[Federal Register Volume 68, Number 72 (Tuesday, April 15, 2003)]
[Notices]
[Pages 18269-18294]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-9026]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from April 4, 2003, through April 17, 2003. The 
last biweekly notice was published on April 1, 2003, (68 FR 15756).

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

[[Page 18270]]

different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By May 15, 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

[[Page 18271]]

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, 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].

Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power 
Plant, Unit 1 (Fermi 1), Monroe County, Michigan

    Date of amendment request: January 28, 2003, (Reference NRC-03-
0011).
    Description of amendment request: The proposed amendment will 
revise the Technical Specifications by:
    1. Section A.1, 2, 4, 8, C.1, D, E.1, H.3.b, I.5, I.7b, I.9.d have 
been previously deleted and the word ``Deleted'' used as a place marker 
to alleviate the need to renumber all sections. This request proposes 
to remove these sections and renumber as appropriate.
    2. Sections C.2 and E.2 cover the Reactor Building and Fuel and 
Repair Building Drains. This request proposes to delete the 
requirements in sections C.2 and E.2, which is all that remains in 
sections C and E. Section C, Reactor Building, and E, Fuel and Repair 
Building, will be deleted in their entirety.
    3. Added, ``Monitoring or sampling for tritium will not be required 
if the sample results have determined that tritium is not present 
during a given evolution'' in Section F. This is to clarify the intent 
of ``During other evolutions resulting in radioactive gaseous 
effluents, the effluents shall be monitored or sampled and analyzed for 
tritium and particulates.''
    4. Section H.1 and 2 cover alarms, including surveillances, allowed 
out of service time, compensatory measures and alarm readouts for 
alarms associated with water intrusion. This request proposed to delete 
these sections on water intrusion alarms.
    5. Sections H.3 and 4 cover required inspections of the facility. 
This request proposes to delete the requirement for radiation 
surveillance of the steam cleaning room access plug, which is Item c. 
of H.3, Fuel and Repair Building.
    This proposal adds the words ``(until made inactive)'' to H.3 
Reactor Building Item c. This request also proposes to delete recording 
liquid waste tank levels, which is Item c. in Section H.4.
    6. Table H-1 lists the required Fermi 1 alarms and their alarm 
points. Only water intrusion alarms are currently covered in this 
table. This request proposed to delete this alarm table.
    7. Editorial changes are included in this proposed request. In 
section I.2, the word ``employes'' will be changed to ``employees''. In 
Section I.2.b the word ``He'' will be changed to ``The Health 
Physicist''. In Section I.7 the word ``his'' will be removed from the 
following sentence, ``The Custodian may temporarily change a procedure 
by Written Order following his determination that the change does not 
constitute a significant increase in the hazards associated with the 
operation.'' In Section I.9.h the word ``usual'' will be changed to 
``unusual''.

    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 using the standards in 10 CFR 50.92(c). The licensee's 
analysis is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident.
    Removing the requirements for water intrusion monitoring, liquid 
waste tanks level recording, and building drains will not 
significantly increase the possibility of an accident as long as the 
probability of an uncontrolled sodium and water reaction is not 
significantly increased. This is accomplished by the amount of 
volume of the area in which the sodium is present where water 
intrusion is currently monitored. It would take a long period of 
time for the water intrusion to reach the sodium piping and this 
would still not increase the probability as long as the piping is 
not breached. When the piping is breached during the sodium 
abatement process, it will be completed under controlled conditions. 
Removal of the instrumentation may delay the discovery of a liquid 
spill but cannot affect the probability of the spill since it is 
only instrumentation. The consequences of an accident will not be 
increased because the previously analyzed accident accounts for all 
of the radioactive material contained within the liquid waste tanks 
and primary sodium to be released. This change will not increase the 
amount of radioactive material. The editorial changes, steam 
cleaning room plug radiation survey deletion, or the clarification 
made to gaseous effluent monitoring for tritium will not 
significantly increase the probability or consequences of an 
accident, because they have no impact on how any systems are 
operated or what systems are removed from the facility.
    2. The proposed change does not create the possibility of a new 
or different accident from any previously evaluated.
    Removing the requirements for water intrusion monitoring and 
liquid waste tanks level recording will not create the possibility 
of a new or different accident from any previously evaluated. The 
accidents these systems monitor for have already been analyzed for, 
including a release of the radioactive sodium during a sodium and 
water reaction and the release of the entire contents of the liquid 
waste tanks. Removing the building drains requirements will not 
cause a different type of accident since the drains only affect 
where liquid flows. Where liquid flows cannot cause an accident 
unless the drains place water where it does not belong. This can 
only impact a liquid water release or sodium accident. The editorial 
changes, survey deletion, and the clarification made to gaseous 
effluent monitoring for tritium will not create the possibility of a 
new or different accident, since they do not introduce any new modes 
of operation of facility equipment.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The removal of the requirements for water intrusion monitoring, 
liquid waste tanks level recording, and building drains may slightly 
reduce the margin of safety, but not significantly. Removing them 
does not in itself introduce water into the sodium containing 
systems. Nor does removing them allow for an unmonitored discharge 
of any radioactive effluents. Discharges are still controlled by 
Section C of the proposed amendment to the Technical Specifications. 
The decommissioning project is now ongoing and the facility no 
longer normally vacant as it was during the initial time following 
facility retirement. In addition, the calculated consequences of 
releasing the radioactive material are small and within 10 CFR 20 
limits. The editorial changes or survey deletion will not 
significantly reduce a margin of safety, because the survey is of a 
floor plug that has been removed from the entrance to an area and 
has no function. The

[[Page 18272]]

clarification made to gaseous effluent monitoring for tritium will 
not significantly reduce a margin of safety since tritium monitoring 
is still required for evolutions involving sodium processing and 
pipe cutting, and during other activities, unless results have 
determined tritium is not present during a given evolution.

    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, NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John Flynn, Esquire, Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Section Chief: Claudia M. Craig.
    Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, 
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
    Date of amendment request: November 25, 2002.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TS) for the Ventilation Filter Testing 
Program (VFTP), Annulus Ventilation System (AVS), Auxiliary Building 
Filtered Ventilation Exhaust System (ABFVES), Fuel Handling Ventilation 
Exhaust System (FHVES), and Control Room Area Ventilation System 
(CRAVS), and containment penetrations.
    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 following discussion is a summary of the evaluation of the 
changes contained in this proposed amendment against the 10 CFR 
50.92(c) requirements to demonstrate that all three standards are 
satisfied. A no significant hazards consideration is indicated if 
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.

First Standard

    Does operation of the facility in accordance with the proposed 
amendment involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    This licensee amendment request proposes amendments to the 
system TS and/or Bases and/or VFTP TS requirements for the AVS, 
ABFVES, FHVES, and CRAVS. It also proposes amendments to the TS and 
Bases for Containment Penetrations. The AVS is in standby during 
normal plant operations and operates only following a Safety 
Injection signal or during a test. It is not an accident initiator. 
The ABFVES is in operation during normal plant operations. However, 
the ABFVES is not used in direct support of any phase of power 
generation or conversion or transmission, shutdown cooling, fuel 
handling operations, or processing of radioactive fluids. Therefore, 
it is not an accident initiator. The FHVES is utilized to support 
fuel handling operations when moving recently irradiated fuel. It is 
not an accident initiator. The CRAVS operates during normal plant 
operations. However, it is not an accident initiator (the CRAVS 
being defined so as to exclude equipment that maintains an 
appropriately low temperature in the control room). The status of 
containment penetrations is required to be controlled so as to 
minimize the consequences of a fuel handling accident or a weir gate 
drop accident. The containment penetrations by themselves are not 
accident initiators. No accident initiators are associated with the 
changes proposed in this license amendment request. For these 
reasons, operation of the facility in accordance with this proposed 
amendment does not involve a significant increase in the probability 
of any accident previously evaluated.
    In support of the proposed amendment, an analysis has been 
performed to determine the radiological consequences of the design 
basis LOCA [loss-of-coolant accident] at Catawba Nuclear Station. 
The analysis made use of the Alternative Source Term (AST) 
methodology and in general conformed to the regulatory positions of 
Regulatory Guide 1.183, [``Alternative Radiological Source Terms for 
Evaluating Design Basis Accidents at Nuclear Power Reactors,'' 
(ML003716792)
    (Draft DG1081 Issued December 1999)] and the draft regulatory 
positions of DG-1111. Total Effective Dose Equivalent (TEDE) 
radiation doses at the Exclusion Area Boundary (EAB), boundary of 
the Low Population Zone (LPZ), and to the control room operators 
were calculated and found to be acceptable.
    TEDE's have been estimated from the radiation doses with the 
current analysis (reported in the UFSAR [Updated Final Safety 
Analysis Report]) using the guidelines of Regulatory Guide 1.183 
modified as reported in Appendix A of Attachment 3 [of the 
licensee's submittal dated November 25, 2002]. These TEDE's are 
compared to the limiting TEDE's from the proposed analysis as 
follows:

                 TEDE's Following the Design Basis LOCA
------------------------------------------------------------------------
                                                       TEDE'S (Rem)
                    Location                     -----------------------
                                                     UFSAR     Proposed
------------------------------------------------------------------------
EAB.............................................        9.95        7.21
LPZ.............................................        1.90        3.97
Control Room....................................        1.57        2.65
------------------------------------------------------------------------

    The new value for the control room TEDE radiation dose is higher 
than the TEDE radiation dose equivalent to the radiation doses 
currently reported in the UFSAR. However, the limiting control room 
TEDE radiation dose reported in this submittal is lower than the 
acceptance criterion by 47%. The new LPZ TEDE radiation dose is 
higher than the equivalent TEDE radiation dose currently 
represented. On the other hand, the margin to the acceptance 
criterion is 84%. The TEDE radiation doses newly computed at the EAB 
for the design basis LOCA is lower than the corresponding equivalent 
EAB TEDE radiation dose currently represented in the UFSAR. The 
margin in the EAB TEDE radiation dose to the guideline value is 71%. 
In all cases, there is significant margin between the newly 
calculated post-LOCA TEDE radiation doses and the corresponding 
regulatory guideline values. In the sense that the margins to the 
germane regulatory guideline values are still large, the new values 
of TEDE radiation doses are comparable to the equivalent TEDE 
associated with the post-LOCA radiation doses currently listed in 
the UFSAR. Therefore, the proposed amendment is determined to not 
result in a significant increase in accident consequences.
    The changes proposed to the TS for Containment Penetrations are 
editorial in nature and will have no effect upon accident 
consequences.
    The changes proposed to the VFTP TS for the AVS, ABFVES, and 
FHVES will not result in a significant increase in any accident 
consequences. The changes to make the penetration values for Unit 2 
consistent with Unit 1 for the AVS, ABFVES, and FHVES are acceptable 
because the appropriate safety factors as delineated in the 
applicable regulatory guideline documents are still maintained. The 
change to the flowrate specified for the ABFVES is consistent with 
the design basis operation of this system. Also, the editorial 
changes proposed to the VFTP TS will have no impact on any 
accidents.
    Operation of the facility in accordance with the proposed 
amendment does not involve a significant increase in the 
consequences of an accident previously evaluated.

Second Standard

    Does operation of the facility in accordance with the proposed 
amendment create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    This proposed amendment does not involve addition, removal, or 
modification of any plant system, structure, or component. These 
changes will not affect the operation of any plant system, 
structure, or components as directed in plant procedures.

[[Page 18273]]

    The analysis performed in support of this license amendment 
request, together with the analyses of the design basis fuel 
handling accident and weir gate drop reported in previously 
submitted and NRC approved license amendment requests, includes full 
scope implementation of AST methodology. This analysis does not 
represent any change in the post-accident operation of any plant 
system, structure, or component.
    Operation of the facility in accordance with this amendment does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.

