[Federal Register Volume 67, Number 141 (Tuesday, July 23, 2002)]
[Notices]
[Pages 48213-48225]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-18242]


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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 June 28, 2002, through July 11, 2002. The 
last biweekly notice was published on July 9, 2002 (67 FR 45560).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. The filing of requests for a 
hearing and petitions for leave to intervene is discussed below.
    By August 22, 2002, 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,\1\

[[Page 48214]]

which is available at the Commission's PDR, located at One White Flint 
North, 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.
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    \1\ The most recent version of Title 10 of the Code of Federal 
Regulations, published January 1, 2002, inadvertently omitted the 
last sentence of 10 CFR 2.714(d) and subparagraphs (d)(1) and (2), 
regarding petitions to intervene and contentions. Those provisions 
are extant and still applicable to petitions to intervene. Those 
provisions are as follows: ``In all other circumstances, such ruling 
body or officer shall, in ruling on--
    (1) A petition for leave to intervene or a request for hearing, 
consider the following factors, among other things:
    (i) The nature of the petitioner's right under the Act to be 
made a party to the proceeding.
    (ii) The nature and extent of the petitioner's property, 
financial, or other interest in the proceeding.
    (iii) The possible effect of any order that may be entered in 
the proceeding on the petitioner's interest.
    (2) The admissibility of a contention, refuse to admit a 
contention if:
    (i) The contention and supporting material fail to satisfy the 
requirements of paragraph (b)(2) of this section; or
    (ii) The contention, if proven, would be of no consequence in 
the proceeding because it would not entitle petitioner to relief.''
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    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, 11555 Rockville Pike (first floor), 
Rockville, Maryland, by the above date. Because of continuing 
disruptions in delivery of mail to United States Government offices, it 
is requested that petitions for leave to intervene and requests for 
hearing be transmitted to the Secretary of the Commission either by 
means of facsimile transmission to 301-415-1101 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
and because of continuing disruptions in delivery of mail to United 
States Government offices, it is requested that copies be transmitted 
either by means of facsimile transmission to 301-415-3725 or by e-mail 
to [email protected]. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the attorney for 
the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, 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, 304-415-4737 or by e-mail to [email protected].

[[Page 48215]]

Dominion Nuclear Connecticut, Inc., Docket No. 50-245, Millstone Power 
Station, Unit No. 1, New London County, Connecticut

    Date of amendment request: May 13, 2002.
    Description of amendment request: The proposed amendment modifies 
the Millstone Nuclear Power Station, Unit No. 1 (MP1) Permanently 
Defueled Technical Specifications (TSs) to change selected MP1 
radiological related TSs. These changes are due to the revision to part 
20 of Title 10 of the Code of Federal Regulations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    It is proposed to revise the Occupational Radiation Exposure 
Report, Radioactive Effluent Controls Program, and High Radiation 
Area Specifications in accordance with TSTF [Technical Specification 
Task Force] travelers 152, 258 and 308, to reflect changes due to 
the revision to 10 CFR part 20.
    These changes do not have an impact on the acceptance criteria 
for any design basis accident described in the Unit No. 1 Defueled 
Safety Analysis Report (DSAR).
    The changes have no impact on plant equipment operation. Since 
the changes are administrative or editorial in nature they cannot 
affect the likelihood or consequences of accidents. Therefore, the 
proposed changes will not increase the probability or consequences 
of an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The revisions to the Occupational Radiation Exposure Report, 
Radioactive Effluent Controls Program, and High Radiation Area 
Specifications in accordance with TSTF travelers 152, 258 and 308 
will have no effect on plant operation. Since the proposed changes 
are solely administrative or editorial in nature, they do not affect 
plant operation in any way.
    The proposed changes do not involve a physical alteration of the 
plant or change the plant configuration (no new or different type of 
equipment will be installed). The proposed changes do not require 
any new or unusual operator actions. The changes do not alter the 
way any structure, system, or component functions and do not alter 
the manner in which the plant is operated. The changes do not 
introduce any new failure modes. Therefore, the proposed changes 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    Since the proposed changes are solely administrative or 
editorial changes to the TS, they do not affect plant operation in 
any way. The proposed changes to each unit's technical 
specifications will revise them to reflect the requirements of the 
current 10 CFR Part 20, standardize terminology, provide clearer 
guidance, clarify inconsistencies, remove extraneous information, 
and result in minor format changes that will not result in any 
technical changes to current requirements.
    The proposed changes have no effect on any safety analyses 
assumptions and therefore [do] not impact any margins of safety. The 
proposed changes do not impact any acceptance criteria for the 
design basis accidents described in the Unit No. 1 DSAR and [do] not 
impact the consequences of accidents previously evaluated. 
Therefore, the proposed changes will not result in a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, CT 06385.
    NRC Section Chief: Stephen Dembek.

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: May 9, 2002.
    Description of amendment request: The amendments would revise the 
licensing basis Steam Generator Tube Rupture sequences for Catawba 
Nuclear Station, Units 1 and 2. Specifically, it is requested that a 
certain single failure scenario potentially leading to steam generator 
overfill be excluded from the design basis steam generator tube rupture 
analysis using the guidance of Regulatory Guide 1.174, ``An Approach 
for Using Probabilistic Risk Assessment in Risk-Informed Decisions on 
Plant Specific Changes to the Licensing Basis.''
    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:

