system (PASS) at the Monticello Nuclear Generating Plant. 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. The notice included a model 
safety evaluation and model no significant hazards consideration 
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 no significant hazards consideration 
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 analysis of the issue 
of no significant hazards consideration is presented below:


[[Page 25656]]



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: Jay E. Silberg, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: L. Raghavan.

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 
approve a selective scope application of an alternative source term 
(AST) for fuel handling accidents (FHAs). Specifically, the amendments 
would revise Technical Specification (TS) 3.9.3, ``Containment 
Penetrations,'' to (1) change the Applicability statement to ``During 
movement of recently irradiated fuel assemblies within containment,'' 
and (2) modify the Required Action for Condition A to eliminate the 
requirement to suspend core alterations and add the requirement to 
suspend movement of recently irradiated fuel assemblies within 
containment if one or more containment penetrations are not in the 
required status.
    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.
    Selective implementation of the Alternative Source Term (AST) 
and those plant systems affected by implementing the proposed 
changes to the TS are not accident initiators and cannot increase 
the probability of an accident. The AST does not adversely affect 
the design or operation of the facility in a manner that would 
create an increase in the probability of an accident. Rather, the 
AST is a methodology used to evaluate the dose consequences of a 
postulated accident.
    The fuel handling accident analysis has demonstrated that the 
dose consequences of a postulated fuel handling accident remain 
within the limits provided sufficient decay has occurred prior to 
the movement of irradiated fuel without taking credit for certain 
mitigation features such as ventilation filter systems and 
containment closure. Irradiated fuel that has not undergone the 
required decay period of 65 hours is defined as recently irradiated 
fuel and the currently approved TS requirements are applicable when 
this recently irradiated fuel is being handled.
    This amendment does not alter the methodology or equipment used 
directly in fuel handling operations. Neither ventilation filter 
system (i.e., the containment purge or drumming area vent stack) is 
used to actually handle fuel. Neither of these systems is an 
accident initiator. Similarly, neither the equipment hatch, 
personnel air locks, any other containment penetrations, nor any 
component thereof is an accident initiator. No other accident 
initiator is affected by the proposed changes.
    The TEDE [total effective dose equivalent] doses from the 
analysis supporting this amendment request have been compared to 
equivalent TEDE doses estimated with the guidelines of RG 
[Regulatory Guide] 1.183 [``Alternative Radiological Source Terms 
for Evaluating Design Basis Accidents at Nuclear Power Reactors''] 
Footnote 7. The new values are shown to be comparable to the results 
of the previous analysis.
    Based on the aforementioned reasons, the proposed amendment does 
not involve a significant increase in the probability or 
consequences of a FHA as previously analyzed.
    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 evaluation of the effects of the proposed changes indicates 
that all design

[[Page 25657]]

standards and applicable safety criteria limits are met. The 
proposed amendment would increase the time during which the 
equipment hatch and personnel air locks could be open during core 
alterations and movement of irradiated fuel. The proposed amendment 
does not involve changes in the operations of these containment 
penetrations. Having these penetrations open does not create the 
possibility of a new accident.
    Therefore, operation of the Point Beach Nuclear Plant in 
accordance with the proposed amendments 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.
    The assumptions and input used in the analysis are conservative 
as noted below. The design basis FHA has been defined to identify 
conservative conditions. The source term and radioactivity releases 
have been calculated pursuant to RG 1.183, Appendix B and with 
conservative assumptions concerning prior reactor operations. The 
control room atmospheric dispersion factor has been calculated with 
conservative assumptions associated with the release. The 
conservative assumptions and input noted above ensure that the 
radiation doses cited in the amendment request are the upper bound 
to radiological consequences of a FHA either in containment or in 
the spent fuel pool. The analysis shows that there is a significant 
margin between the TEDE radiation doses calculated for the 
postulated FHA using the AST and acceptance limits of 10 CFR 50.67 
and RG 1.183. The proposed changes will not degrade the plant 
protective boundaries, will not cause a release of fission products 
to the public, and will not degrade the performance of any 
Structures, Systems, and Components important to safety. Therefore, 
there is no significant reduction in the margin of safety as a 
result of the proposed changes.

    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.