Third Standard

    Does operation of the facility in accordance with the proposed 
amendment involve a significant reduction in the margin of safety?
    Margin of safety is related to confidence in the ability of 
fission product barriers to perform their design functions following 
any of their design basis accidents. These barriers include the fuel 
cladding, the Reactor Coolant System, and the containment. The 
performance of these barriers either during normal plant operations 
or following an accident will not be affected by the changes 
associated with the license amendment request.
    The AVS is associated with the containment fission product 
barrier. Its post-accident operation will not be affected by 
implementation of the amendment to its TS. The operation of the 
ABFVES either during normal plant operations or following an 
accident will not be affected by implementation of the amendment to 
its TS. The operation of the FHVES either during normal plant 
operations or following an accident will not be affected by 
implementation of the amendment to its TS. The operation of the 
CRAVS either during normal plant operations or following an accident 
will not be adversely affected by the proposed changes to its TS 
Bases. The operation of Containment Penetrations following an 
accident will not be adversely affected by the proposed change to 
its TS.
    As noted, an analysis of radiological consequences of the design 
LOCA at Catawba Nuclear Station has been performed in support of 
this license amendment request. The design basis LOCA scenarios were 
selected based on extensive evaluations of Catawba, its design 
basis, and its anticipated response to a design basis LOCA. Credit 
was taken only for safety related systems, structures, and 
components in simulating the mitigation of radiological consequences 
of the LOCA. Limiting values were taken for performance 
characteristics of the Class 1E systems modeled in the analysis. The 
radiological consequences (TEDE radiation doses at the EAB, LPZ, and 
in the control room) are within the regulatory guideline values with 
significant margin.
    The changes proposed to the VFTP TS for the AVS, ABFVES, and 
FHVES will not result in a significant reduction in the margin of 
safety. These changes are supported by regulatory guidance 
documents, and are consistent with existing system operation. Also, 
the editorial changes proposed to the VFTP TS will not have any 
impact on safety.
    Operation of the facility in accordance with the proposed 
amendment does not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, and Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, located in Mecklenburg County, North 
Carolina and York County, South Carolina

    Date of amendment request: November 20, 2002, as supplemented 
January 21, 2003.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) for REQUIRED ACTIONS requiring 
suspension of operations involving positive reactivity additions and 
various NOTES that preclude reduction in boron concentration. The 
proposed changes revise these REQUIRED ACTIONS and NOTES to limit the 
introduction of positive reactivity such that the required margin to 
criticality, the shutdown margin and refueling boron concentration 
limits will still be satisfied. The licensee stated that the changes 
are consistent with the Technical Specification Task Force (TSTF) 
traveler number 286, Revision 2. Associated changes are also proposed 
for the TS Bases. Basis for proposed no significant hazards 
consideration determination: As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    The following discussion is a summary of the evaluation of the 
change contained in this proposed amendment against the 10 CFR 50.92 
(c) requirements to demonstrate that all three standards are 
satisfied. A ``no significant hazards consideration'' is indicated 
if 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.

First Standard

    The proposed changes do not involve any physical alteration of 
plant systems, structures, or components. The proposed changes 
revise ACTIONS in the Catawba Nuclear Station (CNS) and McGuire 
Nuclear Station (MNS) Technical Specifications (TS) that require 
suspending operations involving positive reactivity additions and 
several Limiting Condition for Operation (LCO) Notes that preclude 
reduction in boron concentration. The change revises these ACTIONS 
and LCO Notes to limit the introduction of reactivity such that the 
required SHUTDOWN MARGIN (SDM) or refueling boron concentration will 
still be satisfied. The proposed change ensures that the reactivity 
condition [keff] specified in mode definition, the SDM of 
LCO 3.1.1 and minimum boron concentration requirements of LCO 3.9.1 
are met. Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated in the updated final safety analysis 
report (UFSAR) because the accident analysis assumptions and initial 
conditions will continue to be maintained.

Second Standard

    The proposed changes do not involve any physical alteration of 
plant systems, structures, or components. The proposed changes, 
which allow positive reactivity additions that do not result in the 
SDM or the refueling boron concentration being exceeded, do not 
introduce new failure mechanisms for system structures, or 
components not already considered in the UFSAR. Therefore, the 
possibility of a new or different kind of accident from any accident 
previously evaluated is not created because no new failure 
mechanisms or initiating events have been introduced.

Third Standard

    The proposed changes do not involve a significant reduction in a 
margin of safety because the ability to make the reactor subcritical 
and maintain it subcritical during all operating conditions and 
modes of operation will be maintained. The margin of safety is 
defined by the SDM of LCO 3.1.1 and minimum boron concentration 
requirements of LCO 3.9.1. The proposed changes do not affect these 
operating restrictions and the margin of safety, which assures the 
ability to make and maintain the reactor subcritical, is not 
affected.

    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: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

[[Page 18274]]

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, and Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, located in Mecklenburg County, North 
Carolina and York County, South Carolina

    Date of amendment request: January 31, 2003.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) to incorporate an asymmetrical 
ice mass distribution within the ice condenser containment (ICC) by 
specifying revised safety analysis ice mass quantity requirements for 
three specific radial zones of the ice bed. Associated changes to the 
Bases were also proposed.
    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:

    Duke Energy Corporation (Duke) has concluded that operation of 
Catawba Nuclear Station (CNS) Units 1 & 2, and McGuire Nuclear 
Station (MNS) Units 1 & 2, in accordance with the proposed changes 
to the Technical Specifications (TS) does not involve a significant 
hazards consideration. Duke's conclusion is based on its evaluation, 
in accordance with 10 CFR 50.91(a)(1), of the three standards set 
forth in 10 CFR 50.92(c).
    A. The Proposed Change Does Not Involve a Significant Increase 
In The Probability or Consequences Of An Accident Previously 
Evaluated.
    The only analyzed accidents of possible consideration in regards 
to changes potentially affecting the ice condenser are a loss of 
coolant accident (LOCA) and a high energy line break (HELB) inside 
containment. However, the ice condenser is not postulated as being 
the initiator of any LOCA or HELB. That is because it is designed to 
remain functional following a design basis earthquake, and the ice 
condenser does not interconnect or interact with any systems that 
interconnect or interact with the Reactor Coolant or Main Steam 
Systems. Since these proposed changes do not result in, or require, 
any physical change to the ice condenser that could introduce an 
interaction with the Reactor Coolant or Main Steam Systems, then 
there can be no change in the probability of an accident previously 
evaluated.
    Regarding consequences of analyzed accidents, the ice condenser 
is an engineered safety feature designed, in part, to limit the 
containment sub-compartment and containment vessel pressure 
immediately following the initiation of a LOCA or HELB. Conservative 
sub-compartment and containment pressure analysis [based on the 
proposed changes] shows these criteria will be met if the total ice 
mass within the ice bed is maintained in accordance with the DBA 
[Design Basis Accident] analysis; therefore, the proposed TS SR 
[Surveillance Requirement] changes of these requirements will not 
increase the consequences of any accident previously evaluated.
    Thus, based on the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    B. The Proposed Change Does Not Create The Possibility Of A New 
Or Different Kind Of Accident From Any Accident Previously 
Evaluated.
    As previously described, the ice condenser is not postulated as 
being the initiator of any design basis accident. The proposed 
changes do not impact any plant system, structure or component that 
is an accident initiator. The proposed TSs and TS Bases changes do 
not involve any hardware changes to the ice condenser or other 
change that could create any new accident mechanisms. Therefore, 
there can be no new or different accidents created from those 
already identified and evaluated.
    C. The Proposed Change Does Not Involve A Significant Reduction 
In A Margin Of Safety.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. The performance of the fuel cladding and the reactor coolant 
system will not be impacted by the proposed changes. The Application 
provides a description of additional sub-compartment and containment 
pressure response analysis that has been performed. This analysis 
demonstrates that containment will remain fully capable of 
performing its design function with implementation of the proposed 
changes. Therefore, no safety margin will be significantly impacted.
    Ice Condenser plant historical operating experience has shown 
that the condition of the ice condenser can be ensured to be fully 
capable of performing its specified safety functions with performing 
ice mass verifications and ice mass distribution SRs on an 18 month 
frequency. The request to increase the MNS [McGuire] surveillance 
interval from 9 months to 18 months will provide performance of ice 
mass verification at the end of the fuel cycle, which will verify 
that the maintenance program is effective in maintaining the ice 
mass for the entire fuel cycle. Duke's utilization of the data from 
previous performance of TS required ice mass inspections, and 
additional inspection beyond these requirements, has enabled the 
development of a maintenance program that is reliably predictive 
regarding the specific operating characteristics of each [of] the 
ice beds at Catawba and McGuire Nuclear Stations. This maintenance 
program reliably predicts sublimation and determines which ice 
baskets to replenish prior to beginning a new 18 months operating 
cycle. An ice mass surveillance performed at the conclusion of the 
18 month frequency in an as-found condition verifies that the 
maintenance program is restoring the ice bed operating cycle to 
maintain the ice mass quantity and distribution requirements for 
performance of the intended safety functions.

    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: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

Duke Energy Corporation, Docket No. 50-370, McGuire Nuclear Station, 
Unit 2, Mecklenburg County, North Carolina

    Date of amendment request: January 31, 2003.
    Description of amendment request: The proposed amendment would 
authorize the licensee to change the Updated Final Safety Analysis 
Report (UFSAR) to describe a process for the intentional puncture of an 
irradiated fuel rod in order to transfer the fuel rod gap gasses to a 
collection chamber, and then straighten the fuel rod for storage in a 
broken rod capsule.
    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:

    Duke Energy has evaluated whether or not a significant hazards 
consideration is involved with the proposed amendment by focusing on 
the three standards set forth in 10 CFR 50.92, ``Issuance of 
amendment,'' as discussed below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The bent rod, located in the McGuire Unit 2 spent fuel pool, has 
no interfaces with any primary system, secondary system, or power 
transmission system. All work will be performed in the spent fuel 
pool, with the bent rod located under approximately 23 feet of 
water. None of the systems listed above are modified by the 
activity. No accident initiator or accident mitigation systems, for 
any UFSAR [Updated Final Safety Analysis Report] Chapter 15 
accidents, other than fuel handling accidents, are affected with 
this proposed procedure for degassing and straightening of the 
irradiated Mk-BW fuel rod. For these reasons, the activity does not 
involve an increase in the probability of an accident previously 
evaluated.
    This evolution is bounded by the UFSAR Chapter 15 dropped fuel 
assembly fuel handling accident inside the fuel handling building. 
This accident assumes that the postulated accident occurs 100 hours 
after reactor shutdown, the fuel assembly had 60 GWD/MTU [Gigawatt 
Days/Metric Ton Uranium] burnup, all rods in one fuel

[[Page 18275]]

assembly are ruptured, and the assembly damaged has the highest 
peaking factor. The resultant Exclusion Area Boundary doses for the 
UFSAR Chapter 15 accident are 0.8 Rem Whole Body and 9.1 Rem 
Thyroid.
    For the planned evolution, the cladding on only one rod will be 
breached and the fission product gas contained. This evolution will 
occur approximately ten years after reactor shutdown. The fuel rod 
burnup is only 20.46 GWD/MTU, and the fuel pin peaking factor is 
1.28. Some accident mitigation will be provided by the fuel building 
ventilation system filters, although the majority of the activity 
will be from Kr-85, a noble gas, which is unaffected by these 
filters. The highest potential dose occurs to a worker in the fuel 
building, with whole body doses of less than 3 mRem and a thyroid 
dose of less than 3E-11 mRem. Doses at the Exclusion Area Boundary 
are trivial.
    Should the gas container fail, the offsite activity release and, 
as such, the consequences of this accident will be less than any 
previously evaluated. Analyses have been performed to determine 
upper bounds for the source term, the offsite doses, and the control 
room dose. Both the source term and doses were found to be 
significantly lower than the results of the corresponding design 
basis analyses.
    For the above reasons, it is determined that the intentional 
degassing of the Mk-BW fuel rod does not involve a significant 
increase in either the probability or 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?
    As discussed above, no ``accident initiators'' are affected by 
the proposed activity. The planned evolution is bounded by the 
dropped fuel assembly fuel handling accident inside the fuel 
handling building. The fuel rod straightening and degassing tools 
are no heavier than other fuel handling tools utilized in the spent 
fuel pool during routine operations. A safety tray will be placed on 
top of the racks and below the work area to capture any falling 
debris during the operation. Also a mockup operation will be 
performed at the Framatome facilities to identify and correct any 
deficiencies in the tools and processes.
    For these reasons, the activity will not create the possibility 
of a new or different type of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Margin of safety is associated with the confidence in the 
ability of the fission product barriers (the fuel and fuel cladding, 
the reactor coolant system pressure boundary, and the containment) 
to limit the level of radiation doses to the public. The proposed 
degassing of the fuel rod will intentionally breach the fuel rod 
cladding, but the fuel rod gap gasses will be captured in a 
collection chamber for holdup and later controlled release.
    This evolution will occur beyond a nine year cooling and 
isotopic decay period. The level of activity in the fuel rod is very 
low compared to the level of activity associated with the postulated 
fuel handling accident; the only significant activity remaining is 
approximately 10 Ci [Curies] of Krypton 85. The bent rod will be 
maintained under 23 feet of water. Should the collection chamber 
fail, and the fuel rod gap gas activity released, the highest 
potential dose occurs to a worker in the fuel handling building, 
with whole body doses of less than 3 mRem, and a thyroid dose of 
less than 3E-11 mRem. For this reason, the resulting dose to the 
public is inconsequential. Both offsite doses and doses to the 
control room were found to be small compared to the limits of 10 CFR 
100 and GDC 19. For these reasons, the activity does not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: March 14, 2003.
    Description of amendment request: The licensee requests 
modification of the River Bend Technical Specifications to revise 
several of the Surveillance Requirements (SRs) pertaining to testing of 
the Division 1 and 2 standby diesel generators (DGs). The proposed 
change would modify specific restrictions associated with these SRs 
that prohibit performing required testing in Modes 1 and 2. The 
affected SRs are SR 3.8.1.9 and SR 3.8.1.10.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The DG and its associated emergency loads are accident 
mitigating features, not accident initiating equipment. Therefore, 
there will be no impact on any accident probabilities by the 
approval of the requested amendment.
    The design of plant equipment is not being modified by these 
proposed changes. As such, the ability of the DG to respond to a 
design basis accident will not be adversely impacted by these 
proposed changes. The capability of the DG to supply power in a 
timely manner will not be compromised by permitting performance of 
DG testing during periods of power operation. Additionally, limiting 
testing to only one DG at a time ensures that design basis 
requirements for backup power is met, should a fault occur on the 
tested DG. Therefore, there would be no significant impact on any 
accident consequences.
    Based on the above, the proposed change to permit certain DG 
surveillance tests to be performed during plant operation will have 
no effect on accident probabilities or consequences. 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.
    No new accident causal mechanisms would be created as a result 
of NRC [U.S. Nuclear Regulatory Commission] approval of this 
amendment request since no changes are being made to the plant that 
would introduce any new accident causal mechanisms. Equipment will 
be operated in the same configuration with the exception of the 
plant mode in which the testing is conducted. This amendment request 
does not impact any plant systems that are accident initiators; 
neither does it adversely impact any accident mitigating systems.
    Based on the above, implementation of the proposed changes would 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. The proposed changes to the testing requirements for the DG 
do not affect the operability requirements for the DG, as 
verification of such operability will continue to be performed as 
required. Continued verification of operability supports the 
capability of the DG to perform its required function of providing 
emergency power to plant equipment that supports or constitutes the 
fission product barriers.
    Consequently, the performance of these fission product barriers 
will not be impacted by implementation of this proposed amendment.
    In addition, the proposed changes involve no changes to 
setpoints or limits established or assumed by the accident analysis. 
On this and the above basis, no safety margins will be impacted.
    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

[[Page 18276]]

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: February 27, 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. 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 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 (FR) 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 FR on March 20, 2002 (67 FR 13027). The licensee 
affirmed the applicability of the following NSHC determination in its 
application dated February 27, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

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

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

    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radioisotopes 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: Richard J. Laufer.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of amendment request: March 20, 2003.
    Description of amendment request: This proposed change reflects an 
expanded operating domain for Vermont Yankee Nuclear Power Station (VY) 
resulting from the proposed implementation of the Average Power Range 
Monitor, Rod Block Monitor Technical Specifications/Maximum Extended 
Load Line Limit Analysis (ARTS/MELLLA).