    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?
    No. This proposed amendment requests that steam generator tube 
rupture sequences involving a failure of 125 VDC Distribution Center 
EDE [or EDF] be excluded from consideration in the analysis of the 
design basis steam generator tube rupture event. These sequences 
involve a single failure that potentially degrades the ability to 
terminate auxiliary feedwater flow into a ruptured steam generator 
following a steam generator tube rupture. The inability to terminate 
auxiliary feedwater flow in a timely manner following a steam 
generator tube rupture could result in steam generator overfill.
    The sequences to be excluded do not involve equipment that can 
be considered an accident initiator. Implementation of this 
amendment does not involve any physical changes to the facility. It 
does not affect basic operation of the facility. The probability of 
occurrence of a steam generator tube rupture or any other accident 
previously evaluated will not change following implementation of 
this amendment.
    Elimination of certain sequences from the design basis steam 
generator tube rupture analysis does not adversely affect the 
ability to cool the reactor core and prevent core damage following a 
steam generator tube rupture. The Departure from Nucleate Boiling 
ratio is not adversely impacted.
    The ability to maintain a secondary heat sink and provide water 
to the Reactor Coolant System for makeup, cooling of the core, and 
shutdown margin following a design basis steam generator tube 
rupture is not affected by the changes proposed in this license 
amendment. Neither fuel damage nor clad damage is expected to occur 
for the steam generator tube rupture sequences to be eliminated.
    Should the ruptured steam generator overfill following a design 
basis steam generator tube rupture in one of the sequences to be 
excluded, radioactivity could be released to the environment in 
increased amounts and over a longer time span than predicted in the 
safety analysis. The frequency of occurrence of these steam 
generator tube rupture sequences is low. Should such an event occur, 
the radiological consequences are expected to be below the 
guidelines of 10 CFR 100 and General Design Criteria 19. Under 
nominal conditions, (e.g., nominal atmospheric dispersion factors, 
nominal levels of radioactivity in the Reactor Coolant System, 
etc.), radiological consequences of a steam generator tube rupture 
would be small compared to even the guideline values of the Standard 
Review Plan, Section 15.6.3. There is no significant adverse effect 
on the mitigation of consequences following a steam generator tube 
rupture.
    In summary, operation of the facility in accordance with the 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    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?
    No. The proposed amendment involves elimination of certain 
sequences from the

[[Page 48216]]

design basis steam generator tube rupture analysis. No physical 
changes to the facility are associated with the proposed amendment.
    The sequences to be eliminated involve single failures that 
could adversely affect the ability to terminate auxiliary feedwater 
flow to a ruptured steam generator. The failures associated with 
these sequences are not accident sequence precursors and do not have 
an adverse impact on any accident initiator.
    No new failure modes are created due to implementation of the 
change proposed in this License Amendment Request. Therefore, 
operation of the facility in accordance with the changes proposed in 
this License Amendment Request does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Does operation of the facility in accordance with the proposed 
amendment involve a significant reduction in a margin of safety?
    No. One of the standards by which the consequences of the design 
basis steam generator tube rupture are evaluated is that the 
Departure from Nucleate Boiling Ratio (DNBR) is greater than the 
limit value. Should one of the steam generator tube rupture 
sequences to be excluded occur, the effects relative to steam 
generator overfill would not be manifested until the Control Room 
operators attempt to stop the flow of auxiliary feedwater to the 
ruptured steam generator which is well into the event. The minimum 
DNBR would occur within seconds after reactor trip. Therefore, the 
criterion concerning DNBR is met.
    The risk evaluation demonstrates that the frequency of steam 
generator overfill associated with the steam generator tube rupture 
sequences to be excluded is low (approximately 3.7 E-11 per reactor 
year per Class 1E Train). Additionally, the frequency of a large 
early release is shown to be very low (approximately 3.7 E-15 per 
reactor year per Class 1E Train).
    It is concluded that removal of certain steam generator tube 
rupture sequences from the plant licensing basis as proposed does 
not constitute a significant reduction in a margin of safety.
    Based on this evaluation, it is concluded that operation of the 
facility in accordance with the proposed amendment constitutes no 
significant hazard to the public.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of amendment request: June 7, 2002.
    Description of amendment request: The proposed amendments would 
revise the Updated Final Safety Analysis Report to eliminate credit for 
the flow path from the spent fuel pool to high pressure injection pump 
as one source of primary system makeup following a tornado. The 
proposed amendments would also credit the Standby Shutdown Facility as 
the assured means of achieving safe shutdown for all three Oconee units 
following a tornado.
    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.
    The changes being requested in this amendment request involve 
(1) the elimination of the Spent Fuel Pool (SFP) as a suction source 
to a High Pressure Injection [HPI] pump for primary system make-up, 
and (2) to fully credit the Standby Shutdown Facility (SSF) as the 
primary assured means of achieving safe shutdown of all three units 
following a tornado. Following the modification to fully tornado 
protect the SSF, this facility becomes the station's assured flow 
path for both primary make-up and secondary decay heat removal for 
all three units.
    Although the probability of a severe tornado strike at the 
station does not change, new tornado insights gained from a review 
of the current external event risk analysis have resulted in an 
enhanced risk model that more accurately characterizes station 
tornado damage risk. The proposed changes are part of the revised 
tornado mitigation strategy that provides for an assured, 
deterministic success path rather than the current strategy that is 
based on risk insights and diversity for achieving safe shutdown. 
This effort has resulted in an overall reduction in tornado risk at 
the station and consequently, would not result in a significant 
increase in the consequences of an accident previously evaluated.
    Other than the fortification of walls of existing structures to 
harden them against tornado damage, there are no physical changes to 
the plant structures, systems, or components (SSCs) or operating 
procedures, nor are there any changes to safety limits or set 
points. Also, no new radiological release pathways are created.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The changes being proposed in this amendment request do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated. The initial placement of the SFP-
HPI flow path into the LB (Licensing Bases) was based on 1989 risk 
analyses that showed a potential need for primary make-up due to 
inventory losses from a reactor coolant pump (RCP) seal loss-of-
cooling accident (LOCA). The upgrade of the RCP seals has 
significantly reduced the probability of a seal LOCA and 
subsequently, alleviated the initial reliance on the SFP-HPI flow 
path for primary make-up. If multi-unit primary make-up and decay 
heat removal are required following an event, the tornado protected 
SSF RB[C]MU (Reactor Coolant Makeup) or SSF ASW (Auxiliary Service 
Water) pumps have the capabilities to perform these functions for 
all three units.
    Other than the fortification of walls of existing structures to 
harden them against tornado damage, there are no physical changes to 
the plant SSCs or operating procedures. There are no new hazardous 
materials or potential missiles. It does not introduce the 
possibility of any new or different malfunctions. No safety limits 
or set points are changed.
    3. Involve a significant reduction in a margin or safety.
    As mentioned previously, new tornado insights gained from a 
review of the current external event risk analysis have resulted in 
an enhanced risk model that more accurately characterizes station 
tornado damage risk. The proposed changes are part of the revised 
tornado mitigation strategy that provides for an assured, 
deterministic success path rather than a strategy that is based on 
risk insights and diversity for achieving safe shutdown.
    There are no safety limit, set point, design parameters, or 
operating procedure changes required. The integrity of the fuel 
cladding, reactor coolant system, and containment are preserved. 
Thus, 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: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: May 14, 2002.
    Description of amendment request: The proposed amendment would 
revise Surveillance Requirement (SR) 4.0.3 to extend the delay period, 
before entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period would be extended from