Southern Nuclear Operating Company, Inc, Docket No. 50-364, Joseph M. 
Farley Nuclear Plant, Unit 2, Houston County, Alabama

    Date of amendment request: February 11, 2003.
    Description of amendment request: The proposed amendments would 
allow a 40-month inspection interval for Farley, Unit 2 after the 
completion of the first post-replacement in-service inspection, rather 
than the completion of two consecutive inspections resulting in a 
classification of C-1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed one-time change revises the steam generator (SG) 
inspection interval requirements in TS [technical specification] 
5.5.9.3, ``Inspection Frequencies,'' for the FNP [Farley Nuclear 
Plant] Unit 2 Spring 2004 refueling outage, to allow a 40[-]month 
inspection frequency after one inspection, rather than after two 
consecutive inspections with results that are within the C-1 
category. C-1 category is defined as ``less than 5% of the total 
tubes inspected are degraded tubes and none of the inspected tubes 
are defective.''
    The proposed one-time extension of the FNP Unit 2 SG tube 
inservice inspection interval does not involve changing any 
structure, system, or component, or affect reactor operations. It is 
not an initiator of an accident and does not change any existing 
safety analysis previously analyzed in the FNP's Final Safety 
Analysis Report (FSAR). As such, the proposed change does not 
involve a significant increase in the probability of an accident 
previously evaluated.
    Since the proposed change does not alter the plant design, there 
is no direct increase in SG leakage. Industry experience indicates 
that the probability of increased SG tube degradation would not go 
undetected. Additionally, steps described below will further 
minimize the risk associated with this extension. For example, the 
scope of inspections performed during the last FNP Unit 2 refueling 
outage (i.e., the first refueling outage following SG replacement) 
exceeded the TS requirements for the first two refueling outages 
after SG replacement. That is, more tubes were inspected than were 
required by TS. Currently, FNP Unit 2 does not have a SG damage 
mechanism, and will meet the current industry examination guidelines 
without performing SG inspections during the next refueling outage. 
Additionally, as part of the FNP SG Program, both a Condition 
Monitoring Assessment and an Operational Assessment are performed 
after each inspection and compared to the Nuclear Energy Institute 
(NEI) 97-06, ``Steam Generator Program Guidelines,'' performance 
criteria. The results of the Condition Monitoring Assessment 
demonstrated that all performance criteria were met during the FNP 
Unit 2 Fall 2002 refueling outage, and the results of the 
Operational Assessment show that all performance criteria will be 
met over the proposed operating period. Considering these actions, 
along with the improved SG design and reliability of Westinghouse 
replacement SGs, extending the SG tube inspection frequency does not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change revises the SG inspection frequency 
requirements in TS 5.5.9.3.a for the FNP Unit 2 Spring 2004 
refueling outage, to allow a 40[-]month inspection interval after 
one inspection, rather than after two consecutive inspections, with 
inspection results within the C-1 category.
    The proposed change will not alter any plant design basis or 
postulated accident resulting from potential SG tube degradation. 
The scope of inspections performed during the last FNP Unit 2 
refueling outage (i.e., the first refueling outage following SG 
replacement) significantly exceeded the TS requirements for the 
scope of the first two refueling outages after SG replacement.
    Primary-to-secondary leakage that may be experienced during all 
plant conditions is expected to remain within current accident 
analysis assumptions. The proposed change does not affect the design 
of the SGs, the method of SG operation, or reactor coolant chemistry 
controls. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. The 
proposed change involves a one-time extension to the SG tube 
inservice inspection frequency and therefore will not give rise to 
new failure modes. In addition, the proposed change does not impact 
any other plant systems or components.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The SG tubes are an integral part of the Reactor Coolant System 
(RCS) pressure boundary that are relied upon to maintain the RCS 
pressure and inventory. The SG tubes isolate the radioactive fission 
products in the reactor coolant from the secondary system. The 
safety function of the SGs is maintained by ensuring the integrity 
of the SG tubes. In addition, the SG tubes comprise the heat 
transfer surface between the primary and secondary systems such that 
residual heat can be removed from the primary system.
    SG tube integrity is a function of the design, environment, and 
current physical condition. Extending the SG tube inservice 
inspection frequency by one operating cycle will not alter the 
function or design of the SGs. SG inspections conducted during the 
first refueling outage following SG replacement demonstrated that 
the SGs do not have an active damage mechanism, and the scope of 
those inspections significantly exceeded the scope required by the 
TS. These inspection results were comparable to similar inspection 
results for second generation alloy 690 models of replacement SGs 
installed at other plants, and subsequent inspections at those 
plants yielded results that support this extension request. The 
improved design of the replacement SGs also provides reasonable 
assurance that significant tube degradation is not likely to occur 
over the proposed operating period.