[[Page 18277]]

    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration. The NRC staff has 
reviewed the licensee's analysis against the standards of 10 CFR 
50.92(c). The NRC staff's review is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change involves allowing VY to operate in an 
expanded operating domain. Physical changes provide for enhanced 
instrument performance or were the result of safety analyses that 
support mitigation of design bases accidents. There are no changes 
to radioactive source terms or release pathways. The proposed change 
does not result in any significant change in the availability of 
logic systems or safety-related systems themselves. Required 
protective functions will be maintained. The proposed change does 
not degrade plant design, operation, or the performance of any 
safety system assumed to function in the accident analysis.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility for a new or different kind of 
accident from any previously evaluated.
    The proposed change, which allows VY to operate in an expanded 
operating domain, does not introduce any new accidents or failure 
mechanisms because the change and the effects on existing 
structures, systems and components have been evaluated and found to 
not have any adverse effects. The proposed change will not 
substantially impose new requirements or eliminate any existing 
requirements.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident than those previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change, which allows VY to operate in an expanded 
operating domain, does not alter the manner in which safety limits, 
limiting safety system settings, or limiting conditions for 
operation are determined. There is no impact on the conclusions of 
any safety analysis. The proposed change does not involve any 
increase in calculated off-site dose consequences. The performance 
of equipment will not be significantly affected.
    Therefore, there is no significant reduction in the margin of 
safety as a result of this proposed change.

    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: January 31, 2003.
    Description of amendment request: The proposed amendments would 
change the Technical Specifications (TS) allowable values (AVs) for 
isolation condenser system isolation Function 4.a, Steam Flow-High, and 
Function 4.b, Return Flow-High.
    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 changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed TS changes support the replacement of a 
differential pressure switch with a functionally equivalent 
differential pressure switch. Since there are no functional changes 
and no change in analytical limits, there is no significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Additionally, these changes will not increase the consequences 
of an accident previously evaluated because the proposed changes do 
not adversely impact structures, systems, or components. 
Furthermore, there will be no change in the types or significant 
increase in the amounts of any effluents released offsite as a 
result of the proposed change.
    In summary, the proposed changes do not involve a significant 
increase in the probability of consequences of an accident 
previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The change does not adversely impact the manner in which the 
instrument will operate under normal and abnormal operating 
conditions. Therefore, these changes provide an equivalent level of 
safety and will not create the possibility of a new or different 
kind of accident from any accident previously evaluated. The changes 
in allowed values do not affect the current safety analysis 
assumptions. Therefore, these changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed changes do not affect the probability of failure or 
availability of the affected instrumentation. The revised AVs do not 
affect the analytical limits assumed in the safety analyses for 
actuation of instrumentation. Therefore, the proposed changes do not 
result in a 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 
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.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of amendment request: February 17, 2003.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (ITS) 3.6.3 ``Containment Isolation 
Valves,'' to allow verification by administrative means of isolation 
devices in high radiation areas, and isolation devices that are locked, 
sealed or otherwise secured. The specific Conditions and Surveillance 
Requirements (SR) in ITS 3.6.3 that will be affected are: (1) Condition 
A--Required Action A.2, (2) Condition B--Required Actions B.1 and B.2, 
(3) Condition C--Required Action C.2, and (4) SR 3.6.3.3 and SR 
3.6.3.4. The licensee stated that the changes are consistent with the 
NUREG-1430, ``Standard Technical Specifications: Babcock and Wilcox 
Plants,'' Revision 2, and Standard Technical Specification Task Force 
(TSTF) Traveler TSTF-440. Associated changes are also proposed for the 
ITS Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does not involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed License Amendment Request (LAR) will revise the 
position verification requirements for manual containment isolation 
devices that are locked, sealed, or otherwise secured in the closed 
position. The proposed changes will allow the use of administrative 
controls to verify the position of these types of devices when they 
are being used to meet the Required Actions of ITS 3.6.3 Condition 
A, Condition B or Condition C, and will exclude these valves from 
Surveillance Requirement (SR) 3.6.3.3 and

[[Page 18278]]

SR 3.6.3.4 physical position verification requirements.
    The design function of the affected containment isolation 
valves, and the initial conditions for accidents that require these 
valves to be closed, will not be affected by the proposed changes. 
Therefore, the changes will not increase the probability or 
consequences of an accident previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously analyzed.
    The proposed license amendment will revise the position 
verification requirements for manual containment isolation devices 
that are locked, sealed, or otherwise secured in the closed 
position.
    No changes to the actual position/status of these valves are 
proposed by this amendment. The proposed amendment will not result 
in changes to the design, physical configuration or operation of the 
plant. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does not involve a significant reduction in the margin of 
safety.
    Changes to the position verification requirements of normally 
closed manual containment isolation valves that are locked, sealed, 
or otherwise secured do not change the position/status of these 
valves. The proposed amendment does not impact the ability of these 
valves to perform their design function of controlling containment 
leakage rates during design basis radiological accidents. Therefore, 
the proposed amendment does not result in a reduction of 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602-1551.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: March 11, 2003.
    Description of amendment request: The proposed amendment would 
change the operating license to authorize the licensee to revise the 
updated final safety analysis report (UFSAR) by deleting a footnote 
stating that the Nuclear Regulatory Commission (NRC) does not endorse 
the reactor building crane as single-failure-proof.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    For heavy load handling associated with the spent fuel pool, 
Section 5.1.4(2) of NUREG-0612 states ``The effects of heavy load 
drops in the reactor building should be analyzed to show that the 
evaluation criteria of Section 5.1 are satisfied.''
    An alternative to this is Section 5.1.4(1): ``The reactor 
building crane, and associated lifting devices used for handling of 
* * * heavy loads, should satisfy the single-failure-proof 
guidelines of Section 5.1.6 of this report.''
    The upgraded crane and handling systems satisfy the guidelines 
of Section 5.1.6. The evaluation criteria of NUREG-0612, Section 5.1 
are met with a single-failure-proof crane that satisfies the 
guidelines of Section 5.1.6, or consequence analysis that satisfies 
Section 5.1.4(2).
    Section 5.2 of NUREG-0612 states that an evaluation of fault 
trees shows that: ``(1) The likelihood for unacceptable consequences 
in terms of excessive releases of gap activity or potential for 
criticality due to accidental dropping of postulated heavy loads 
after implementation of the guidelines of Section 5.1 is very low; 
and (2) The potential for unacceptable consequences is comparable 
for any of the alternatives evaluated for fault trees, indicating 
the relative equivalency between alternatives.''
    Since the NRC fault tree evaluation shows that the potential for 
unacceptable consequences is comparable for the two alternatives in 
Section 5.1.4 of NUREG-0612, the proposed request does not 
significantly change the potential for unacceptable consequences to 
the plant in conducting heavy load handling above the spent fuel 
pool. The probability of a load drop accident caused by use of the 
reactor building crane has been reduced to where it is so small to 
be considered not credible within regulatory accepted standards. The 
reason for this is attributed to the following:
    (a) The reactor building crane is single-failure-proof. In 1985, 
the DAEC [Duane Arnold Energy Center] Reactor Building Crane was 
modified to meet the requirements of NUREG-0554 ``Single Failure 
Proof Cranes for Nuclear Power Plants.'' The design of the Ederer 
hoist and trolley system was evaluated in a Staff SER [Safety 
Evaluation Report] of the Generic Licensing Topical Report EDR-1, 
Rev. 3, for Ederer's Nuclear Safety-Related Extra Safety and 
Monitoring (X-SAM) Cranes, dated August 3, 1983.
    (b) The rigging used with the crane will be single-failure-proof 
per Section 5.1.6 of NUREG-0612.
    (c) The requirements of NUREG-0612 Phase 1 have been 
implemented. The NRC provided a Safety Evaluation (SE) and Technical 
Evaluation Report (TER) by letter dated June 12, 1984 that concluded 
that the guidelines of NUREG-0612, Sections 5.1.1 and 5.3 had been 
satisfied and that Phase I of this issue for the DAEC was 
acceptable.
    Therefore, this proposed change will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The crane has been upgraded to meet single-failure-proof 
requirements in accordance with the applicable provisions of NUREG-
0612 and NUREG-0554. The use of a single-failure-proof crane with 
rigging and procedures that implement the requirements of NUREG-0612 
assures that a cask drop is not credible. The loading on the single-
failure-proof crane will not exceed the design rated load of the 
crane.
    Rigging for critical loads will meet NUREG-0612 requirements for 
single-failure-proof handling systems whenever a critical load is to 
be lifted over safety related equipment, or over the spent fuel 
pool, or over the cask when it is in the reactor building and loaded 
with fuel. When a cask is loaded on the crane hook, the crane 
trolley and bridge movements will be maintained within well defined 
limits of operation.
    The loading conditions, load combinations, allowable stress 
limits, and methods of analysis used in the evaluations are 
consistent with the current licensing basis for the DAEC and NRC 
approved methods.
    Therefore, this proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    In 1985, the reactor building crane was upgraded to single-
failure-proof in compliance with NUREG-0554. The upgraded crane and 
handling system is in compliance with NUREG-0612, Sections 5.1.1 and 
5.1.6. The NRC in NUREG-0612, Section 5.2 documented their review of 
the potential consequences of a load drop when handled by a single-
failure-proof crane using single-failure-proof rigging compared with 
other alternatives and concluded as follows: ``The likelihood for 
unacceptable consequences in terms of excessive releases of gap 
activity or potential for criticality due to accidental dropping of 
postulated heavy loads after implementation of the guidelines of 
Section 5.1 is very low.''
    This means that a load drop is considered to be unlikely within 
regulatory accepted standards when the load is handled by a single-
failure-proof crane and handling system, and performed in accordance 
with Section 5.1 of NUREG-0612. A single-failure-proof crane design 
incorporates the applicable design basis event that in this case is 
a seismic event. A load drop is of such low probability that it is 
considered unlikely when it is handled with the reactor building 
crane since the crane and its handling systems satisfy the NUREG-
0612 criteria for a single-failure-proof crane. Therefore, any load 
lifted over the spent fuel pool using the reactor building crane has 
a very low probability of falling into the spent fuel pool