[[Page 48217]]

the current limit of ``* * * up to 24 hours to permit the completion of 
the surveillance 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 interval, whichever is greater.'' In 
addition, the following requirement would be added to SR 4.0.3: ``A 
risk evaluation shall be performed for any Surveillance delayed greater 
than 24 hours and the risk impact shall be managed.''
    The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice 
of opportunity for comment in the Federal Register on June 14, 2001 (66 
FR 32400), on possible amendments concerning missed surveillances, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on September 28, 2001 (66 FR 
49714). The licensee affirmed the applicability of the following NSHC 
determination in its application dated May 14, 2002.
    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 proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an 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 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. A 
missed surveillance will not, in and of itself, introduce new 
failure modes or effects and any increased chance that a standby 
system might fail to perform its safety function due to a missed 
surveillance would not, in the absence of other unrelated failures, 
lead to an accident beyond those previously evaluated. The addition 
of a requirement to assess and manage the risk introduced by the 
missed surveillance will further minimize possible concerns. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.

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

    The extended time allowed to perform a missed surveillance does 
not result in a significant reduction in the margin of safety. As 
supported by the historical data, the likely outcome of any 
surveillance is verification that the LCO (Limiting Condition for 
Operation) is met. Failure to perform a surveillance within the 
prescribed frequency does not cause equipment to become inoperable. 
The only effect of the additional time allowed to perform a missed 
surveillance on the margin of safety is the extension of the time 
until inoperable equipment is discovered to be inoperable by the 
missed surveillance. However, given the rare occurrence of 
inoperable equipment, and the rare occurrence of a missed 
surveillance, a missed surveillance on inoperable equipment would be 
very unlikely. This must be balanced against the real risk of 
manipulating the plant equipment or condition to perform the missed 
surveillance. In addition, parallel trains and alternate equipment 
are typically available to perform the safety function of the 
equipment not tested. Thus, there is confidence that the equipment 
can perform its assumed safety function.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

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

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

    Date of amendment request: June 12, 2002.
    Description of amendment request: The proposed amendment would 
revise Surveillance Requirement 3.0.3 to extend the delay period, 
before entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period would be extended from the current limit 
of ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is less'' to ``* * * up to 24 hours or up to the 
limit of the specified Frequency, whichever is greater.'' In addition, 
the following requirement would be added to SR 3.0.3: ``A risk 
evaluation shall be performed for any Surveillance delayed greater than 
24 hours and the risk impact shall be managed.''
    The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice 
of opportunity for comment in the Federal Register on June 14, 2001 (66 
FR 32400), on possible amendments concerning missed surveillances, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on September 28, 2001 (66 FR 
49714). The licensee affirmed the applicability of the following NSHC 
determination in its application dated June 12, 2002.
    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 proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a

[[Page 48218]]

standby system might fail to perform its safety function due to a 
missed surveillance is small and would not, in the absence of other 
unrelated failures, lead to an increase in consequences beyond those 
estimated by existing analyses. The addition of a requirement to 
assess and manage the risk introduced by the missed surveillance 
will further minimize possible concerns. Therefore, this change does 
not involve a significant increase in the probability or 
consequences of an 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 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. A 
missed surveillance will not, in and of itself, introduce new 
failure modes or effects and any increased chance that a standby 
system might fail to perform its safety function due to a missed 
surveillance would not, in the absence of other unrelated failures, 
lead to an accident beyond those previously evaluated. The addition 
of a requirement to assess and manage the risk introduced by the 
missed surveillance will further minimize possible concerns. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.

