[Federal Register Volume 67, Number 43 (Tuesday, March 5, 2002)]
[Notices]
[Pages 10006-10022]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-5000]


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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 February 8, 2002 through February 21, 2002. 
The last biweekly notice was published on February 19, 2002 (67 FR 
7410).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period.

[[Page 10007]]

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 NRC'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 April 4, 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, which is 
available at the NRC'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 Systems (ADAMS) Public Electronic Reading Room on the 
internet at the NRC web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or an Atomic 
Safety and Licensing Board, designated by the Commission or by the 
Chairman of the Atomic Safety and Licensing Board Panel, will rule on 
the request and/or petition; and the Secretary or the designated Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland, by the above date. A copy of the petition should 
also be sent to the Office of the General Counsel, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and 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 
Systems (ADAMS) Public

[[Page 10008]]

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].

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

    Date of amendment request: September 10, 2001.
    Description of amendment request: The proposed amendment would 
revise the requirements in Technical Specifications (TSs), Sections 
3.4.A.7.c and 3.4.A.8.c, to determine operability of core spray pumps 
and system components by verification rather than testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed TS changes are not associated with accident 
initiators. The proposed changes are, however, associated with 
emergency core cooling requirements for loss of coolant mitigation. 
This event is a loss of coolant from the reactor vessel when the 
plant is shutdown and was evaluated in the NRC [Nuclear Regulatory 
Commission] Safety Evaluation Report supporting License Amendment 
No. 12, dated January 21, 1976. The proposed changes contained in 
this request do not affect the assumptions or conclusions of that 
evaluation and do not impact the physical characteristics of the 
core spray and fire protection systems. Therefore, the proposed 
changes do not degrade the ability of the core spray and fire 
protection systems to perform their intended accident mitigation 
function. The proposed changes to core spray pump/component and fire 
protection system operability verification versus demonstration in 
TS 3.4.A.7.c and core spray pump/component operability verification 
versus demonstration in TS 3.4.A.8.c provide an alternate means of 
determining equipment operability without reliance on frequent 
testing. The clarification of the extent of core spray system 
operability verification in TS 3.4.A.7.c does not change any 
existing requirements. Therefore, the proposed changes to TS 
3.4.A.7.c and 3.4.A.8.c do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS changes are not associated with accident 
initiators. They are changes that provide an alternate means of 
determining equipment operability while eliminating frequent 
testing.
    The proposed changes to TS 3.4.A.7.c and 3.4.A.8.c do not 
involve the addition of any new plant structure, system or component 
(SSC). Similarly, the proposed TS changes do not involve physical 
changes to an existing SSC nor do they modify any current operating 
parameters. Providing an alternate means of determining equipment 
operability does not alter the functional capability of any accident 
mitigation system. The clarification of the extent of core spray 
system operability verification in TS 3.4.A.7.c does not change any 
existing requirements. Therefore, the proposed changes do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed changes to TS 3.4.A.7.c and 3.4.A.8.c are not 
associated with accident initiators and do not introduce new SSCs or 
physically impact existing SSCs. They are changes that provide an 
alternate means (i.e., verification) of determining core spray and 
fire protection system component operability. The capability of the 
necessary core spray and fire protection components to provide the 
required core cooling flow is demonstrated during surveillance 
testing. While the proposed changes revise the method of determining 
the operability of the core spray and fire protection system in the 
reduced availability mode, they do not degrade the ability of the 
systems to perform their intended function. The clarification of the 
extent of core spray system operability verification in TS 3.4.A.7.c 
does not change any existing requirements. Therefore, the proposed 
TS 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: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Joel Munday, Acting.