    The NRC staff has reviewed the licensee's analysis and, based on 
this

[[Page 25658]]

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

Southern Nuclear Operating Company, Inc, Docket No. 50-364, Joseph M. 
Farley Nuclear Plant, Unit 2, Houston County, Alabama

    Date of amendment request: March 31, 2003.
    Description of amendment request: The proposed amendments would 
modify Surveillance Requirement (SR) 3.4.11.1, for Farley, Unit 2 only 
by the addition of the following note that states, ``Not required to be 
performed for Unit 2 for the remainder of operating cycle 16 for 
Q2B31MOV800B.'' In addition, a temporary Technical Specification SR 
3.4.11.4 is added to provide compensatory action for this block valve 
while SR 3.4.11.1 is suspended. Further, this SR requires that power to 
the Farley, Unit 2 Power Operated Relief Valve Q2B31MOV800B be checked 
at least every 24 hours for the remainder of operating cycle 16.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change to Surveillance Requirement (SR) 3.4.11.1 
suspends the requirement to cycle test the Unit Two pressurizer 
power operated relief valve (PORV) block valve Q2B31MOV8000B for the 
remainder of operating cycle 16. This change will eliminate the 
remaining scheduled cycle tests for the PORV block valve during 
operating cycle 16. SR 3.4.11.4 is added to provide compensatory 
measures for verifying power available to the block valve at least 
every 24 hours. At the end of cycle 16, the proposed changes will no 
longer be in effect. Suspension of the cycle tests for the PORV 
block valve Q2B31MOV8000B may result in a small decrease in 
assurance that the block valve would cycle if required to isolate a 
stuck open PORV. However, experience with these valves has shown 
them to be very reliable and suspension of the remaining tests will 
not appreciably reduce reliability of the valve. There is no 
relationship between packing leakage on the PORV block valve and a 
postulated stuck open PORV. The proposed compensatory measure of 
verifying block valve power available on a 24 hour basis adds 
additional assurance that the block valve will close if demanded. 
Therefore, the probability of a previously evaluated accident 
remains acceptable is not significantly increased.
    The proposed changes do not affect the consequences of a 
previously analyzed accident since the magnitude and duration of 
analyzed events are not impacted by this change. The dose 
consequences of the proposed change are bounded by LOCA [loss-of-
coolant accident] analyses. Therefore, the consequences of a 
previously evaluated accident are unchanged.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes involve no change to the physical plant. 
They allow for suspension of the PORV block valve Q2B31MOV8000B 
cycle tests for a limited time and provide for compensatory action 
to verify power to the PORV block valve. This valve provides an 
isolation function for a postulated stuck open or leaking 
pressurizer PORV. This condition is an analyzed event since it is 
bounded by the FNP [Farley Nuclear Plant] LOCA analyses. In addition 
to the isolation function, the block valve is required to remain 
open to allow the associated PORV to function automatically to 
control reactor coolant system (RCS) pressure. These changes do not 
impact the open function of the block valve since the normal 
position is open.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The physical plant is unaffected by these changes. The proposed 
changes do not impact accident offsite dose, containment pressure or 
temperature, emergency core cooling system (ECCS) or reactor 
protection system (RPS) settings or any other parameter that could 
affect a margin of safety. The elimination of cycle testing of the 
PORV block valve Q2B31MOV8000B for the remainder of the Unit Two 
operating cycle and the addition of the proposed compensatory action 
that enhances assurance of valve operation are somewhat offsetting.