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

    The extended time allowed to perform a missed surveillance does 
not result in a significant reduction in the margin of safety. As 
supported by the historical data, the likely outcome of any 
surveillance is verification that the LCO [Limiting Condition for 
Operation] is met. Failure to perform a surveillance within the 
prescribed frequency does not cause equipment to become inoperable. 
The only effect of the additional time allowed to perform a missed 
surveillance on the margin of safety is the extension of the time 
until inoperable equipment is discovered to be inoperable by the 
missed surveillance. However, given the rare occurrence of 
inoperable equipment, and the rare occurrence of a missed 
surveillance, a missed surveillance on inoperable equipment would be 
very unlikely. This must be balanced against the real risk of 
manipulating the plant equipment or condition to perform the missed 
surveillance. In addition, parallel trains and alternate equipment 
are typically available to perform the safety function of the 
equipment not tested. Thus, there is confidence that the equipment 
can perform its assumed safety function.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

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

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

    Date of amendment request: May 31, 2001.
    Description of amendment request: The proposed amendments would 
change Appendix A, Technical Specifications (TS), of Facility Operating 
License Nos. NPF-11 and NPF-18. Specifically, the proposed change 
modifies TS Surveillance Requirement (SR) 3.6.1.3.8 to reduce to number 
of excess flow check valves (EFCVs) required to be tested every 24 
months. The proposed SR will require that a representative sample of 
reactor instrumentation line EFCVs actuate to the isolation position on 
an actual or simulated instrumentation line break signal every 24 
months. All reactor instrumentation line EFCVs will be tested at least 
once every 10 years (nominal). The proposed change implements Technical 
Specification Task Force Traveler 334 (TSTF-334), ``Relaxed 
Surveillance Frequency for Excess Flow Check Valve Testing,'' 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:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change to LaSalle County Station, Unit 1 and Unit 2 
Technical Specifications (TS) modifies TS Surveillance Requirement 
(SR) 3.6.1.3.8 to reduce the number of excess flow check valves 
(EFCVs) required to be tested every 24 months. The proposed SR will 
require that a representative sample of reactor instrumentation line 
EFCVs actuate to the isolation position on an actual or simulated 
instrumentation line break signal every 24 months. All reactor 
instrumentation line EFCVs will be tested at least once every 10 
years (nominal).
    The performance of EFCV surveillance testing is not a precursor 
to any accident previously evaluated and is not related to the 
frequency of instrument line failures. Thus, the proposed change to 
modify the test frequency associated with EFCV surveillance does not 
have any effect on the probability of an accident previously 
evaluated.
    The performance of the EFCV surveillance testing does provide 
assurance that the EFCV will perform as designed. The LaSalle County 
Station radiological dose assessment for an instrument line break is 
documented in the LaSalle County Station UFSAR Table 15.6-4, 
``Instrument Line Break Radiological Effects.'' The assessment does 
not credit performance of the EFCV to limit instrument line flows 
during an assumed break. These estimated doses are significantly 
below the regulatory dose limits listed in 10 CFR 100, ``Reactor 
Site Criteria.'' The proposed change does not change the assumptions 
or the estimated doses associated with a LaSalle County Station 
instrument line break. Thus, the radiological consequences of any 
accident previously evaluated are not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change modifies TS SR 3.6.1.3.8 to reduce the 
number of excess flow check valves (EFCVs) required to be tested 
every 24 months while requiring all EFCVs to be tested at least once 
every 10 years (nominal). The proposed change does not affect the 
performance of any LaSalle County Station structure, system, or 
component credited with mitigating any accident previously 
evaluated. The proposed change to modify the surveillance will not 
affect the control parameters governing unit operation or the 
response of plant equipment to transient conditions. The proposed 
change does not introduce any new equipment, modes of system 
operation or failure mechanisms.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed change for LaSalle County Station, Units 1 and 2, 
implements Technical Specification Task Force Traveler 334 (TSTF-
334), ``Relaxed Surveillance Frequency for Excess Flow Check Valve 
Testing,'' Revision 2. TSTF-334 notes that its implementation is 
only allowed for plants for which General Electric Nuclear Energy 
Topical Report NEDO-32977-A, ``Excess Flow Check Valve Testing 
Relaxation,'' is applicable. In addition, an EFCV performance 
criteria and basis must be developed to ensure that the corrective 
action program can provide meaningful feedback for appropriate 
corrective actions.
    LaSalle County Station, in accordance with Topical Report NEDO-
32977-A, has performed a plant-specific radiological dose assessment 
for an instrument line break, EFCV failure rate analysis, release 
frequency initiated by an instrument line break analysis and has 
proposed a corrective action program

[[Page 48219]]