[[Page 18279]]

accidentally or as a result of a design basis event.
    Therefore, this proposed amendment will not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Alvin Gutterman, Morgan Lewis, 1111 
Pennsylvania Avenue NW., Washington, DC 20004.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: January 29, 2003.
    Description of amendment request: The proposed amendment would 
change the drywell leakage and sump monitoring detection section of the 
Technical Specifications (TSs). These proposed changes clarify the 
definitions and restructure the coolant leakage section of the TSs and 
revise unidentified leakage and total leakage requirements. The 
revisions add a TS Limiting Condition for Operation for leakage-
detection instrumentation being inoperable. This request supercedes the 
Nuclear Management Company's license amendment request of October 8, 
2002, as supplemented November 8, 2002, which was previously noticed in 
the Federal Register on October 17, 2002 (67 FR 64144).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed Technical Specification changes do not introduce 
new equipment or new equipment operating modes, nor do the proposed 
changes alter existing system relationships. Additionally, the 
proposed changes do not affect any accident previously evaluated in 
the Monticello Updated Safety Analysis Report (USAR). The changes 
simply redefine the parameters for evaluation of leakage in the 
drywell. The evaluation criteria for drywell leakage have been 
refocused into the areas that are most susceptible to IGSCC 
[intergranular stress corrosion cracking]. Consequently, the 
probability of an accident previously evaluated is not significantly 
increased.
    The equipment referenced in the proposed changes is still 
required to monitor the reactor coolant system operational leakage 
to ensure appropriate action is taken before the integrity of the 
reactor coolant pressure boundary is impaired. As a result, 
operation of the facility with the proposed changes will continue to 
meet the licensing basis and applicable guidelines. As such, the 
consequences of any accident previously evaluated are not 
significantly affected.
    Therefore, the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed changes do not involve physical alterations of the 
plant; no new or different type of equipment will be installed; nor 
are there significant changes in the methods governing normal plant 
operation. The changes simply redefine the parameters for evaluation 
of leakage in the drywell. The evaluation criteria for drywell 
leakage have been refocused into the areas that are most susceptible 
to IGSCC. Additionally, the changes do not create any new failure 
mechanisms, malfunctions, or accident initiators not already 
considered in the design and licensing bases.
    Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously analyzed.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The proposed amendment redefines the parameters for evaluation 
of leakage in the drywell. There are no physical alterations of the 
plant; no new or different type of equipment will be installed; nor 
are there significant changes in the methods governing normal plant 
operation. Additionally, the proposed changes do not exceed or alter 
a design basis or safety limit as established in the Monticello 
licensing basis.
    Therefore, these proposed changes will 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: 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. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of amendment request: February 11, 2003.
    Description of amendment request: The proposed amendments would 
revise technical specification (TS) 5.5.9, ``Ventilation Filter Testing 
Program (VFTP)'' by (1) incorporating filter test face velocity limits 
for the control room special ventilation system, auxiliary building 
special ventilation system, spent fuel pool special and inservice purge 
ventilation system, and shield building ventilation system; and (2) 
making editorial changes. The proposed amendments would also delete the 
additional conditions in Appendix B of the Operating Licenses which 
require the licensee to complete an evaluation of the maximum test face 
velocity for the ventilation systems in TS 5.5.9. The additional 
conditions also require the licensee to submit a license amendment 
request for a TS amendment to specify the maximum test face velocity if 
the maximum actual face velocity is the greater than 110 percent of 40 
fpm. Additionally, the proposed amendments would revise the penetration 
and system bypass limit from 0.05 percent to 0.5 percent for the 
ventilation systems.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Revision of the Allowable Filtration Penetration and System Bypass

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This license amendment request proposes to increase the 
penetration and system bypass limit for the control room special 
ventilation system, auxiliary building special ventilation system, 
spent fuel pool special and inservice purge ventilation system and 
shield building ventilation system from 0.05% to 0.5%. These 
ventilation systems are included in the plant design to mitigate 
accident consequences and are not assumed accident initiators, thus, 
this change does not involve a significant increase in the 
probability of an accident. This change will assure that the subject 
ventilation systems will perform within their intended design ranges 
thus, this change assures that the consequences of an accident are 
not increased.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    This proposed change does not alter the design, function, or 
operation of any plant component and does not install any new or 
different equipment. The malfunction of safety related equipment, 
assumed to be operable in the accident analyses, would not be caused 
as a result of the proposed Technical Specification change. No new

[[Page 18280]]

failure mode has been created and no new equipment performance 
burdens are imposed. Therefore the possibility of a new or different 
kind of accident from those previously analyzed has not been 
created.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    This license amendment request proposes to increase the 
penetration and system bypass limit for the control room special 
ventilation system, auxiliary building special ventilation system, 
spent fuel pool special and inservice purge ventilation system and 
shield building ventilation system from 0.05% to 0.5%. Site dose 
analyses are required to demonstrate that regulatory dose limits are 
met using Technical Specification allowed penetration and system 
bypass with an appropriate safety factor as an input to the 
evaluation. Since the dose analyses have not been modified to credit 
0.05% penetration and system bypass, this proposed change has no 
effect on the dose analyses which demonstrate that the regulatory 
limits are satisfied. Since the NRC regulatory limits must continue 
to be met and the safety factor will not be changed by this proposed 
Technical Specification change, this change does not involve a 
significant reduction in the margin of safety.

Addition of Filter Test Face Velocities

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This license amendment request proposes to add filter test face 
velocity minimum values for the control room special ventilation 
system, auxiliary building special ventilation system, spent fuel 
pool special and inservice purge ventilation system and shield 
building ventilation system. These ventilation systems are included 
in the plant design to mitigate accident consequences and are not 
assumed accident initiators, thus, this change does not involve a 
significant increase in the probability of an accident. This change 
will assure that the subject ventilation systems will perform within 
their intended design ranges thus, this change assures that the 
consequences of an accident are not increased.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    This proposed change does not alter the design, function, or 
operation of any plant component and does not install any new or 
different equipment. The malfunction of safety related equipment, 
assumed to be operable in the accident analyses, would not be caused 
as a result of the proposed Technical Specification change. No new 
failure mode has been created and no new equipment performance 
burdens are imposed. Therefore the possibility of a new or different 
kind of accident from those previously analyzed has not been 
created.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    This license amendment request proposes to add filter test face 
velocity minimum values for the control room special ventilation 
system, auxiliary building special ventilation system, spent fuel 
pool special and inservice purge ventilation system and shield 
building ventilation system. These additional Technical 
Specification limits on system performance assures these ventilation 
systems are tested and maintained within their designed function 
limits and may increase the margin of safety for these systems. 
Therefore this change does not involve a significant reduction in 
the margin of safety.

Editorial and Administrative Changes

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This license amendment request proposes editorial changes to 
Technical Specification Section 5.5.9, including replacement of 
ventilation system names with abbreviations and miscellaneous 
changes associated with addition of a new paragraph to this section, 
and proposes an administrative change to delete the Operating 
License Additional Condition for each unit that relates to NRC 
Generic Letter 99-02. Since these changes are editorial or 
administrative, they do not change any plant operating limits or 
technical requirements. Therefore these changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    This proposed change does not alter the design, function, or 
operation of any plant component and does not install any new or 
different equipment. The malfunction of safety related equipment, 
assumed to be operable in the accident analyses, would not be caused 
as a result of the proposed technical specification change. No new 
failure mode has been created and no new equipment performance 
burdens are imposed. Therefore, the possibility of a new or 
different kind of accident from those previously analyzed has not 
been created.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    This license amendment request proposes editorial changes to 
Technical Specification Section 5.5.9, including replacement of 
ventilation system names with abbreviations and miscellaneous 
changes associated with addition of a new paragraph to this section, 
and proposes an administrative change to delete the Operating 
License Additional Condition for each unit that relates to NRC 
Generic Letter 99-02. Since these changes are editorial or 
administrative, they do not change any plant operating limits or 
technical requirements. Therefore these 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 requests involve no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: March 11, 2003.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.1.4, ``Rod Group Alignment 
Limits,'' and TS 3.1.7, ``Rod Position Indication,'' to allow up to 1 
hour of soak time following substantial rod movement during which 
individual rod position indicators may not be within its limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This license amendment request proposes to allow up to one hour 
of soak time following substantial rod movement during which time 
the rod position indication may be outside its limits. This would 
allow an additional hour for rod position indication to be 
inoperable or a control rod to be misaligned prior to entry into a 
Technical Specification LCO [Limiting Condition for Operation] 
Condition and Required Actions.
    Rod position indication instrumentation is not an assumed 
accident initiator and thus this change does not involve a 
significant increase in the probability of an accident. Rod position 
indication instrumentation provides information on control rod 
position. Inoperable rod position indication instrumentation for an 
additional hour does not make a rod misaligned. The consequences of 
a rod misaligned for an additional hour are considered separately, 
thus inoperable rod position indication instrumentation, by itself, 
for an additional hour does not involve an increase in the 
consequences of an accident.
    This license amendment request may allow a misaligned rod to be 
undetected for an additional hour. Plant safety analyses consider 
two types of rod misalignment events, static misalignment and a 
dropped rod. This license amendment request does not involve a 
significant increase in the probability of a misaligned control rod 
event because the one-hour time extension does not affect the 
control rod drive system features, whose failure would result in 
either type of misalignment. This proposed one-hour time extension 
for a control rod to be misaligned does not involve a significant 
increase in the

[[Page 18281]]

consequences of a rod misalignment event as follows. The analyses 
show that a single dropped rod event, without any operator 
intervention, does not result in any fuel pin failure, therefore the 
rod drop event is not time dependent and an additional hour with the 
misalignment undetected and unmitigated does not increase the 
consequences of the event. Multiple rod drop events cause the 
reactor to trip and therefore an additional hour would not have any 
impact on this event.
    In the static misalignment event, one or more control rods are 
assumed to be statically misplaced from the allowed position. This 
situation might occur if a rod were left behind when inserting or 
withdrawing banks, or if a single rod were to be withdrawn. The 
analysis of this event is bounded by modeling the most limiting 
configuration which is the control banks at the full power insertion 
limit except for a single control rod fully withdrawn. The analyses 
show that, without any operator intervention, a single fully 
withdrawn rod event does not result in any fuel pin failure, 
therefore the static rod misalignment event is not time dependent 
and an additional hour with the misalignment undetected and 
unmitigated does not increase the consequences of the event. 
Multiple rod misalignment events are bounded by the single rod 
misalignment analyses and therefore an additional hour would not 
have any impact on this event.
    Therefore this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    This proposed change does not alter the design, function, or 
operation of any plant component and does not install any new or 
different equipment. The malfunction of safety related equipment, 
assumed to be operable in the accident analyses, would not be caused 
as a result of the proposed technical specification change. No new 
failure mode has been created and no new equipment performance 
burdens are imposed. Therefore the possibility of a new or different 
kind of accident from those previously analyzed has not been 
created.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    This license amendment request proposes to allow up to one hour 
of soak time following substantial rod movement during which time 
the rod position indication may be outside its limits. This would 
allow an additional hour for rod position indication instrumentation 
to be inoperable or a control rod to be misaligned prior to entry 
into a Technical Specification LCO Condition and Required Actions.
    The rod position indication system is an instrumentation system 
that provides indication to the operators that a control rod may be 
misaligned. Inoperable individual rod position indication 
instrumentation does not by itself in any way harm or impact reactor 
operation. Inoperable rod position indication instrumentation may 
impair the ability of the operators to detect a misaligned rod. The 
impact of inoperable rod position indication instrumentation may be 
offset by availability of other indications that a rod is misaligned 
such as nuclear instrumentation indication that reactor power has 
shifted to one side of the core or thermocouple indication that the 
core temperatures increased in one region of the core and/or 
decreased in another region of the core.
    The Prairie Island staff is not aware of a misaligned control 
rod in more than 50 reactor-years of plant operation. The likelihood 
of a misaligned rod at Prairie Island is small and the likelihood of 
a misaligned rod coincident with inoperable rod position indication 
during the allowed one-hour extension is smaller.
    The addition of one hour soak time for the rod position 
indication instrumentation will allow the operators and engineers to 
focus on monitoring the reactor performance without unnecessary 
entry into LCO Conditions and Required Actions with the concomitant 
administrative activities. Thus, these changes may enhance plant 
safety and reliability of equipment.
    In conclusion, the proposed addition of an LCO Note in LCO 3.1.4 
and 3.1.7 does not involve a significant reduction in the margin of 
safety because rod position indication instrumentation inoperability 
by itself does not impact plant safety, rod misalignment is 
unlikely, there may be other indications of rod misalignment, rod 
misalignment coincident with rod position indication instrumentation 
inoperability within the one hour extension is unlikely, and plant 
safety may be enhanced by avoiding unnecessary LCO Condition entry.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: March 19, 2003.
    Description of amendment request: The proposed amendments would 
revise the Technical Specification (TS) 5.3, ``Plant Staff 
Qualifications.'' The proposed amendments would revise requirements 
that have been superseded based on licensed operator training programs 
being accredited by the National Academy for Nuclear Training (NANT) 
and promulgation of the revised 10 CFR part 55, ``Operators' 
Licenses,'' which became effective on May 26, 1987.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed Technical Specification change is an administrative 
change to clarify the current requirements for licensed operator 
qualifications and the licensed operator training program. With this 
change, the Technical Specifications continue to meet the current 
requirements of 10 CFR [Part] 55.
    Although licensed operator qualifications and training may have 
an indirect impact on accidents previously evaluated, the NRC 
considered this impact during the rulemaking process, and by 
promulgation of the revised 10 CFR [Part] 55 rule, concluded that 
this impact remains acceptable as long as the licensed operator 
training programs are certified to be accredited and are based on a 
systems approach to training. The Prairie Island Nuclear Generating 
Plant licensed operator training program is accredited by the 
National Academy for Nuclear Training and is based on a systems 
approach to training. The proposed Technical Specification change 
takes credit for the National Academy for Nuclear Training 
accreditation of the licensed operator training program. The 
Technical Specification requirements for all other plant staff 
qualifications remain unchanged.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed Technical Specification change is an administrative 
change to clarify the current requirements for licensed operator 
qualifications and the licensed operator training program and to 
conform to the revised 10 CFR [Part] 55.
    As discussed above, although licensed operator qualifications 
and training may have an indirect impact on the possibility of a new 
or different kind of accident from any accident previously 
evaluated, the NRC considered this impact during the rulemaking 
process, and by promulgation of the revised rule, concluded that 
this impact remains acceptable as long as licensed operator training 
programs are certified to be accredited and based on a systems 
approach to training. As previously noted, the Prairie Island 
Nuclear Generating Plant licensed operator training program is 
accredited by the National Academy for Nuclear Training and is based 
on a systems approach to training. The proposed Technical