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

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of amendment request: February 26, 2003.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications Section 5.5.17, ``Containment Leakage 
Rate Testing Program,'' to reflect a one time deferral of the Type-A 
Containment Integrated Leak Rate Test (ILRT). The 10-year interval 
between ILRTs is to be extended to 15 years from the previous ILRTs 
that were completed in March 2002 for Unit 1 and March 1995 for Unit 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed Technical Specifications change does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed revision to Technical Specifications 5.5.17, 
``Containment Leakage Rate Testing Program,'' involves a one-time 
extension to the current interval for Type A containment leak 
testing. The current test interval of ten (10) years would be 
extended on a one-time basis to no longer than fifteen (15) years 
from the last Type A test. The proposed Technical Specifications 
change does not involve a physical change to the plant or a change 
in the manner in which the plant is operated or controlled. The 
reactor containment is designed to provide an essentially leak tight 
barrier against the uncontrolled release of radioactivity to the 
environment for postulated accidents. As such, the reactor 
containment itself and the testing requirements invoked to 
periodically demonstrate the integrity of the reactor containment 
exist to ensure the plant's ability to mitigate the consequences of 
an accident, and do not involve the prevention or identification of 
any precursors of an accident.
    The proposed change involves only the extension of the interval 
between Type A containment leakage tests. Type B and C containment 
leakage tests will continue to be performed at the frequency 
currently required by plant Technical Specifications. Industry 
experience has shown, as documented in NUREG-1493 [``Performance-
Based Containment Leak-Test Program''], that Type B and C 
containment leakage tests have identified a very large percentage of 
containment leakage paths and that the percentage of containment 
leakage paths that are detected only by Type A testing is very 
small. VEGP [Vogtle Electric Generating Plant] test history supports 
this conclusion. NUREG-1493 concluded, in part, that reducing the 
frequency of Type A containment leak tests to once per twenty (20) 
years leads to an imperceptible increase in risk. The integrity of 
the reactor containment is subject to two types of failure mechanism 
which can be categorized as (1) activity based and (2) time based. 
Activity based failure mechanisms are defined as degradation due to 
system and/or component modifications or maintenance. Local leak 
rate test requirements and administrative controls such as design 
change control and procedural

[[Page 25659]]

requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the reactor 
containment itself combined with the containment inspections 
performed in accordance with ASME Section XI, the Maintenance Rule, 
and the containment coatings program serve to provide a high degree 
of assurance that the containment will not degrade in a manner that 
is detectable only by Type A testing.
    2. The proposed Technical Specifications change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    The proposed revision to Technical Specifications involves a 
one-time extension to the current interval for Type A containment 
leak testing. The reactor containment and the testing requirements 
invoked to periodically demonstrate the integrity of the reactor 
containment exist to ensure the plant's ability to mitigate the 
consequences of an accident and do not involve the prevention or 
identification of any precursors of an accident. The proposed 
Technical Specifications change does not involve a physical change 
to the plant or the manner in which the plant is operated or 
controlled.
    3. The proposed Technical Specifications change does not involve 
a significant reduction in a margin of safety.
    The proposed revision to Technical Specifications involves a 
one-time extension to the current interval for Type A containment 
leak testing. The proposed Technical Specifications change does not 
involve a physical change to the plant or a change in the manner in 
which the plant is operated or controlled. The specific requirements 
and conditions of the Containment Leakage Rate Testing Program, as 
defined in Technical Specifications, exist to ensure that the degree 
of reactor containment structural integrity and leak tightness that 
is considered in the plant safety analysis is maintained. The 
overall containment leakage rate limit specified by Technical 
Specifications is maintained. The proposed change involves only the 
extension of the interval between Type A containment leakage tests. 
Type B and C containment leakage tests will continue to be performed 
at the frequency currently required by plant Technical 
Specifications.
    VEGP and industry experience strongly support the conclusion 
that Type B and C testing detects a large percentage of containment 
leakage paths and that the percentage of containment leakage paths 
that are detected only by Type A testing is small. The containment 
inspections performed in accordance with ASME Section XI, the 
Maintenance Rule, and the containment coatings program serve to 
provide a high degree of assurance that the containment will not 
degrade in a manner that is detectable only by Type A testing.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: John A. Nakoski.

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.

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

    Date of application for amendments: April 8, 2003 as supplemented 
April 22, 2003.
    Brief description of amendments: To revise, for one time only, a 
portion of Surveillance Requirement 3.5.2.3 of the Technical 
Specifications for the emergency core cooling system (ECCS). The 
revision will extend, until the refueling outage in the fall of 2003, 
the verification that the ECCS safety injection hot leg injection lines 
are full of water.
    Date of publication of individual notice in the Federal Register: 
April 16, 2003 (68 FR 18712).
    Expiration date of individual notice: May 1, 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 published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by e-mail to [email protected].