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

    Date of amendment request: September 11, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications, Section 3.9, ``Refueling,'' to 
incorporate compensatory provisions which permit fuel-handling 
operations without the refueling interlocks operable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff reviewed the licensee's analysis and has 
performed its own, which is presented below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    No. The proposed amendment involves refueling interlock operability 
requirements during refueling operations. The only design-basis 
accident described in the Oyster Creek Updated Final Safety Analysis 
Report (UFSAR) for cold shutdown or refueling conditions is a 
postulated fuel handling (dropped bundle) accident. The refueling 
interlocks are not postulated to cause, and are not involved in the 
mitigation of such an accident. Thus, the proposed amendment does not 
affect the safety function of the refueling interlocks. The proposed 
alternative actions provide an equivalent level of protection against 
inadvertent criticality during fuel handling operations. Therefore, 
this amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the amendment create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed amendment does not affect accident initiators or 
precursors because it does not alter any design parameter, condition, 
equipment configuration, or manner in which the unit is operated. 
Further, it does not alter or prevent the ability of structures, 
systems, or components to perform their intended safety or accident 
mitigating functions. Accordingly, the proposed amendment does not 
create a new or different kind of accident from any accident previously 
evaluated.
    3. Does the amendment involve a significant reduction in a margin 
of safety?
    No. The proposed amendment does not change any design parameter, 
analysis methodology, safety limits or acceptance criteria. The revised 
requirement (i.e., proposed alternative) will continue to ensure 
against inadvertent criticality during fuel handling operations. 
Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.
    Based on the NRC staff's review, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the

[[Page 10009]]

NRC staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Joel Munday, Acting.

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

    Date of amendment request: August 14, 2001.
    Description of amendment request: The proposed amendment revises 
Section 6, ``Administrative Controls,'' of the Technical Specifications 
(TSs) to delete Section 6.5.4, ``Independent Onsite Safety Review 
Group,'' and all associated subsections. The licensee will revise its 
Operational Quality Assurance Plan to incorporate conforming changes to 
provide its proposed alternative independent nuclear safety oversight 
provisions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    This change involves deletion of the TS requirements for the 
Independent Onsite Safety Review Group [IOSRG]. To satisfy the 
NUREG-0737 [``Clarification of TMI Action Plan Requirements,'' 
November 1980] guidance concerning organizational independence, the 
proposed IOSRG alternative provides for technical expertise by 
onsite engineering and licensing organizations. These site 
engineering and licensing organizations report through the Site 
Vice-President and are independent of the production reporting chain 
through the plant manager. Additionally, high-level management 
positions are located in the corporate and regional offices for 
these engineering and licensing organizations which set policy and 
have responsibility for governance and oversight of these functional 
areas. These corporate and regional high-level positions are not in 
the management chain for power production.
    Organizational and procedural changes at TMI Unit 1 [Three Mile 
Island Nuclear Station, Unit 1] following the issuance of NUREG-0737 
have resulted in improvements to the review processes that meet the 
intent of the requirements [of] NUREG-0737 for an IOSRG. Therefore, 
inclusion of the IOSRG in the plant or plant support organization is 
unneccessary. In light of the considerable improvement in the 
processes listed above, the contribution of three full time 
engineers assigned as a separate group to address nuclear safety 
oversight is not significant in comparison to the contribution of 
the overall organization. This change does not affect assumptions 
contained in the plant safety analyses, the physical design and/or 
operation of the plant, nor does it affect Technical Specifications 
that preserve safety analysis assumptions. No Technical 
Specification Limiting Condition for Operation, Action Statement, or 
Surveillance Requirement is affected by this change. The proposed 
change does not alter design, function, operation, or reliability of 
any plant component. This change does not involve a physical 
modification to the plant, a mode of operation, or a change to the 
UFSAR [Updated Final Safety Analysis Report] transient analyses. 
Normal and accident dose to plant personnel or to the public are 
unaffected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    This change to remove the IOSRG from the TS[s] is administrative 
in nature and does not affect the assumptions contained in the plant 
safety analyses, the physical design and/or modes of plant operation 
defined in the plant operating license that preserve safety analysis 
assumptions.
    This proposed change does not introduce a new mode of plant 
operation or surveillance requirement, nor involve a physical 
modification to the plant. The proposed change does not alter the 
design, function, or operation of any plant system or component.
    Therefore, the change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    This change only involves Technical Specification Section 6, 
``Administrative Controls,'' which does not include any margins of 
safety. None of the proposed changes involve a physical modification 
to the plant, a new mode of operation, an instrument setpoint, or a 
change to the UFSAR transient analyses. No Limiting Safety System 
Setting, Technical Specification Limiting Condition for Operation, 
Action Statement, or Surveillance Requirement is affected. 
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: Edward J. Cullen, Jr., Esquire, Vice 
President, General Counsel and Secretary, Exelon Generation Company, 
LLC, 300 Exelon Way, Kennett Square, PA 19348.
    NRC Section Chief: Joel T. Munday (Acting).