[[Page 18282]]

Specification change takes credit for the National Academy for 
Nuclear Training accreditation of the licensed operator training 
program. The Technical Specification requirements for all other 
plant staff qualifications remain unchanged.
    Additionally, the proposed Technical Specification change does 
not affect plant design, hardware, system operation, or procedures. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed Technical Specification change is an administrative 
change to clarify the current requirements applicable to licensed 
operator qualifications and the licensed operator training program. 
With this change the Technical Specifications continue to be 
consistent with the requirements of 10 CFR [Part] 55. The Technical 
Specification qualification requirements for all other plant staff 
remain unchanged.
    Licensed operator qualifications and training can have an 
indirect impact on a margin of safety. However, the NRC considered 
this impact during the rulemaking process, and by promulgation of 
the revised 10 CFR [Part] 55, determined that this impact remains 
acceptable when licensees maintain a licensed operator training 
program that is accredited and based on a systems approach to 
training. As noted previously, the Prairie Island Nuclear Generating 
Plant licensed operator training program is accredited by the 
National Academy for Nuclear Training and is based on a systems 
approach to training.
    The NRC has concluded, as stated in NUREG-1262, ``Answers to 
Questions at Public Meetings Regarding Implementation of Title 10, 
Code of Federal Regulations, Part 55 on Operators' Licenses,'' that 
the standards and guidelines applied by the Institute for Nuclear 
Power Operations' National Academy for Nuclear Training in their 
training accreditation program are equivalent to those put forth or 
endorsed by the NRC. As a result, maintaining a National Academy for 
Nuclear Training accredited, systems approach based licensed 
operator training program is equivalent to maintaining an NRC 
approved licensed operator training program which conforms with 
applicable NRC Regulatory Guides or NRC endorsed industry standards. 
The margin of safety is maintained by virtue of maintaining the 
National Academy for Nuclear Training accredited licensed operator 
training program.
    In addition, the NRC published NRC Regulatory Issue Summary 
2001-01, ``Eligibility of Operator License Applicants,'' dated 
January 18, 2001, ``to familiarize addressees with the NRC's current 
guidelines for the qualification and training of reactor operator 
(RO) and senior operator (SO) license applicants.'' This document 
again acknowledges that the Institute for Nuclear Power Operations' 
National Academy for Nuclear Training guidelines for education and 
experience, outline acceptable methods for implementing the NRC's 
regulations in this area.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment requests: February 6, 2003.
    Description of amendment requests: The proposed license amendments 
would revise Surveillance Requirements (SRs) 3.3.1.2 and 3.3.1.3 of TS 
3.3.1, ``Reactor Trip System Instrumentation,'' of the Diablo Canyon 
Technical Specifications. The change to SR 3.3.1.2 is consistent with 
NRC-approved Industry/Technical Specifications Task Force (TSTF) 
Standard Technical Specification Change Traveler, TSTF-371. The change 
to SR 3.3.1.3 is editorial in nature.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change to Technical Specifications (TS) 
Surveillance Requirement (SR) 3.3.1.2 and SR 3.3.1.3 is consistent 
with the NRC approved Industry/Technical Specifications Task Force 
Standard Technical Specification Change Traveler, TSTF-371, and 
NUREG-1431, ``Standard Technical Specifications, Westinghouse 
Plants,'' Revision 2.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since there are 
no hardware changes. The reactor trip system (RTS) instrumentation 
will be unaffected. Protection systems will continue to function in 
a manner consistent with the plant design basis. All design, 
material, and construction standards that were applicable prior to 
the request are maintained.
    The probability and consequences of accidents previously 
evaluated in the Updated Final Safety Analysis Report (UFSAR) are 
not adversely affected because the change to the nuclear 
instrumentation system (NIS) power range channel daily surveillance 
assures the conservative response of the channel even at part-power 
levels.
    The proposed change modifies the NIS power range channel daily 
surveillance requirement to help assure the NIS power range 
functions are tested in a manner consistent with the safety analysis 
and licensing basis.
    The proposed change will not affect the probability of any event 
initiators. There will be no degradation in the performance of, or 
an increase in the number of challenges imposed on, safety-related 
equipment assumed to function during an accident situation. There 
will be no change to normal plant operating parameters or accident 
mitigation performance.
    The proposed change will not alter any assumptions or change any 
mitigation actions in the radiological consequence evaluations in 
the USAR.
    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.
    There is no hardware change or change in the method by which any 
safety-related plant system performs its safety function. This 
change will not affect the normal method of plant operation or 
change any operating parameters. No performance requirements or 
response time limits will be affected. The NIS power range high trip 
setpoint adjustment requirements, prior to adjusting indicated power 
in a decreasing power direction, will ensure the reactor power level 
is consistent with assumptions made in the safety analysis and 
licensing basis.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed change. There will be no adverse effect or 
challenges imposed on any safety-related system as a result of the 
change.
    This amendment does not alter the design or performance of the 
Eagle 21 System, NIS, or Solid State Protection System used in the 
plant protection systems.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change requires a revision to the criteria for 
implementation of NIS power range channel adjustments based on 
secondary power calorimetric calculations; however, the change does 
not eliminate any RTS surveillances or alter the frequency of 
surveillances required by the Technical Specifications. The revision 
to the criteria for implementation of the daily surveillance will 
have a conservative effect on the performance of the NIS power range 
channels, particularly at part-power conditions. The nominal trip 
setpoints specified in the Technical Specification Bases and the 
safety analysis

[[Page 18283]]

limits assumed in the transient and accident analyses are unchanged. 
None of the acceptance criteria for any accident analysis is 
changed.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. There will be no impact on the overpower 
limit, departure from nucleate boiling ratio limits, heat flux hot 
channel factor (FQ), nuclear enthalpy rise hot channel factor (FDH), 
loss of coolant accident peak cladding temperature, peak local power 
density, or any other margin of safety. The radiological dose 
consequence acceptance criteria listed in the Standard Review Plan 
will continue to be met.
    The imposition of appropriate surveillance testing requirements 
will not reduce any margin of safety since the change will assure 
that safety analysis assumptions on reactor power are verified on a 
periodic frequency.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment requests: February 28, 2003.
    Description of amendment requests: The proposed license amendments 
would revise Technical Specification (TS) 3.3.1, ``Reactor Trip System 
(RTS) Instrumentation,'' to add Surveillance Requirement (SR) 3.3.1.16 
to function 3.a, ``Power Range Neutron Flux Rate-High Positive Rate 
Trip,'' in Table 3.3.1-1. The amendments would also eliminate periodic 
pressure sensor response time testing (RTT) and periodic protection 
channel RTT.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since there are 
no hardware changes.
    The design of the Reactor Trip System (RTS) instrumentation, 
specifically the positive flux rate trip (PFRT) function, will be 
unaffected. The reactor protection system will continue to function 
in a manner consistent with the plant design basis. All design, 
material, and construction standards that were applicable prior to 
the request are maintained.
    The proposed change imposes additional surveillance requirements 
to assure safety-related structures, systems, and components are 
verified to be consistent with the safety analysis and licensing 
basis. In this specific case, a response time verification 
requirement will be added to the PFRT function.
    The Technical Specification Bases changes do not result in a 
condition where the design, material, or construction standards that 
were applicable prior to change are altered. The same RTS and 
engineered safety features actuation system instrumentation is being 
used; the time response allocations/modeling assumptions in the 
Updated Final Safety Analysis Report (UFSAR) Chapter 15 analyses are 
still the same; only the method of verifying time response is 
changed. The proposed change will not change any system interface 
and could not increase the likelihood of an accident since these 
events are independent of this change.
    The proposed change will not affect the probability of any event 
initiators. There will be no degradation in the performance of, or 
an increase in the number of challenges imposed on safety-related 
equipment assumed to function during an accident situation. There 
will be no change to normal plant operating parameters or accident 
mitigation performance.
    The proposed activity will not change, degrade or prevent 
actions or alter any assumptions previously made in evaluating the 
radiological consequences of an accident described in the UFSAR.
    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.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. This change will not affect the normal method of plant 
operation or change any operating parameters. No performance 
requirements will be affected; however, the proposed change does 
impose additional surveillance requirements for the PFRT function. 
These additional requirements are consistent with assumptions made 
in the safety analysis and licensing basis.
    This change does not alter the performance of the process 
protection racks, nuclear instrumentation, and logic systems used in 
the plant protection systems. These systems will still have their 
response time verified by test before being placed in operational 
service. Changing the method of verifying instrument response for 
these systems (assuring equipment operability) from time response 
testing to channel and calibration checks will not create any new 
[accident] initiators or scenarios. Periodic surveillance of these 
systems will continue and may be used to detect degradation that 
could cause the response time characteristic to exceed the total 
allowance. The total response time allowance for each function 
bounds all degradation that cannot be detected by periodic 
surveillance.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this change. There will be no adverse effects or challenges 
imposed on any safety-related system as a result of this change.
    Therefore the proposed change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. There will be no impact on the overpower 
limit, departure from nucleate boiling ratio limits, heat flux hot 
channel factor, nuclear enthalpy rise hot channel factor, loss of 
coolant accident peak cladding temperature, peak local power 
density, or any other margin of safety. The radiological dose 
consequence acceptance criteria listed in the Standard Review Plan 
will continue to be met.
    The safety analysis limits assumed in the transient and accident 
analyses are unchanged. None of the acceptance criteria for any 
accident analysis are changed. The imposition of additional 
surveillance requirements maintains the margin of safety by assuring 
that the affected safety analysis assumptions on equipment response 
time are verified on a periodic frequency.
    This change does not affect the total system response time 
assumed in the safety analysis. The periodic system response time 
verification method for the process protection racks, nuclear 
instrumentation, and logic systems are modified to allow use of 
engineering data. The method of verification still provides 
assurance that the total system response is within that defined in 
the safety analysis, since calibration tests will continue to be 
performed and may be used to detect any degradation which might 
cause the response time to exceed the total allowance. The total 
response time allowance for each function bounds all degradation 
that cannot be detected by periodic surveillance. Based on the 
above, it is concluded that the proposed change does not result in a 
significant reduction in margin with respect to plant safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) 
are satisfied. Therefore, the NRC staff proposes to determine that 
the amendment requests involve no significant hazards consideration.

[[Page 18284]]

    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas 
and Electric Company, P.O. Box 7442, San Francisco, California 
94120.
    NRC Section Chief: Stephen Dembek.

Pacific Gas and Electric Company, Docket No. 50-323, Diablo Canyon 
Nuclear Power Plant, Unit No. 2, San Luis Obispo County, California