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

    Date of application for amendment: September 26, 2002.
    Brief description of amendment: The amendment revises Technical 
Specification 3.9.1, ``Refueling Equipment Interlocks,'' to allow in-
vessel fuel movement to continue if the refueling interlocks become 
inoperable. Specifically, the amendment adds Required Action A.2.1 to 
immediately block control rod withdrawal and Required Action A.2.2 to 
perform a verification that all of the control rods are fully inserted.
    Date of issuance: April 28, 2003.

[[Page 25660]]

    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 154.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 26, 2002 (67 
FR 70764).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 28, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application of amendments: December 4, 2002.
    Brief description of amendments: The amendments revised the 
Technical Specification 3.7.6 to require a minimum combined inventory 
of 155,000 gallons and remove the Condensate Storage Tank as a source 
of the combined inventory.
    Date of Issuance: April 30, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 330, 330 & 331.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 31, 2003 (68 FR 
2801).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 30, 2003.
    No significant hazards consideration comments received: No

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: December 6, 2002.
    Brief description of amendment: The amendment increased the 
surveillance interval of the local power range monitor calibrations 
from 1000 megawatt-days/ton to 200 megawatt-days/ton.
    Date of issuance: May 1, 2003.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 277.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5674).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 1, 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: August 16, 2002, as supplemented 
on March 26, April 16, and April 19, 2003.
    Brief description of amendment: The amendment modified Technical 
Specification (TS) 3/4.10.A, ``Refueling Interlocks,'' and TS 3/4.10.D, 
``Multiple Control Rod Removal,'' to provide an alternative required 
action if the refueling interlocks became inoperable during fuel 
movements in the reactor vessel. The amendment allowed fuel movements 
to continue in the reactor vessel should the refueling equipment 
interlocks become inoperable.
    Date of issuance: April 21, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 199.
    Facility Operating License No. DPR-35: Amendment revised the TSs.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75872).
    The March 26, April 16, and April 19, 2003, letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated April 21, 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: January 23, 2003, as 
supplemented February 24, and April 17, 2003.
    Brief description of amendment: This amendment modifies the Pilgrim 
Nuclear Power Station Technical Specification (TS) requirements for the 
Emergency Core Cooling System (ECCS) during shutdown conditions. The 
amendment changes the Core Spray and Low Pressure Coolant Injection 
System's TS requirements to be applicable during the Run, Startup, and 
Hot Shutdown Modes. The amendment also modifies the High Drywell 
Pressure Instrumentation TSs to require the instrumentation to be 
Operable during the Run, Startup and Hot Shutdown Modes. Unnecessary TS 
requirements are removed based on the plant's operating Mode. Other 
changes are administrative in nature.
    Date of issuance: April 22, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 200.
    Facility Operating License No. DPR-35: Amendment revised the TSs.
    Date of initial notice in Federal Register: March 18, 2003 (68 FR 
12952).
    The supplements dated February 24, and April 17, 2003, provided 
additional information that clarified the application, and did not 
expand the scope of the application or change the staff's original 
proposed no significant hazards consideration determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated April 22, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 18, 2002.
    Brief description of amendment: The amendment revises several 
Technical Specifications Limiting Conditions for Operations and 
Administrative sections to correct or clarify certain requirements and 
information.
    Date of issuance: April 23, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 157.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75871).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 23, 2003.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-237, Dresden Nuclear 
Power Station, Unit 2, Grundy County, Illinois

    Date of application for amendment: January 31, 2003, as 
supplemented March 7, 2003.
    Brief description of amendment: The amendment revises the safety 
limit