to ensure continued EFCV performance. The result of the assessment 
and analyses meets the overall requirements to allow implementation 
TSTF-334, Revision 2.
    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 
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.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of amendment request: June 28, 2002.
    Description of amendment request: The proposed amendment would 
revise the Unit 1 Operating License and Technical Specifications to 
increase the licensed power level to 3304 megawatts thermal (MWt), or 
1.66 percent greater than the current licensed power level of 3250 MWt.
    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 of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    Probability of Occurrence of an Accident Previously Evaluated--
In support of this measurement uncertainty recapture power uprate, a 
comprehensive evaluation was performed for nuclear steam supply 
system (NSSS) and balance of plant (BOP) components and analyses 
that could be affected by this change. A power calorimetric 
uncertainty calculation was performed, and the effect of increasing 
plant power by 1.66 percent on the plant's design and licensing 
basis was evaluated. The result of these evaluations is that all 
plant components will continue to be capable of performing their 
design function at an uprated core power of 3304 megawatts thermal 
(MWt). In addition, an evaluation of the accident analyses 
demonstrates that applicable analysis acceptance criteria continue 
to be met. No accident initiators are affected by this uprate and no 
challenges to any plant safety barriers are created by this change.
    Consequences of an Accident Previously Evaluated--This change 
does not affect the release paths, the frequency of release, or the 
source term for release for any accidents previously evaluated in 
the Updated Final Safety Analysis Report. Structures, systems, and 
components (SSC) required to mitigate transients remain capable of 
performing their design functions, and thus were found acceptable. 
The reduced uncertainty in the feedwater flow input to the power 
calorimetric measurement ensures that applicable accident analyses 
acceptance criteria continue to be met, to support operation at a 
core power of 3304 MWt. Analyses performed to assess the effects of 
mass and energy remain valid. The source terms used to assess 
radiological consequences have been reviewed and determined to bound 
operation at the uprated condition.
    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 scenarios, failure mechanisms, or single 
failures are introduced as a result of the proposed changes. The 
installation of the Caldon Leading Edge Flow Meter (LEFM) 
CheckPlusTM system has been analyzed, and failures of 
this system will have no adverse effect on any safety-related system 
or any SSCs required for transient mitigation. SSCs previously 
required for the mitigation of a transient remain capable of 
fulfilling their intended design functions. The proposed changes 
have no adverse effects on any safety-related system or component 
and do not challenge the performance or integrity of any safety-
related system.
    This change does not adversely affect any current system 
interfaces or create any new interfaces that could result in an 
accident or malfunction of a different kind than previously 
evaluated. Operating at a core power level of 3304 MWt does not 
create any new accident initiators or precursors. The reduced 
uncertainty in the feedwater flow input to the power calorimetric 
measurement ensures that applicable accident analyses acceptance 
criteria continue to be met, to support operation at a core power of 
3304 MWt. Credible malfunctions continue to be bounded by the 
current accident analysis of record or re-analysis demonstrates that 
applicable acceptance criteria continue to be met.
    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 margins of safety associated with this Measurement 
Uncertainty Recapture Uprate Program are those pertaining to core 
power. This includes those associated with the fuel cladding, 
Reactor Coolant System (RCS) pressure boundary, and containment 
barriers. A comprehensive engineering review was performed to 
evaluate the 1.66 percent increase in the licensed core power from 
3250 MWt to 3304 MWt. The 1.66 percent increase required that 
revised NSSS design thermal and hydraulic parameters be established, 
which then served as the basis for all of the NSSS analyses and 
evaluations. This engineering review concluded that no design 
transient modifications are required to accommodate the revised NSSS 
design conditions. NSSS systems and components were evaluated and it 
was concluded that the NSSS equipment has sufficient margin to 
accommodate the 1.66 percent power uprate. NSSS accident analyses 
were either evaluated or revised for the 1.66 percent power uprate. 
In all cases the evaluations and re-analyses demonstrate that the 
applicable analyses acceptance criteria continue to be met. As such, 
the margins of safety continue to be bounded by the current analyses 
of record for this change.
    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: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: June 28, 2002.
    Description of amendment request: The proposed amendment would 
delete requirements from the Technical Specifications (TSs) and, as 
applicable, other elements of the licensing bases to maintain a Post-
Accident Sampling System (PASS). Licensees were generally required to 
implement PASS upgrades as described in NUREG-0737, ``Clarification of 
TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory 
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Assess Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades was an outcome of the 
lessons learned from the accident that occurred at TMI, Unit 2. 
Requirements related to PASS were imposed by Order for many facilities 
and were added to, or included in, the TSs for nuclear power reactors 
currently licensed to operate. However, lessons learned and 
improvements implemented over the last 20 years have shown that the 
information obtained from PASS can be readily obtained

[[Page 48220]]

through other means, or is of little use in the assessment and 
mitigation of accident conditions.
    The Nuclear Regulatory Commission (NRC) staff issued a notice of 
opportunity for comment in the Federal Register on December 27, 2001 
(66 FR 66949) on possible amendments to eliminate PASS, including a 
model safety evaluation and model no significant hazards consideration 
(NSHC) determination, using the consolidated line item improvement 
process. The NRC staff subsequently issued a notice of availability of 
the models for referencing in license amendment applications in the 
Federal Register on March 20, 2002 (67 FR 13027). The licensee affirmed 
the applicability of the NSHC determination in its application dated 
June 28, 2002. The NSHC determination is restated below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC 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 [a] 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 [a] margin of safety.

    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard J. Laufer.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: June 7, 2002.
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Nuclear Power Plant Technical Specification (TS) 
Sections for administrative changes: (1) Section 1--``Definitions,'' 
(2) Section 2--``Safety Limits and Limiting Safety System Settings,'' 
(3) Section 5--``Design Features,'' and (4) Section 6--``Administrative 
Controls.'' The administrative changes include capitalizing defined 
words, formatting section titles, renumbering pages and correcting 
miscellaneous grammar and punctuation errors.
    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.
    The proposed changes will not alter the intent of the TS. 
Reformatting the TS sections and correcting typographical, 
grammatical and format inconsistencies are administrative in nature. 
There is no impact on accident initiators or plant equipment, and 
therefore does not affect the probability or consequences of an 
accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not involve a change to the physical 
plant or operations. Since these are administrative changes they do 
not contribute to accident initiation. Therefore, the proposed 
changes do not produce a new accident scenario or produce a new type 
of equipment malfunction.
    3. Involve a significant reduction in the margin of safety.
    Since these are administrative changes, they do not involve a 
significant reduction in the margin of safety. The proposed changes 
do not affect plant equipment or operation. Safety limits and 
limiting safety system settings are not affected by this change.

    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.

[[Page 48221]]

    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: L. Raghavan.