    Date of amendment request: March 3, 2003.
    Description of amendment request: The proposed license amendment 
would revise Technical Specification (TS) 5.5.9, ``Steam Generator 
Tube Surveillance Program,'' and TS 5.6.10, ``Steam Generator Tube 
Inspection Report,'' for Diablo Canyon Power Plant (DCPP) Unit 2, to 
apply a probability of detection (POD) of 1.0 to the bobbin 
indication in the steam generator (SG) 4 tube at row 44, column 45 
at the second tube support plate (TSP) on the hot leg side (R44C45-
2H) for the beginning of cycle (BOC) voltage distribution for the 
DCPP Unit 2 BOC Cycle 12 operational assessment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The use of probability of detection (POD) of 1.0 for the bobbin 
indication in the Diablo Canyon Power Plant (DCPP) Unit 2 steam 
generator (SG) 4 tube at row 44, column 45 at the second tube 
support plate (TSP) on the hot leg side (R44C45-2H) for the 
beginning of cycle (BOC) voltage distribution for the DCPP Unit 2 
BOC cycle 12 operational assessment does not increase the 
probability of an accident. Based on industry and plant specific 
bobbin detection data for outside diameter stress corrosion cracks 
(ODSCC) within the SG tube support plate region, large voltage 
bobbin indications, such as those the size of indication R44C45-2H, 
can be detected with 100 percent certainty. Since large voltage 
ODSCC bobbin indications within the SG TSP can be detected, they 
will not be left in service, and therefore these indications should 
not be included in the voltage distribution for the purpose of 
operational assessments. Therefore, these large voltage indications 
will not result in an increase in the probability of a steam 
generator tube rupture (SGTR) accident or an increase in the 
consequences of a SGTR or main steam line break (MSLB) accident.
    Therefore, the proposed changes will not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The use of a POD of 1.0 for the DCPP Unit 2 R44C45-2H bobbin 
indication for the BOC voltage distribution for the DCPP Unit 2 BOC 
cycle 12 operational assessment concerns the SG tubes and can only 
affect the SGTR accident. Since the SGTR accident is already 
considered in the Final Safety Analysis Report Update, there in [is] 
no possibility to create a design basis accident which has not been 
previously evaluated.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The use of POD of 1.0 for the DCPP Unit 2 R44C45-2H bobbin 
indication for the BOC voltage distribution for the DCPP Unit 2 BOC 
cycle 12 operational assessment does not involve a significant 
reduction in a margin of safety. The applicable margin of safety 
potentially impacted is the Technical Specification 5.6.10, ``Steam 
Generator Tube Inspection Report,'' projected end-of-cycle leakage 
for a MSLB accident and the projected end-of-cycle probability of 
burst. Based on industry and plant specific bobbin detection data 
for ODSCC within the SG tube support plate region, large voltage 
bobbin indications, such as those the size of indication R44C45-2H, 
can be detected with 100 percent certainty and will not be left in 
service. Therefore these indications should not be included in the 
voltage distribution for the purpose of operational assessments. 
Therefore, these large voltage indications will not result in a 
significant increase in the actual end-of-cycle leakage for a MSLB 
accident or the actual end-of-cycle probability of burst.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas 
and Electric Company, P.O. Box 7442, San Francisco, California 
94120.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: February 14, 2003.
    Description of amendment request: The proposed amendment would 
extend the surveillance test intervals and allowed out-of-service 
times for the end-of-cycle recirculation pump trip instrumentation.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR), Section 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would extend the allowed out-of-service 
times (AOTs) and surveillance test intervals (STIs) for the end of 
cycle recirculation pump trip (EOC-RPT) instrumentation system. No 
changes are being made to any EOC-RPT instrumentation setpoints or 
components. The effect of the proposed changes is to reduce the 
potential for unnecessary plant scrams or transients. The proposed 
changes were evaluated in General Electric Company Topical Report 
GENE-770-06-1-A which concluded that they do not result in a 
degradation in overall plant safety.
    Since the proposed changes do not affect any accident initiator, 
and since the EOC-RPT instrumentation will remain capable of 
performing its design function, 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.
    Extending the AOTs and STIs for the EOC-RPT instrumentation does 
not change the design function or operation of any plant equipment. 
Additionally, no new modes of plant operation are involved with 
these changes.
    Therefore, the proposed changes do 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.
    No changes are being made to any plant instrumentation setpoints 
or to the required level of redundancy. The proposed changes were 
evaluated in General Electric Company Topical Report GENE-770-06-1-
A, which concluded that they do not result in a degradation in 
overall plant safety.
    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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: January 29, 2003.
    Description of amendment request: The licensee proposed 
administrative and editorial changes to the Salem Nuclear Generating 
Station (Salem), Unit No. 1 and Unit No. 2 Technical Specifications 
(TSs) as follows: (1) The second equation in Salem Unit No. 2 TS

[[Page 18285]]

Limiting Condition for Operation 3.2.2 on page 3/4 2-5 will be revised; 
(2) Salem Unit No. 2 TS Table 3.3-6 will be revised to indicate that 
one operable channel of containment air particulate activity reactor 
coolant system (RCS) leakage detection instrumentation is required for 
operation in Modes 1 through 4; (3) Salem Unit No. 1 TS 3/4.7.6 Action 
Statements ``d.'' (for Modes 1, 2, 3 and 4) and ``e.'' (for Modes 5 and 
6) will be revised to refer to Action 25 in TS Table 3.3-6; and (4) 
Salem Unit No. 2 TS 3/4.7.6 Action Statements ``d.'' (for Modes 1, 2, 3 
and 4) and ``e.'' (for Modes 5 and 6) will be revised to refer to 
Action 28 in TS Table 3.3-6.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the TSs are administrative or editorial 
in nature and do not change the intent of any Technical 
Specification requirement. No changes are being made to any plant 
systems, structures or components (SSCs).
    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 administrative and editorial changes to the TSs do 
not change the design function or operation of any plant equipment. 
Additionally, no new modes of plant operation are involved with 
these changes.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes are administrative and editorial 
corrections to the TSs that do not affect the ability of plant SSCs 
to perform their design basis accident functions. 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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: March 11, 2003.
    Description of amendment request: The amendment application 
requests a revision to the Unit 1 defueled Technical Specifications 
administrative controls section to propose changes in organizational 
responsibilities. Specifically, the proposed change identifies that the 
Vice President, Engineering & Technical Services would be responsible 
for decommissioning activities. Additionally, the Station Manager would 
be designated as having approval authority for activities within the 
Station Manager's organization.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. This is a request to revise the San Onofre Nuclear 
Generating Station, Unit 1 permanently defueled technical 
specifications administrative controls. The proposed administrative 
changes are due to a realignment of the Unit 1 Decommissioning 
Project into the Engineering & Technical Services organization and 
the establishment of the Station Manager position within the Nuclear 
Generation organization. The proposed changes identify the Vice 
President, Engineering & Technical Services to be responsible for 
decommissioning activities and provides the Station Manager the 
opportunity to approve procedures and changes to procedures and 
changes to the Process Control Program that are under the Station 
Manger's responsibility. Therefore, the proposed changes do not 
involve an increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different type of accident 
from any accident previously evaluated?
    No. The proposed changes are administrative. Therefore, the 
proposed changes do not involve the possibility of a new or 
different type of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety?
    No. The proposed changes are administrative. Therefore, the 
proposed changes do not involve a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis. These 
administrative changes do not affect the design or operation of the 
facility 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: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Acting Section Chief: Mark Thaggard.

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

    Date of amendments request: March 25, 2003.
    Description of amendments request: The proposed amendments would 
revise Technical Specification 3.5.2, ``ECCS--Operating,'' Surveillance 
Requirement (SR) 3.5.2.5. Specifically, the proposed change would 
replace the requirement to verify specific surveillance test values for 
the Emergency Core Cooling System (ECCS) pumps with the requirement to 
verify the developed head for each ECCS pump in accordance with the 
Inservice Testing Program. This new requirement is identical to SR 
3.5.2.4 in NUREG-1432, ``Standard Technical Specifications, Combustion 
Engineering Plants,'' Revision 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Deleting the specific surveillance test values for Emergency 
Core Cooling System (ECCS) pumps from Surveillance Requirement (SR) 
3.5.2.5 does not affect the probability of occurrence or 
consequences of an accident previously evaluated because ECCS pumps 
are for accident mitigation and do not contribute to initiation of 
accidents. Periodic surveillance testing of the ECCS pumps in 
accordance with the Inservice Testing (IST) program provides 
assurance that the pumps will perform as assumed in the safety 
analysis. There is no change to the safety analysis.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 18286]]

    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    ECCS pumps are for accident mitigation and do not contribute to 
accident initiation. The ECCS system will still be verified capable 
of meeting its emergency core cooling and IST requirements. There is 
no change to the safety analysis.
    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.
    There is no change to the safety analysis. Testing of the ECCS 
pumps as required by the IST Program combined with the existing 
Technical Specification 3.5.2--``ECCS--Operating'' surveillance 
requirements ensure that the ECCS requirements remain met without a 
significant reduction in a margin of safety. Therefore, there is no 
significant reduction in a margin of safety.
    Based on the above, SCE [Southern California Edison Company] 
concludes that the proposed amendments present no significant 
hazards consideration under the standards set forth in 10 CFR 
50.92(c), and, accordingly, a finding of ``no significant hazards 
consideration'' is justified.

    The NRC 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 involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: March 13, 2003.
    Description of amendment request: The proposed amendments would 
modify the Sequoyah Nuclear Plant, Units 1 and 2, Operating Licenses 
DPR-77 and DPR-79. This proposed request provides Technical 
Specification (TS) change 03-01 that would revise the limiting 
condition for operation for TS Section 3.5.1, ``Cold Leg Injection 
Accumulators'' and TS Section 3.5.5, ``Refueling Water Storage Tank.'' 
This revision would modify the single boron concentration requirement 
by inserting a table that defines the minimum and maximum amount of 
boron that is required for accident mitigation based on the number of 
tritium producing rods in the core.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change modifies the required boron concentration 
for the cold leg accumulators (CLAs) and refueling water storage 
tank (RWST). The proposed values have been verified to maintain the 
required accident mitigation safety function for the CLAs and RWST. 
The CLAs and RWST safety function is to mitigate accidents that 
require the injection of borated water to cool the core and to 
control reactivity. These functions are not potential sources for 
accident generation and the modification of the boron concentration 
that supports event mitigation will not increase the potential for 
an accident. Therefore, the possibility of an accident is not 
increased by the proposed changes. The boron levels for this change 
are based on the number or tritium producing rods in the core. As 
the number of rods is increased the need for additional shutdown 
boron also increases. This effect has been evaluated with the same 
methodology utilized for previous NRC approved amendments associated 
with tritium production. This methodology ensures that the impact of 
tritium producing rods is adequately compensated for by the required 
boron concentrations and has been incorporated into the proposed 
revision. Since the boron levels will continue to maintain the 
safety function of the CLAs and RWST in the same manner as currently 
approved, the consequences of an accident is not increased by the 
proposed changes.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change only modifies boron concentrations for 
accident mitigation functions of the CLAs and RWST. These functions 
do not have a potential to generate accidents as they only serve to 
perform mitigation functions associated with an accident. The 
proposed requirements will maintain the mitigation function in an 
identical manner as currently approved. There are no plant equipment 
or operational changes associated with the proposed revision other 
than the adjustment of the boron level in the CLAs and RWST. 
Therefore, since the CLA and RWST functions are not altered and the 
plant will continue to operate without change, the possibility of a 
new or different kind of an accident is not created.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    This change proposes boron concentration requirements that 
support the accident mitigation functions of the CLAs and RWST 
equivalent to the currently approved limits. The proposed change 
does not alter any plant equipment or components and does not alter 
any setpoints utilized for the actuation of accident mitigation 
system or control functions. The proposed boron values have been 
verified to provide the same level of reactivity control for 
accident mitigation. Therefore, the proposed change will not involve 
a significant reduction in a margin of safety.

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

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

    Date of amendment request: March 12, 2003.
    Description of amendment request: The proposed amendment would 
revise the Updated Final Safety Analysis Report (UFSAR) and the 
Technical Specification Bases. The revision would update the quality 
assurance criteria and the basis for the seismic qualification of the 
ducting installed as part of the suspended ceiling air delivery system 
in the main control room (MCR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The design function of the MCR ducting system is to support 
pressurization and cooling of the control room during normal and 
accident conditions. The MCR ducting is a passive plant feature and 
does not act as an accident initiator. Consequently, the changes in 
the MCR ducting system and suspended ceiling quality assurance (QA) 
requirements and qualification methodology do not result in an 
increase in the probability of an accident previously evaluated.
    For the principal design basis accidents, Loss of Coolant 
Accident (LOCA), Internal Flood, Steam Generator Tube Rupture 
(STGR), Main Steam Line Break (MSLB), etc., the integrity of the MCR 
HVAC [heating, ventilation, and air conditioning] system, including 
the suspended ceiling, will not be compromised. These accidents do 
not have a structural effect on the MCR. This means that for 
postulated radiological or toxic chemical accidents, the ability to 
both pressurize and

[[Page 18287]]

maintain MCR temperatures within the design limits is unaffected by 
the limited QA and newly defined seismic requirements for the air 
delivery components.
    An accident that involves a fire that affects the MCR or the 
habitability of the MCR was not a consideration for the 
qualification of the air distribution components. A fire of this 
nature will result in plant operation from the Auxiliary Control 
Room which is supported by a separate heating, ventilation and air 
conditioning (HVAC) supply system.
    An earthquake (including the Design Basis SSE [safe shutdown 
earthquake]) is the only event for which the design basis for the 
MCR HVAC and suspended ceiling is potentially challenged. A seismic 
qualification report by an industry seismic expert concludes that 
the air delivery components will remain in place, will retain their 
structural integrity such that flow will not be impeded, and the 
pressure boundary will not be lost during and subsequent to a design 
basis seismic event. Further, as assured by TVA's qualification 
report, the suspended ceiling will remain in place during and 
subsequent to a seismic event or accident. Thus, the revised QA and 
seismic qualification requirements for the MCR air delivery 
components and suspended ceiling will not result in loss of safety 
function for any design basis accident or event. Consequently, the 
accident dose consequences as previously evaluated in the UFSAR are 
not affected by the proposed license amendment.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The MCR air delivery components addressed by the proposed 
amendment are not an accident initiator and therefore, failure of 
these components will not initiate a design basis accident. In 
addition, the subject air delivery components and suspended ceiling 
have been seismically qualified, as previously discussed, and a 
determination has been made that they will not fail during a design 
basis accident. Therefore, the air delivery components and suspended 
ceiling will continue to perform their safety function during normal 
and accident conditions. Consequently, this activity does not create 
a possibility of a new or different type of accident than any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The changes addressed in TVA's proposed amendment are 
associated with changes in QA requirements and seismic qualification 
methodology for safety related air delivery components and for the 
suspended ceiling. The change does not affect specific HVAC 
equipment safety limits, design limits, set points, or other 
critical parameters. In addition, the new seismic analysis 
methodology and limited QA requirements ensure that these components 
will continue to perform their safety functions during normal and 
accident conditions. The previously implied margin of safety against 
structural or functional failure of the air delivery components or 
suspended ceiling during and after a design basis SSE has not been 
reduced. Consequently, the MCR HVAC system or suspended ceiling 
margin of safety has not been significantly reduced by this proposed 
amendment.