[[Page 25661]]

minimum critical power ratio for Unit 2 for two loop operation and for 
single loop operation.
    Date of issuance: April 22, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 199.
    Facility Operating License No. DPR-19: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 4, 2003 (68 FR 
10279). The supplement dated March 7, 2003, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 22, 2003.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: April 10, 2002, as supplemented 
March 10, 2003.
    Brief description of amendments: The amendments revised the 
Technical Specifications to relocate emergency diesel generator 
maintenance inspection requirements from Section 4.8.1.1.2.e.1 to the 
Technical Requirements Manual.
    Date of issuance: April 18, 2003.
    Effective date: As of the date of issuance and shall include the 
relocation of the emergency diesel generator maintenance requirements 
of Technical Specification 4.8.1.1.2.e.1 to the Technical Requirements 
Manual within 30 days.
    Amendment Nos.: 165 and 128.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 28, 2002 (67 FR 
36926). The supplement dated March 10, 2003, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 18, 2003.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: November 27, 2002.
    Brief description of amendments: These amendments deleted TS 
6.8.4.c, ``Post-accident Sampling,'' and thereby eliminated the 
requirements to have and maintain the post accident sampling system for 
Limerick Generating Station, Units 1 and 2. The amendments also 
addressed related changes to TS 6.8.4.a, ``Primary Coolant Sources 
Outside Containment.''
    Date of issuance: April 25, 2003.
    Effective date: As of date of issuance and shall be implemented 
within 180 days.
    Amendment Nos.: 166 and 129.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 21, 2003 (68 FR 
2802).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 25, 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 4, 2002.
    Brief description of amendment: This amendment revises the pressure 
temperature limits for 22- and 32-effective full power years for Perry 
Nuclear Power Plant. The June 4, 2002, application also contained a 
request for exemption from applying Appendix G of the 1995 American 
Society of Mechanical Engineers Boiler and Pressure Vessel Code and 
approval for using Code Case N-640, which permits the use of the plain 
strain fracture toughness (KIc) curve instead of the crack arrest 
fracture toughness (KIa) curve for reactor pressure vessel materials in 
determining the P-T limits.
    Date of issuance: April 29, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 127.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75878).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 29, 2003.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket No. 50-389, St. Lucie Plant, 
Unit No. 2, St. Lucie County, Florida

    Date of amendment request: October 15, 2002, as supplemented 
February 28, 2003.
    Description of amendment request: The amendment modifies the 
reactor coolant system flow rate from 363,000 gallons per minute (gpm) 
to 355,000 gpm in Technical Specifications (TSs) Table 3.3-2 and in a 
footnote for Table 2.2-1.
    Date of Issuance: April 18, 2003.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 131.
    Facility Operating License No. NPF-16: Amendment revised the TSs.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68737).
    The February 28, 2003, supplement did not affect the original 
proposed no significant hazards determination, or expand the scope of 
the request as noticed in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 18, 2003.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendments: July 23, 2002.
    Brief description of amendments: The amendments revise certain 18-
month surveillance requirements by eliminating the condition that 
testing be conducted ``during shutdown,'' or ``during the COLD SHUTDOWN 
or REFUELING MODE'' (i.e., shutdown conditions).
    Date of issuance: April 22, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 275 and 257.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 17, 2002 (67 
FR 58647).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 22, 2003.
    No significant hazards consideration comments received: No.

[[Page 25662]]