South Carolina Electric & Gas Company (SCE&G), South Carolina Public 
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, 
Unit No. 1, Fairfield County, South Carolina

    Date of amendment request: June 27, 2002.
    Description of amendment request: SCE&G is proposing a revision to 
the Technical Specifications (TS) for the Virgil C. Summer Nuclear 
Station (VCSNS) to add an Allowed Outage Time (AOT) to Table 3.3-3, 
Engineered Safety Features Actuation System (ESFAS) instrumentation, 
Action Statement 16.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    South Carolina Electric & Gas Company (SCE&G) has evaluated the 
proposed changes to the VCSNS TS described above against the 
Significant Hazards Criteria of 10 CFR 50.92 and has determined that 
the changes do not involve any significant hazard. The following is 
provided in support of this conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The addition of an ACTION STATEMENT and the addition of an AOT 
(and its associated actions if not met) for a TS action statement 
are neither an accident initiator or precursor. The ESFAS actuates 
in response to an accident and has a mitigating function. Increasing 
the TS requirements for specific TS instrument loops provides 
additional assurance that the channels will be capable of performing 
their design function in the event of a DBA [design-basis accident]. 
The ability of the operations staff to respond to an evaluated 
accident or plant transient will not be hampered. This change 
provides conservative requirements to assure that the design basis 
of the plant is maintained.
    Addition of conservative changes to the Engineered Safety 
Feature Actuation System Instrumentation [does] not contribute to 
the initiation of any accident evaluated in the FSAR [Final Safety 
Analysis Report]. Supporting factors are as follows:

--The changes provide consistency between Tables 3.3-2, 3.3-3, and 
4.3-2, resulting in a one-for-one correlation between the functional 
units in those tables. These changes are conservative and consistent 
with the Standard Technical Specifications, NUREG-1431, Rev. 2.

    There are no deletions from the Technical Specifications made by 
these changes, nor relaxation in any applicability, action, or 
surveillance requirements.

--Overall plant performance and operation [are] not altered by the 
proposed changes. There are to be no plant hardware changes as a 
result of this proposed change and only minimal procedural changes.

    Therefore, since the Engineered Safety Feature Actuation System 
Instrumentation [is] treated more conservatively, the probability of 
occurrence or consequences of an accident evaluated in the VCSNS 
FSAR will be no greater than the original design basis of the plant.
    Therefore, the change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes provide consistency between Tables 3.3-2, 
3.3-3, and 4.3-2, resulting in a one-for-one correlation between the 
functional units in those tables. Additionally, the addition of an 
ACTION STATEMENT and an AOT with conservative requirements are 
intended to assure that the plant is in a safe configuration and can 
meet accident analyses assumptions. These changes are conservative 
and consistent with the Improved Technical Specifications, NUREG-
1431, Rev. 2. No new accident initiator mechanisms are introduced 
since:

--No physical changes to the Engineered Safety Feature Actuation 
System Instrumentation are made.
--No deletions from the Technical Specifications are made.
--No relaxation in any applicability, action, or surveillance 
requirements [is] made.

    Since the safety and design requirements continue to be met and 
the integrity of the reactor coolant system pressure boundary is not 
challenged, no new accident scenarios have been created. Therefore, 
the types of accidents defined in the FSAR continue to represent the 
credible spectrum of events to be analyzed [that] determine safe 
plant operation.
    3. Does this change involve a significant reduction in margin of 
safety?
    The proposed change requires that an instrument channel for an 
Engineered Safety Feature [remains] operable or be restored to 
operability within a reasonable time period, otherwise a controlled 
shutdown is required. This conforms to the safety analysis where the 
plant and its systems, structures and components must be capable of 
performing the safety function while a DBA is occurring, in the 
presence of a worst case single failure.
    This is not a reduction in a margin of safety, since it restores 
the margin that was designed into the plant.
    The proposed changes provide consistency between Tables 3.3-2, 
3.3-3, and 4.3-2, resulting in a one-for-one correlation between the 
functional units in those tables. These changes are conservative and 
consistent with the Standard Technical Specifications, NUREG-0452, 
Rev. 5.
    The proposed changes impose more restrictive operating 
limitations, and their use provides increased assurance that the 
Engineered Safety Feature Actuation System Instrumentation remains 
operable. Since the changes are conservative additions, it is 
concluded that the changes do not involve a significant reduction in 
the margin of safety. This is not a reduction in a margin of safety, 
since it restores the margin that was designed into the plant.
    Pursuant to 10 CFR 50.91, the preceding analyses [provide] a 
determination that the proposed Technical Specifications change 
poses no significant hazard as delineated by 10 CFR 50.92.

    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: Thomas G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: May 8, 2002.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications (TS) Figure 2.1.1-1, ``Reactor Core 
Safety Limits;'' Table 3.3.1-1, ``Reactor Trip System 
Instrumentation;'' and the associated Bases B 2.1.1 and B 3.3.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
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 can be implemented without adverse impact to 
the safety analyses and plant systems. Implementation of the revised 
VEGP OT[Delta]T [Overtemperature Delta Temperature] and OP[Delta]T 
[Overpower Delta temperature] reactor trip setpoints will continue 
to ensure that fuel melt and departure from nucleate boiling (DNB) 
criteria are met. In addition, the setpoint changes will improve 
operating margin to the OT[Delta]T and OP[Delta]T reactor trip 
setpoints. The setpoints provide reactor protection and are not 
event initiators and therefore do not affect the probability of 
occurrence of an accident previously evaluated.
    There is no change in the radiological consequences of any 
accident since the fuel clad, the reactor coolant system pressure 
boundary, and the containment are not changed, nor will the 
integrity of these physical barriers be challenged. In addition, the 
proposed change will not change, degrade, or prevent any reactor 
trip system actuations.