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

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

    Date of amendment request: March 24, 2003.
    Description of amendment request: The proposed amendment would 
revise the design and licensing basis failure modes and effects 
analysis for specific valves in the essential raw cooling water system, 
component cooling water system, and control air system. Tennessee 
Valley Authority has identified a condition where containment 
integrity, accident flood levels, and sump boron concentrations 
subsequent to a high-energy line break events could not be assured 
automatically as stated in the updated final safety analysis report 
(UFSAR). In certain postulated events, manual actions may be required 
using equipment not currently evaluated in the UFSAR.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously 
evaluated[?]
    Response: No.
    The manual actions required by this change are only needed after 
a high energy line break (HELB) accident, such as a loss-of-coolant-
accident (LOCA), main steam line break (MSLB), feedwater line break 
accidents, etc., has occurred inside containment and a single 
failure of an outboard containment isolation valve to close has 
occurred on one of four specific lines inside containment. In this 
event, the manual actions ensure containment is isolated, which is 
consistent with the current design. Consequently, the manual actions 
of isolating the air and water lines after an accident do not affect 
the frequency of any accident previously evaluated in the Updated 
Final Safety Analysis Report (UFSAR).
    The UFSAR currently indicates that the containment vessel design 
and the containment isolation system automatically ensure 
containment integrity is maintained and thus ensure that release of 
radioactive material from containment remains below allowable limits 
during and subsequent to an accident. Current UFSAR Failure Modes 
and Effects Analysis (FMEA) for the affected essential raw cooling 
water (ERCW), component cooling system (CCS), and control air system 
(CAS) valves indicate a single failure of the outboard containment 
isolation valve in conjunction with a concurrent accident and 
consequential (due to interaction) failure of the system piping 
inside containment, has no adverse effect on the plant; thus, 
containment integrity is ensured automatically. This change revises 
these evaluations to indicate manual actions are required to ensure 
containment integrity in the event of an HELB and single failure of 
an outboard containment isolation valve. Evaluations have been 
performed to ensure adequate instrumentation and time is available 
to recognize the need and to manually isolate an affected line 
subsequent to an HELB if the outboard containment isolation valve 
does not close. The emergency procedures have been revised that 
require manual actions to be performed to isolate CAS, ERCW, and CCS 
and to open and close a post accident sampling facility (PASF) 
cooling water supply valve. The Operations Staff has confirmed that 
the subject containment lines can be isolated within the allowable 
time and without exceeding the dose limitations as required by 10 
CFR [Part] 50, Appendix A, General Design Criteria (GDC) 19, 
``Control Room.''
    Evaluations have indicated that adequate instrumentation, time, 
and staffing are available to manually isolate the lines into 
containment. Operator actions are achievable and can be accomplished 
without heroic actions. Therefore, containment integrity from 
overpressurization or flooding is maintained within the current 
design basis analysis, and the radiological consequences of an 
accident will not be increased by this change. Consequently, 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.
    This change implements manual actions to isolate four specific 
containment lines in lieu of automatic containment isolation for 
previously identified accidents. The manual actions are required to 
maintain containment integrity from overpressurization, containment 
flood levels, sump pH levels, and emergency core cooling system 
(ECCS) water boron concentrations subsequent to an HELB inside 
containment concurrent with a single failure of an outboard 
containment isolation valve on a CAS, ERCW, or CCS line. The UFSAR 
FMEA evaluations will be revised by this proposed change to include 
the failure modes and associated manual actions.
    NRC Information Notice (IN) 97-78, ``Crediting of Operator 
Actions in Place of Automatic Actions and Modifications of

[[Page 18288]]

Operator Actions, including Response Times,'' provided guidance to 
the industry concerning use of operator actions in place of 
automated system or component actuation. IN 97-78 states: In those 
instances where licensees consider temporary or permanent changes to 
the facility which credit operator actions, the NRC has relied on 
the guidance provided in * * * ANSI/ANS 58.8, ``Time Response Design 
Criteria for Safety-related Operator Actions,'' * * * for evaluating 
such changes. The American Nuclear Society (ANS)-58.8, establishes 
the requirements for safety-related operator actions, which are 
summarized as follows: (1) The specific operator actions required, 
(2) the potentially harsh or inhospitable environmental conditions 
expected, (3) ingress/egress paths taken by the operators to 
accomplish functions, (4) procedural guidance for required actions, 
(5) operator training and qualifications to carry out actions, (6) 
any additional support personnel and/or equipment to carry out 
actions, (7) information required by the control room staff to 
determine whether action is required, including qualified 
instrumentation to diagnose the situation and to verify that the 
action is successfully, (8) ability to recover from credible errors 
in performance of manual actions, and the expected time required to 
make such a recovery, and (9) consideration of risk significance of 
operator actions.
    The manual actions implemented by this change can be completed 
within the guidance and criteria provided in IN 97-78 and ANS-58.8. 
Consequently, the manual actions can be credited in the mitigation 
of the specific accidents. With credit for the manual actions to 
isolate the affected lines subsequent to an accident inside 
containment, the type of accidents and consequences currently 
evaluated in the UFSAR, remains the same. Therefore, the proposed 
change does not create the possibility of new or different kinds of 
accidents from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety[?]
    Response: No.
    This change establishes requirements for manual actions to 
isolate one air line and three water lines subsequent to an accident 
inside containment concurrent with a single failure of a containment 
isolation valve to close. The manual actions ensure air or water 
cannot continue to enter containment with a single failure of an 
outboard containment isolation valve when the line pressure boundary 
inside containment is lost due to an accident and associated pipe 
interactions. The safety-related configuration of the lines 
(outboard motor operated valve and inboard check valve) continues to 
ensure the containment environment is automatically prevented from 
exiting the line to outside the containment. Safety-related 
instrumentation is available to inform operators that the manual 
actions are required, and operators have been trained in the 
requirements for addressing the failures of valves to close. In 
addition, adequate time and resources are available to perform the 
manual actions. The manual actions meet the criteria for safety-
related operator actions contained in NRC IN 97-78 and ANS-58.8. 
Further, the proposed change to allow credit for the manual actions 
does not affect the offsite and Main Control Room dose consequences 
of the accidents currently reported in UFSAR Chapter 15, Accident 
Analyses. 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: March 6, 2003.
    Brief description of amendments: Technical Specifications Section 
1.1 ``Definitions'' for Engineered Safety Feature (ESF) Response Time 
and Reactor Trip System (RTS) Response Time require U.S. Nuclear 
Regulatory Commission (NRC) review and approval of any methodology used 
to allocate response times in lieu of measuring them. The application 
requests NRC review and approval of a topical report to allow the use 
of allocated signal processing and actuation logic response times in 
the overall verification of the protection system channel response 
time.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not result in a condition where the 
design, material, and construction standards that were applicable 
prior to the change are altered. The same RTS and ESFAS [Engineered 
Safety Feature Actuation System] instrumentation are being used and 
the time response allocations and modeling assumptions in the 
Chapter 15 safety analysis are unchanged. Only the method of 
verifying the time response is changed. The proposed change will not 
modify any system interface and could not increase the likelihood of 
an accident since these events are independent of this change. The 
proposed activity will not change, degrade, or prevent actions or 
alter any assumptions previously made in evaluating the radiological 
consequences of an accident described in the FSAR [Final Safety 
Analysis Report]. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not alter the performance of the 
process protection racks, the nuclear instrumentation, or the logic 
systems used in the plant protection systems. Periodic surveillance 
of these systems will continue and may be used to detect degradation 
that could cause the response time characteristics to exceed the 
total allowance. Changing the method of periodically verifying 
instrument response for these systems from response time testing to 
calibration and channel checks will not create any new accident 
initiators or scenarios. Periodic surveillance of these systems will 
continue and may be used to detect degradation that could cause the 
response time characteristic to exceed the total allowance. The 
total time response allowance for each function bounds all 
degradation that cannot be detected by periodic surveillance. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not affect the total system response 
time assumed in the safety analysis. The periodic response time 
verification method for the Process protection racks, the nuclear 
instrumentation and the logic systems is modified to allow the use 
of actual test data or engineering data. The method of verification 
still provides assurance that the total system response time is 
within that defined in the safety analysis, since calibration tests 
will continue to be performed and may be used to detect any 
degradation which might cause the response time to exceed the total 
allowance. The total response time allowance for each function 
bounds all degradation that cannot be detected by

[[Page 18289]]

periodic surveillance. Therefore the proposed change does not 
involve a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: March 18, 2003.
    Brief description of amendments: The proposed amendment would 
delete certain of the Surveillance Requirements in Technical 
Specification 3.6.3 entitled ``Containment Isolation Valves.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Do the proposed changes involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since there are 
no hardware changes. Protection systems will continue to function in 
a manner consistent with the plant design basis. All design, 
material, and construction standards that were applicable prior to 
the request are maintained.
    The probability and consequences of accidents previously 
evaluated in the FSAR [Final Safety Analysis Report] are not 
adversely affected.
    The proposed changes will not involve a significant increase in 
the probability of any event initiators. There will be no 
degradation in the performance of, or an increase in the number of 
challenges imposed on, safety-related equipment assumed to function 
during an accident situation. There will be no change to normal 
plant operating parameters or accident mitigation performance.
    The proposed changes will not alter any assumptions or change 
any mitigation actions in the radiological consequence evaluations 
in the FSAR.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve any physical alteration of 
the units. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
are no setpoints at which protective or mitigative actions are 
initiated that are affected by the proposed change. The proposed 
change will not alter the manner in which equipment operation is 
initiated, nor will the function demands on credited equipment be 
changed. No alteration in the procedures, which ensure the unit 
remains within analyzed limits, is proposed, and no change is being 
made to procedures relied upon to respond to an off-normal event. As 
such, no new failure modes are being introduced. The proposed change 
does not alter assumptions made in the safety analyses.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    (3) Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will not adversely affect operation of plant 
equipment and will not result in a change to the setpoints at which 
protective actions are initiated. None of the acceptance criteria 
for any accident analysis is changed. There will be no effect on the 
manner in which safety limits or limiting safety system settings are 
determined nor will there be any effect on those plant systems 
necessary to assure the accomplishment of protection functions. 
There will be no impact on the overpower limit, departure from 
nucleate boiling ratio (DNBR) limits, heat flux hot channel factor 
(FQ), nuclear enthalpy rise hot channel factor (FDH), loss of 
coolant accident peak cladding temperature (LOCA PCT), peak local 
power density, or any other margin of safety. The radiological dose 
consequence acceptance criteria listed in the Standard Review Plan 
will continue to be met.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and 2, Louisa County, Virginia

    Date of amendment request: December 13, 2002.
    Description of amendment request: The proposed amendments will 
extend the Completion Time of Technical Specification (TS) 3.8.7, 
Inverters-Operating, Required Action A.1, from 24 hours to 14 days for 
an inoperable inverter on either Train H or Train J.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:

    1. The proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change to extend the Completion Time for an 
inoperable inverter from 24 hours to 14 days does not alter any 
plant equipment or operating practices in such a manner that the 
probability of an accident is increased. In addition, this proposed 
change will not alter assumptions relative to the mitigation of an 
accident or transient event.
    The licensee performed an evaluation to determine the risk 
significance of the proposed change. This risk evaluation concluded 
that the increases in annual core damage frequency (CDF) and large 
early release frequency (LERF) associated with the proposed change 
can be characterized as ``very small changes'' by Regulatory Guide 
(RG) 1.174, ``An Approach for Using Probabilistic Risk Assessment in 
Risk-Informed Decisions on Plant-Specific Changes to the Licensing 
Basis.'' Additional evaluation by the licensee determined that the 
incremental conditional core damage probability (ICCDP) and 
incremental conditional large early release probability (ICLERP) 
associated with the proposed change are within the acceptance 
criteria in RG 1.177, ``An Approach for Plant-Specific, Risk-
Informed Decisionmaking: Technical Specifications.'' Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change to extend the Completion Time for an 
inoperable inverter has been evaluated for its effect on plant 
safety. The licensee's risk-informed evaluation concluded that the 
increases in annual CDF and LERF associated with the proposed change 
can be characterized as ``very small changes'' by RG 1.174. The

[[Page 18290]]

ICCDP and ICLERP associated with the proposed change are within the 
acceptance criteria in RG 1.177. Therefore, the proposed change does 
not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Dominion Nuclear Connecticut, Inc., Millstone Power Station, 
Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, 
Connecticut 06385.
    NRC Section Chief: John A. Nakoski.

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].