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendments: January 14, 2003.
    Brief description of amendments: The amendments modify Technical 
Specification (TS) 3.7.5.1 to add an exception to Limiting Condition 
for Operation 3.0.4 for the control room emergency ventilation system 
(CREVS). This exception allows movement of irradiated fuel assemblies 
to begin while one of the two CREVS pressurization trains is 
inoperable, provided the appropriate TS action requirements are 
implemented. The amendments are consistent with the standard TSs for 
Westinghouse plants (NUREG 1431, Revision 2, ``Standard Technical 
Specifications, Westinghouse Plants,'' dated April 30, 2001).
    Date of issuance: April 25, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 276 and 258.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 4, 2003 (68 FR 
10280).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 25, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 26, 2001, as 
supplemented by letters dated June 7 and November 22, 2002.
    Brief description of amendment: The amendment revised Section 6.0, 
``Administrative Controls,'' of the Technical Specifications (TSs) to 
clarify and relocate existing requirements, make wording improvements, 
and make the TSs consistent with the Unit 2 TSs. The revised Section 
6.0 is consistent with the ``Standard Technical Specifications for 
General Electric plants, BWR [Boiling Water Reactor]/4'' (NUREG-1433, 
Revision 2).
    Date of issuance: April 23, 2003.
    Effective date: April 23, 2003, to be implemented within 90 days of 
issuance.
    Amendment No.: 181.
    Facility Operating License No. DPR-63: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 8, 2002 (67 FR 
928).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 23, 2003.
    No significant hazards consideration comments received: No.
    Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin
    Date of application for amendment: July 26, 2002, as supplemented 
February 27, March 14, March 19, March 21 (2 letters), and April 3, 
2003.
    Brief description of amendment: The amendment revises technical 
specifications for use of Westinghouse 422 VANTAGE + nuclear fuel with 
PERFORMANCE + features.
    Date of issuance: April 4, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 167.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 3, 2002 (67 
FR 56322).
    The supplemental letters 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 4, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 19, 2002, as 
supplemented February 28, 2003.
    Brief description of amendment: The amendment relocates TS 
Surveillance Requirement (SR) 4.6.B.2, ``Reactor Vessel Temperature and 
Pressure,'' and the associated TS Bases to Section 4.2 of the Updated 
Safety Analysis Report. It also implements the Boiling Water Reactor 
Vessel and Internals Project reactor pressure vessel integrated 
surveillance program at Monticello and demonstrates compliance with the 
requirements of Title 10 of the Code of Federal Regulations, Part 50, 
Appendix H, ``Reactor Vessel Material Surveillance Program 
Requirements.''
    Date of issuance: April 22, 2003
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 135.
    Facility Operating License No. DPR-22: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 29, 2002 (67 FR 
66012).
    The supplement of February 28, 2003, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the Nuclear 
Regulatory Commission staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 22, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 8, 2002, and its supplements 
dated December 3, 2002, and March 4, 2003.
    Brief description of amendment: The amendment modifies Technical 
Specification 2.3.a, ``Emergency Core Cooling System,'' to extend the 
allowed outage time for a single low pressure safety injection pump 
from the existing 24 hours to 7 days. In addition, the word ``pump'' 
has been replaced with the word ``train.''
    Date of issuance: April 29, 2003
    Effective date: April 29, 2003, and shall be implemented within 60 
days from the date of issuance.
    Amendment No.: 217.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68740).
    The supplemental letters dated December 3, 2002, and March 4, 2003, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed or revise the 
proposed technical specification changes and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 29, 2003.
    No significant hazards consideration comments received: No.

Saxton Nuclear Experimental Corporation (SNEC) and GPU Nuclear, Inc., 
Docket No. 50-146 Saxton Nuclear Experimental Facility (SNEF), Bedford 
County, Pennsylvania

    Date of application for amendment: February 2, 2000, as 
supplemented on

[[Page 25663]]

June 23, August 11, September 18 and December 4, 2000; January 30, 
February 14, March 15 and 19, June 20, July 2 and September 4, 2001; 
and January 11 and 24, February 4, May 22 and 28, July 11, August 20, 
September 17, 23, 24, and 26, October 10, and December 16, 2002.
    Brief description of amendment: The amendment revises Amended 
Facility License No. DPR-4 for the SNEF to annotate approval of the 
SNEF License Termination Plan.
    Date of issuance: March 28, 2003.
    Effective date: Date of issuance to be implemented no later than 30 
days from the date of issuance.
    Amendment No.: 18.
    Amended Facility License No. DPR-4: Amendment added a new license 
condition to require the licensees to implement and maintain in effect 
all provisions of the approved SNEF License Termination Plan.
    Date of initial notice in Federal Register: November 29, 2000.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 28, 2003.
    No significant hazards consideration comments received: No.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendments request: November 5, 2001, as supplemented by 
letters dated October 23, 2002, and January 15, 2003.
    Brief description of amendments: The proposed amendments convert 
the current Technical Specification (TS) Section 6.0 of the STP, Units 
1 and 2, TS to the Improved Technical Specifications based on NUREG-
1431, ``Standard Technical Specification for Westinghouse Plants.''
    Date of issuance: April 24, 2003.
    Effective date: As of its date of issuance and shall be implemented 
within 6 months from the date of issuance.
    Amendment Nos.: Unit 1-151; Unit 2-139.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 5, 2002 (67 FR 
5335).
    The October 23, 2002, and January 15, 2003, supplemental letters 
provided clarifying information that was within the scope of the 
original Federal Register notice (67 FR 5335) and did not change the 
initial no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 24, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 28, 2003.
    Brief description of amendment: The amendment approves the use of 
an alternate methodology using a through-bolted connection frame to 
restore the steam generator (SG) compartment roof after replacement of 
the SGs, and a revision of the Updated Safety Analysis Report (UFSAR) 
to reflect the approval of the methodology.
    Date of issuance: April 25, 2003.
    Effective date: As of the date of issuance, to be incorporated into 
the UFSAR at the time of its next update.
    Amendment No.: 184.
    Facility Operating License No. DPR-77: Amendment revises the UFSAR.
    Date of initial notice in Federal Register: March 14, 2003 (68 FR 
12382).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 25, 2003.
    No significant hazards consideration comments received: No