[[Page 48222]]

    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?
    The proposed change can be implemented without adverse impact to 
the safety analyses and plant systems. Implementation of the revised 
VEGP OT[Delta]T and OP[Delta]T reactor trip setpoints will continue 
to ensure that fuel melt and departure from nucleate boiling (DNB) 
criteria are met. In addition, the setpoint changes will improve 
operating margin to the OT[Delta]T and OP[Delta]T reactor trip 
setpoints. The revised OT[Delta]T and OP[Delta]T reactor trip 
setpoints would not create any new transients nor would they 
invalidate the OT[Delta]T and OP[Delta]T design bases.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    The proposed change can be implemented without adverse impact to 
the safety analyses and plant systems. Implementation of the revised 
VEGP OT[Delta]T and OP[Delta]T reactor trip setpoints will continue 
to ensure that fuel melt and departure from nucleate boiling (DNB) 
criteria are met. In addition, the setpoint changes will improve 
operating margin to the OT[Delta]T and OP[Delta]T reactor trip 
setpoints. The margin of safety provided by the Technical 
Specifications is not significantly affected because the proposed 
changes are based on the same accident acceptance limits, i.e., the 
OT[Delta]T and OP[Delta]T design bases continue to be met.

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

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

    Date of application request: June 17, 2002. This application 
supercedes the December 6, 2001, application that was noticed in the 
Federal Register on February 5, 2002 (67 FR 5340).
    Description of amendment request: The proposed amendment would 
revise the following Technical Specifications (TSs): (1) TS 3.3.6, 
``Containment Purge Isolation Instrumentation;'' (2) TS 3.3.7, 
``Control Room Emergency Ventilation System (CREVS) Instrumentation;'' 
(3) TS 3.3.8, ``Emergency Exhaust System (EES) Actuation 
Instrumentation;'' and (4) TS 3.9.4, ``Containment Penetrations.'' The 
revisions to the TSs affect limiting conditions for operation (LCOs), 
the required actions for LCOs, surveillance requirements, and tables 
specifying requirements on instrumentation. The revisions to the TSs 
are to allow the equipment hatch and the emergency air lock to be open 
in refueling outages during core alterations and/or movement of 
irradiated fuel within containment.
    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 changes will allow the containment equipment hatch 
[and the emergency air lock] to be open during CORE ALTERATIONS and 
movement of irradiated fuel assemblies inside containment. The 
status of the containment equipment hatch or the emergency air lock 
during refueling operations has no [e]ffect on the probability of 
the occurrence of any accident previously evaluated. The proposed 
revision does not alter any plant equipment or operating practices 
in such a manner that the probability of an accident is increased. 
Since the consequences of a fuel handling accident inside 
containment with an open containment hatch [or emergency air lock] 
are bounded by the current analysis described in the FSAR [Final 
Safety Analysis Report] and the probability of an accident is not 
affected by the status of the containment equipment hatch [or 
emergency air lock], the proposed change[s] [do] not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not create any new failure modes for any 
system or component, nor do they adversely affect plant operation. 
No new equipment will be added and no new limiting single failures 
will be created. The plant will continue to be operated within the 
envelope of the existing safety analysis.
    Therefore, the proposed changes do not create 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 previously determined radiological dose consequences for a 
fuel handling accident inside containment with the [equipment hatch 
or the] air lock doors open remain bounding for the proposed 
changes. These previously determined dose consequences were 
determined to be well within the limits of 10 CFR 100 and they meet 
the acceptance criteria of SRP [NRC Standard Review Plan] section 
15.7.4 and GDC [NRC General Design Criterion] 19.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

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

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

    Date of application request: June 17, 2002.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3.3.1, ``Reactor Trip System (RTS) 
Instrumentation,'' by adding Surveillance Requirement (SR) 3.3.1.16 to 
Function 3 of TS Table 3.3.1-1. The amendment would add a requirement 
to verify the reactor trip system response times are within limits 
every 18 months on a staggered test basis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. 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 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 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

[[Page 48223]]

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 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. 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. These additional 
requirements are 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 this amendment. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of this amendment.
    This amendment does not alter the design or performance of the 
7300 Process Protection System, Nuclear Instrumentation System, 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.
    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 (F[Delta]H), 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.
    The safety analysis limits assumed in the transient and accident 
analyses are unchanged. None of the acceptance criteria for any 
accident analysis is changed. The imposition of additional 
surveillance requirements increases the margin of safety by assuring 
that the affected safety analysis assumptions on equipment response 
time 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 request involves no significant hazards consideration.
    Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: June 27, 2002. This application revises 
the application of September 27, 2001, that was originally noticed in 
the Federal Register on October 17, 2001 (66 FR 52805).
    Description of amendment request: The proposed amendment would 
revise Section 5.3.1.1 of the Technical Specifications to state new 
education and experience eligibility requirements for operator license 
applicants.
    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 TS change is an administrative change to clarify 
the current requirements for licensed operator qualifications and 
licensed operator training program. [The change conforms] to the 
current requirements of 10 CFR 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 55 rule, concluded that this 
impact remains acceptable as long as the licensed operator training 
program is certified to be accredited and is based on a systems 
approach to training. WCNOC's [Wolf Creek Nuclear Operating 
Corporation's] licensed operator training program is accredited by 
INPO [Institute for Nuclear Power Operations] and is based on a 
systems approach to training. The proposed TS change takes credit 
for the INPO accreditation of the licensed operator training 
program. The TS requirements for all other unit staff qualifications 
remain unchanged.
    Therefore, the proposed change does not involve a signification 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed TS change is an administrative change to clarify 
the current requirements for licensed operator qualifications and 
licensed operator training program and to conform to the revised 10 
CFR 55.
    As noted 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 [10 CFR 55] rule, concluded that this 
impact remains acceptable as long as the licensed operator training 
program is certified to be accredited and based on a systems 
approach to training. As previously noted, WCNOC's licensed operator 
training program is accredited by INPO and is based on a systems 
approach to training. The proposed TS change takes credit for the 
INPO accreditation of the licensed operator training program. The TS 
requirements for all other unit staff qualifications remain 
unchanged.
    Additionally, the proposed TS 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. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed TS change is an administrative change to clarify 
the current requirements applicable to licensed operator 
qualifications and licensed operator training program. This change 
is consistent with the requirements of 10 CFR 55. The TS 
qualification requirements for all other unit 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 55 [rule], 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, WCNOC's licensed operator training 
program is accredited by INPO 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 INPO in their training 
accreditation program are equivalent to those put forth or endorsed 
by the NRC. As a result, maintaining an INPO accredited, systems 
approach based licensed operator training program is equivalent to 
maintaining an NRC approved licensed operator training program which 
conform with applicable NRC Regulatory Guides or NRC endorsed 
industry standards. The margin of safety is maintained by virtue of 
maintaining an INPO accredited licensed operator training program.
    In addition, the NRC has recently published NRC Regulatory Issue 
Summary 2001-01, ``Eligibility of Operator License Applicants,'' 
dated January 18, 2001, ``to familiarize addresses with the NRC's 
current guidelines for the qualification and training