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

    Date of application for amendment: November 27, 2002.
    Brief description of amendment: This amendment deletes technical 
specification (TS) 5.5.3, ``Post Accident Sampling,'' and thereby 
eliminates the requirements to have and maintain the post accident 
sampling system at the Clinton Power Station, Unit 1. The amendment 
also addresses related changes to TS 5.5.2, ``Primary Coolant Sources 
Outside Containment.''
    Date of issuance: March 21, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days.
    Amendment No.: 155.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 21, 2003 (68 FR 
2797).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 21, 2003.
    No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of application for amendment: April 10, 2002, as supplemented 
February 12, 2003.
    Brief description of amendment: The amendment deleted Technical 
Specification 4.6.1.c, related to 24-month emergency diesel generator 
surveillance, and relocated these requirements to the Updated Final 
Safety Evaluation Report (UFSAR).
    Date of issuance: April 3, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days, including the relocation of the emergency diesel 
generator maintenance requirements of Technical Specification 4.6.1.c 
to the Updated Final Safety Analysis Report (UFSAR), as was described 
in the licensee's application dated April 10, 2002, and evaluated in 
the NRC staff's safety evaluation dated April 3, 2003, and which 
relocation shall be included in the next scheduled update of the UFSAR 
pursuant to 10 CFR 50.71(e).
    Amendment No.: 243.
    Facility Operating License No. DPR-50: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 28, 2002 (67 FR 
36926).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 3, 2003.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of application for amendments: July 24, 2002, as supplemented 
February 21, 2003.
    Brief Description of amendments: The amendments revise the 
Technical Specifications Section 3.1.7, ``Standby Liquid Control (SLC) 
System,'' to reflect modifications being made to the system as a result 
of transition to the GE14 fuel design.
    Date of issuance: March 25, 2003.
    Effective date: March 25, 2003.
    Amendment Nos.: 227 and 255.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications and Appendix B, ``Additional 
Conditions.''
    Date of initial notice in Federal Register: August 20, 2002 (67 FR 
53984).
    The February 21, 2003, supplement contained clarifying information 
only and did not change the initial no significant hazards 
consideration determination or expand the scope of the initial Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 25, 2003.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket No. 50-324, Brunswick Steam 
Electric Plant, Unit 2, Brunswick County, North Carolina

    Date of amendment request: November 7, 2002, as supplemented 
February 17, 2002.
    Brief description of amendment: The amendment revises the Minimum 
Critical Power Ratio (MCPR) Safety Limit contained in Technical 
Specification 2.1.1.2 from 1.09 to 1.11

[[Page 18291]]

for two recirculation loop operation and from 1.10 to 1.13 for single 
recirculation loop operation.
    Date of issuance: March 25, 2003.
    Effective date: March 25, 2003.
    Amendment No.: 254.
    Facility Operating License No. DPR-62: Amendment changes the 
Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75869). The February 17, 2003, supplement contained clarifying 
information only and did not change the initial no significant hazards 
consideration determination or expand the scope of the initial Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 25, 2003.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of application for amendment: August 28, 2002, as supplemented 
November 21, 2002.
    Brief description of amendment: This amendment revises the 
Technical Specifications (TS) by adding Topical Report EMF-2328 (P)(A), 
``PWR Small Break LOCA Evaluation Model, S-RELAP5 Based'' as reference 
in the TS to allow the licensee to update the methodologies that are 
used for safety analyses for the Shearon Harris Nuclear Power Plant, 
Unit 1. The amendment also relocates referenced methodologies within TS 
6.9.1.6.2 to group mechanical design methodologies together.
    Date of issuance: March 28, 2003.
    Effective date: March 28, 2003.
    Amendment No.: 114.
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 15, 2002 (67 FR 
63691). The November 21, 2002, supplement contained clarifying 
information only and did not change the initial proposed no significant 
hazards consideration determination or expand the scope of the initial 
application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 28, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 23, 2002, as supplemented 
December 20, 2002, and February 27, 2003.
    Brief description of amendment: The amendment revises the Fermi 2 
Technical Specifications (TSs) to allow a one-time deferral of the Type 
A primary containment integrated leak rate test. Specifically, TS 
5.5.12, ``Primary Containment Leakage Rate Testing Program,'' would be 
revised to extend the current interval for performing the containment 
Type A test to 15 years.
    Date of issuance: March 27, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 153.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 25, 2002 (67 FR 
42817).
    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 March 27, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 6, 2001, as 
supplemented on December 27, 2001, and July 15, August 6, and October 
29, 2002.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) associated with the spent fuel pool (SFP). 
Specifically, the amendment increases the allowable nominal average 
fuel assembly enrichment from 4.5 weight percent (w/o) Uranium-235 (U-
235) to 4.85 w/o U-235 for all regions of the SFP, the new fuel storage 
racks (dry), and the reactor core; allows fuel to be located in the 40 
storage cells in Region B of the SFP that are currently empty and 
blocked; credits SFP soluble boron for reactivity control during normal 
conditions; and reduces the Boraflex reactivity credit in Regions A and 
B of the SFP.
    Date of issuance: April 1, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 274.
    Facility Operating License No. DPR-65: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: February 19, 2002 (67 
FR 7414). The supplement dated December 27, 2001, provided a revision 
to the licensee's analysis of the issue of no significant hazards 
consideration, as originally provided in the November 6, 2001, 
application. The supplements dated July 15, August 6, and October 29, 
2002, 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 1, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 3, 2002, as supplemented on 
January 23, 2003.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.4.9, ``Pressurizer,'' to increase the pressurizer 
water level limit when the plant is in MODE 3 (Hot Standby). The 
pressurizer water level limit for MODES 1 and 2 (Power Operation and 
Startup) remains unchanged. The amendment also revises TS 3.8.4, ``DC 
Sources--Operating,'' to remove the notes that refer to the one-time 
amendment allowing the online replacement of station batteries 31 and 
32. The notes were no longer applicable since the batteries have been 
replaced.
    Date of issuance: March 25, 2003.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 216.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 9, 2002 (67 FR 
45566).
    The January 23 letter provided clarifying information that did not 
enlarge the scope of the original Federal Register notice or change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 25, 2003.
    No significant hazards consideration comments received: No.

[[Page 18292]]

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

    Date of application for amendment: December 4, 2002, which replaces 
the original applications dated May 1, 2002.
    Brief description of amendment: The proposed amendment would change 
the Pilgrim Nuclear Power Station Technical Specification (TS) Figures 
3.6.1, 3.6.2, and 3.6.3 to extend the applicability of the current 
reactor pressure vessel pressure-temperature (P-T) curves through the 
end of Operating Cycle (OC) 16. The current P-T curves were approved 
for use in License Amendment 190, dated April 13, 2001, and are limited 
to use through the end of OC 14. The proposed change would delete the 
20 and 32 Effective Full Power Year curves and replace the wording of 
the title blocks to allow use through the end of OC 16.
    Date of issuance: March 28, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 197.
    Facility Operating License No. DPR-35: This amendment revised the 
TS.
    Date of initial notice in Federal Register: February 18, 2003 (68 
FR 7816).
    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.

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

    Date of amendment request: August 19, 2002, as supplemented by 
letter dated December 19, 2002.
    Brief description of amendment: The amendment revises the Technical 
Specifications by: (1) Modifying the wording of the current 
Surveillance Requirement (SR) 4.0.1 and SR 4.0.3 to be consistent with 
NUREG-1431, Revision 2, Improved Standard Technical Specifications 
(ISTS) wording for SR 3.0.1 and SR 3.0.3; and (2) modifying the ISTS 
wording, adopted in Item (1), above, for SR 4.0.3 to extend the delay 
period, before entering a Limiting Condition for Operation, following a 
missed surveillance. The delay period is extended from the current 
limit of up to 24 hours ``* * * when the allowable outage time limits 
of the ACTION requirements are less than 24 hours'' to ``* * * up to 24 
hours or up to the limit of the specified surveillance interval, 
whichever is greater.'' In addition, the following requirement is added 
to SR 4.0.3: ``A risk evaluation shall be performed for any 
Surveillance delayed greater than 24 hours and the risk impact shall be 
managed.''
    Date of issuance: March 21, 2003.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 187.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications and Surveillance Requirements.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5670).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 21, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 20, 2002.
    Brief description of amendment: This amendment approves several 
administrative changes to the Waterford Steam Electric Station, Unit 3 
Technical Specifications (TSs) to revise, correct, or clarify certain 
titles, page numbers, and heading information. It also revises 
personnel and committee titles that have been changed, revises 
administrative reporting requirements to conform to 10 CFR 50.4, and 
deletes redundant or unnecessary requirements from TSs 5.4, 6.6, and 
6.7.
    Date of issuance: April 3, 2003.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 188.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5673).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 3, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 16, 2002.
    Brief description of amendments: The amendments modify the Unit 3 
allowable value Technical Specification, and the Units 2 and 3 
surveillance requirements Technical Specification for the reactor 
protection system scram discharge volume water level-high function.
    Date of issuance: April 3, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 198/191.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68737).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 3, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 27, 2002.
    Brief description of amendments: These amendments delete Technical 
Specification (TS) 5.5.3, ``Post Accident Sampling,'' and thereby 
eliminate the requirements to have and maintain the post accident 
sampling system at the LaSalle County Station, Units 1 and 2. The 
amendments also address related changes to TS 5.5.2, ``Primary Coolant 
Sources Outside Containment.''
    Date of issuance: March 21, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days.
    Amendment Nos.: 158/144.
    Facility Operating License Nos. NPF-11 and NPF-18: 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 March 21, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 22, 2002.
    Brief description of amendments: The amendments modify the required 
surveillance interval from monthly to quarterly for calibration of the 
trip units associated with the instrumentation channels of the 
Anticipated Transient Without Scram-Recirculation Pump Trip system.
    Date of issuance: April 1, 2003.

[[Page 18293]]

    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 213 and 207.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 1, 2002 (67 FR 
61682). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 1, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 16, 2002, as 
supplemented January 28, 2003.
    Brief description of amendment: The amendment would revise the 
Technical Specification values for the 4 kilovolt degraded-voltage and 
loss-of-voltage relays.
    Date of issuance: March 26, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 256.
    Facility Operating License No. DPR-74: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68739).
    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 March 26, 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: December 19, 2002.
    Brief description of amendment: The amendment revised the Technical 
Specifications to add the definition of shutdown margin (SDM), 
incorporate new, more restrictive SDM limits, add the associated 
limiting condition for operation actions and completion times for each 
applicable operating condition if the SDM is not met, and add 
surveillance requirements for verifying SDM.
    Date of issuance: March 27, 2003.
    Effective date: March 27, 2003.
    Amendment No.: 180.
    Facility Operating License No. DPR-63: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 21, 2003 (68 FR 
2806).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 27, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 29, 2002, as supplemented 
by letter dated January 24, 2003.
    Brief description of amendment: The amendment changes the 
surveillance requirement of TS 5.5.12, ``Primary Containment Leakage 
Rate Testing Program,'' to allow a one-time 5-year extension to the 10-
year interval for performing the next Type A containment integrated 
leakage rate test (ILRT). The change allows ILRT testing within 15 
years from the last ILRT, which was performed in September 1993.
    Date of issuance: March 21, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 249.
    Facility Operating License No. DPR 49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 30, 2002 (67 FR 
21291).
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 21, 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: April 22, 2002, as supplemented 
October 25, 2002, January 23, and February 12, 2003.
    Brief description of amendment: The amendment changes TS 
Surveillance Requirement 4.7.A.2.b, ``Primary Containment Integrity,'' 
to allow a one-time, 5-year extension to the 10-year interval for 
performing the next Type A containment integrated leakage rate test 
(ILRT). The change allows ILRT testing within 15 years from the last 
ILRT, which was performed in March 1993.
    Date of issuance: March 31, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 134.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 3, 2002 (67 
FR 56324).
    The October 25, 2002, January 23, and February 12, 2003, 
supplements 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 March 31, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 25, 2002, as supplemented 
on October 21, 2002.
    Brief description of amendment: The amendment would modify 
Technical Specification (TS) requirements for missed surveillance tests 
in TS 4.0.3 using the Consolidated Line Item Improvement Program, 
modify TS 4.0.1 to be consistent with the Standard Technical 
Specifications (STS), and incorporate a TS Bases Control Program in 
Section 6.0 in accordance with the STS.
    Date of issuance: March 31, 2003.
    Effective date: March 31, 2003.
    Amendment No.: 145.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75883)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 31, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendments request: April 4, 2002, as supplemented by 
letter dated January 9, 2003.
    Brief Description of amendments: The amendments revise Technical

[[Page 18294]]

Specifications 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 1994 
for Unit 1 and March 1995 for Unit 2.
    Date of issuance: March 21, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 159/150.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68743).
    The supplement, dated January 9, 2003, provided clarifying 
information that did not change the scope of the April 4, 2002, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 21, 2003.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant, 
Unit 2, Hamilton County, Tennessee

    Date of application for amendment: November 15, 2002, as 
supplemented by letters dated February 19, 2003, and February 26, 2003.
    Description of amendment: This one-time condition establishes 
special provisions and requirements for safe operation of Unit 2 while 
heavy load lifts are performed during the Unit 1 steam generator 
replacement project. The provisions for heavy load lifts are described 
in Topical Report 24370-TR-C-002, which was previously submitted on 
April 15, 2002, for NRC review and approval. The topical report 
contains prerequisite actions for heavy load movement, active 
monitoring during heavy load movement, and compensatory measures in 
response to the unlikely event of a heavy load drop.
    Date of issuance: March 26, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days of issuance.
    Amendment No.: 273.
    Facility Operating License No. DPR-79: Amendment revises the 
Operating License.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75885).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 26, 2003.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 7th day of April, 2003.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 03-9026 Filed 4-14-03; 8:45 am]
BILLING CODE 7590-01-P