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

    Date of application for amendment: February 28, 2003.
    Description of amendment: The amendment approves a revision of the 
SQN Updated Final Safety Analysis (UFSAR) to include a change to the 
methodology for connecting reinforcing steel bars during restoration of 
the Unit 1 concrete shield building dome as part of the steam generator 
replacement project. This modification to the shield building concrete 
dome is necessary to support removal of the original steam generators 
and installation of the replacement steam generators.
    Date of issuance: April 24, 2003.
    Effective date: As of the date of issuance to be incorporated into 
the UFSAR at the time of its next update.
    Amendment No.: 283.
    Facility Operating License No. DPR-77: Amendment revises the 
Operating License.
    Date of initial notice in Federal Register: March 17, 2003 (68 FR 
12718).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 24, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 8, 2003, as supplemented 
April 22, 2003.
    Brief description of amendment: The amendment revises, for one time 
only, a portion of Surveillance Requirement (SR) 3.5.2.3 of the Watts 
Bar Technical Specifications for the emergency core cooling system 
(ECCS). The revision extends, until the refueling outage in the fall of 
2003, the verification that the ECCS safety injection hot leg injection 
lines are full of water. SR 3.5.2.3 currently requires a verification 
frequency of 31 days.
    Date of issuance: May 1, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 43.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes. (68 FR 18712 dated April 16, 2001). That notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by May 16, 2003, but indicated that if the Commission makes a 
final no significant hazards consideration determination, any such 
hearing would take place after issuance of the amendment. The April 22, 
2003, letter provided clarifying information that did not change the 
initial proposed no significant hazards consideration determination or 
expand the scope of the original request.
    The Commission's related evaluation of the amendments, finding of 
exigent circumstances, and final no significant hazards consideration 
determination are contained in a Safety Evaluation dated May 1, 2003.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

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

    Date of application for amendment: October 3, 2002.
    Brief description of amendment: The amendment revises Limiting 
Condition for Operation 3.1.8, ``Physics Tests Exceptions--Mode 2,'' to 
reduce the required number of channels from four

[[Page 25664]]

to three channels for certain functions in Table 3.3.1-1, ``Reactor 
Trip System Instrumentation.''
    Date of issuance: April 21, 2003.
    Effective date: April 21, 2003, and shall be implemented within 60 
days of the date of issuance.
    Amendment No.: 154.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 26, 2002 (67 
FR 70771).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 21, 2003.
    No significant hazards consideration comments received. No.

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

    Date of amendment request: October 1, 2002.
    Brief description of amendment: The amendment revises Limiting 
Condition for Operation 3.1.8, ``Physics Tests Exceptions--Mode 2,'' to 
reduce the required number of channels from four to three channels for 
certain functions in Table 3.3.1-1, ``Reactor Trip System 
Instrumentation.''
    Date of issuance: April 21, 2003.
    Effective date: April 21, 2003, and shall be implemented within 90 
days of the date of issuance.
    Amendment No.: 151.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68746).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 21, 2003.
    No significant hazards consideration comments received: No.

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

    Brief description of amendment: The amendment revises the YNPS 
License and Technical Specifications to delete operational and 
administrative requirements that would no longer be required once the 
spent nuclear fuel has been transferred from the spent fuel pool to the 
Independent Spent Fuel Storage Installation.
    Date of issuance: April 17, 2003.
    Effective date: April 17, 2003.
    Amendment No.: 157.
    Facility Operating License No. DPR-3. Amendment revises the License 
and Technical Specifications.
    Date of initial notice in Federal Register: February 18, 2003 (68 
FR 7823).
    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.

    Dated at Rockville, Maryland, this 5th day of May 2003.

For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 03-11697 Filed 5-12-03; 8:45 am]
BILLING CODE 7590-01-P