[[Page 48224]]

of reactor operator (RO) and senior operator (SO) license 
applicants.'' This document again acknowledges that the INPO 
National Academy for Nuclear Training (NANT) 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 request involves 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: Stephen Dembek.

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

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

    Date of application for amendment: August 24, 2001, as supplemented 
June 11, 2002.
    Brief description of amendment: The amendment revises the control 
room emergency filtration system requirements in Technical 
Specification 3.7.3, ``Control Room Emergency Filtration (CREF) 
System,'' based on NRC-approved Industry/Technical Specification Task 
Force (TSTF) Standard Technical Specification Traveler TSTF-287, 
Revision 5, ``Ventilation System Envelope Allowed Outage Times.''
    Date of issuance: June 28, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 149.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 28, 2002 (67 FR 
36929). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 28, 2002.
    No significant hazards consideration comments received: No.

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 application for amendments: December 20, 2001.
    Brief description of amendments: The amendments revised the 
Technical Specifications 5.6.5.b to eliminate the revision number and 
dates from the list of topical reports that contain the analytical 
methods used to determine the core operating limits.
    Date of issuance: July 2, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 199 and 192.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 5, 2002 (67 FR 
10010). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 2, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 20, 2001.
    Brief description of amendments: The amendments revised the 
Technical Specification 5.6.5.b to eliminate the revision number and 
dates from the list of topical reports that contain the analytical 
methods used to determine the core operating limits.
    Date of issuance: July 10, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 203 and 184.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2921). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 10, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application of amendments: December 20, 2001.
    Brief description of amendments: The amendments revised the 
Technical Specifications 5.6.5.b to eliminate the revision number and 
dates from the list of topical reports that contain the analytical 
methods used to determine the core operating limits.
    Date of Issuance: July 9, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 326, 326 and 327.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 5, 2002 (67 FR 
10011).

[[Page 48225]]

The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated July 9, 2002.
    No significant hazards consideration comments received: No.

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: February 6, 2002, as supplemented by 
letter dated June 7, 2002.
    Brief description of amendment: The amendment relocates the 
requirements for Main Steam Isolation Valve isolations on certain area 
temperatures from Technical Specification Section 3.3.6.1, ``Primary 
Containment and Drywell Isolation Instrumentation,'' to the Technical 
Requirements Manual.
    Date of issuance: July 11, 2002.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 124.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 19, 2002 (67 FR 
12601). The June 7, 2002, supplemental letter provided clarifying 
information that did not change the scope of the original Federal 
Register notice or the original no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 11, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 31, 1999, as supplemented 
by letters dated June 1, July 14, and October 14, 1999, February 11, 
April 4 and 13, June 30, July 31, September 12 and 13, and October 23, 
2000, May 31, October 18, 2001, and February 6, March 27, April 26, and 
June 11 and 12, 2002 (two letters).
    Brief description of amendment: The amendment provides for the full 
conversion of the Current Technical Specifications to the Improved 
Technical Specifications.
    Date of issuance: July 3, 2002.
    Effective date: As of the date of issuance to be implemented within 
120 days.
    Amendment No.: 274.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 8, 1999, (64 
FR 60584), December 13, 1999, (64 FR 69574) and November 28, 2001 (66 
FR 59595). The letters subsequent to the November 28, 2001, Federal 
Register notice did not change the technical content of the Federal 
Register notices, and did not change the scope of the proposed action. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated July 3, 2002.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant (KNPP), Kewaunee County, Wisconsin

    Date of application for amendment: April 17, 2002.
    Brief description of amendment: The amendment revises the KNPP 
Technical Specification (TS) 6.3, ``Plant Staff Qualifications,'' to 
change the title of the Superintendent Plant Radiation Protection to 
the Radiation Protection Manager. In addition, the licensee informed 
the Nuclear Regulatory Commission staff of its intention to reformat TS 
6.3 using MicroSoft Word format.
    Date of issuance: June 28, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 161.
    Facility Operating License No. DPR-43: Amendment revised the TSs.
    Date of initial notice in Federal Register: May 28, 2002 (67 FR 
36932). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 28, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 7, 2001, as supplemented 
December 14, 2001 and April 1, 2002.
    Brief description of amendment: Revised the Technical 
Specifications (TSs) to add a new condition and associated actions to 
Limiting Condition for Operation 3.8.1, ``AC Sources Operating,'' to 
allow one diesel generator to be out of service for 14 days.
    Date of issuance: July 1, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 39.
    Facility Operating License No. NPF-90: Amendment revised the TSs.
    Date of initial notice in Federal Register: September 19, 2001 (66 
FR 48292). The supplemental letters provided clarifying information 
that was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated July 1, 2002.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 12th of July, 2002.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 02-18242 Filed 7-22-02; 8:45 am]
BILLING CODE 7590-01-P