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck 
Plant, Middlesex County, Connecticut

    Date of amendment request: September 10, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3/4.9.7 and corresponding Bases to 
address use of a single-failure-proof handling system, as defined by 
NUREG-0612 (``Control of Heavy Loads at Nuclear Power Plants'') and 
NUREG-0554 (``Single-Failure-Proof Cranes For Nuclear Power Plants''). 
The modifications will allow handling loads in excess of 1,800 pounds 
near or over the Spent Fuel Pool. The anticipated types of heavy loads 
include the combination of a spent fuel storage canister and transfer 
cask.
    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.
    Concerning the application of a single-failure-proof handling 
system for handling heavy loads near or over the Spent Fuel Pool, 
NUREG-0612, ``Control of Heavy Loads at Nuclear Power Plants'' 
asserts that the probability of an accidental load drop while 
handling loads over the spent fuel is insignificant.
    Under the proposed amendment, the evaluation criteria of NUREG-
0612, Section 5.1 are satisfied by the combination of (a) the 
continued implementation of procedures and the practices for both 
the Fuel Handling Cranes and the Yard Crane that provide conformance 
with the guidelines of Section 5.1.1 of NUREG-0612, and (b) the 
application of a single-failure-proof handling system that satisfies 
the criteria of NUREG-0612, Sections 5.1.2(1) and 5.1.6 for the 
movement of any load with a weight greater than 1800 pounds either 
(i) over any spent fuel assembly in the Spent Fuel Pool or (ii) near 
or over any area of the Spent Fuel Pool, including the Spent Fuel 
Cask Laydown Area.
    The proposed amendment retains existing restrictions on crane 
travel for the Fuel Handling Cranes, which are not qualified to the 
single-failure-proof criteria of NUREG-0612. These retained 
restrictions continue to support the existing safety analysis of 
Section 15.2.2, ``Fuel Handling Accident'' of the

[[Page 10010]]

UFSAR [Updated Final Safety Analysis Report], Reference (9) [Haddam 
Neck Plant UFSAR Change 34 dated August 2, 2000].
    Additionally, the proposed amendment corresponds to the 
application of a single-failure-proof handling system to fulfill the 
NUREG-0612 Phase II condition that is required prior to the handling 
of a spent fuel cask near or over any area of the Spent Fuel Pool.
    Therefore, the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes will allow the handling by a single-
failure-proof handling system of loads in excess of 1800 pounds over 
fuel assemblies in any region of the Spent Fuel Pool, including the 
Spent Fuel Cask Laydown Area.
    Additionally, the proposed changes correspond to the application 
of a single-failure-proof handling system for the fulfillment of the 
required condition for the handling of spent fuel casks near or over 
any area of the Spent Fuel Pool. This required condition is 
identified in the documentation for the NRC Issuance of License 
Amendment 125, Ref. (7) [Letter from US NRC to CYAPCO, dated April 
26, 1990] and it is acknowledged in the CYAPCO submittal for the 
proposed license amendment that was issued as License Amendment 188, 
Ref. (5) [Letter from J. F. Opeka (CYAPCO) to US NRC, ``Haddam Neck 
Plant Proposed Revision to Technical Specifications Spent Fuel Pool 
Capacity Expansion,'' Letter Number B15136, dated March 31, 1995.] 
and the NRC Issuance of License Amendment 195, Ref. (6) [Letter from 
T. L. Fredrichs (NRC) to R. A. Mellor (CYAPCO), ``Haddam Neck Plant-
Issuance of Amendment RE: Relocation of Requirements to Licensee--
Controlled Documents (TAC No. MA5756),'' dated October 19, 1999].
    NUREG-0612, Section 5.1.2 identifies that the capability of a 
single-failure-proof handling system to handle heavy loads has been 
identified as equivalent in risk to the capabilities of a non-
single-failure-proof heavy load handling system that complies with 
the criteria of one of the other three alternative sets from NUREG-
0612 (including alternative criteria that include analyses 
concerning postulated heavy load drops).
    A structural evaluation of the heavy load interfaces within the 
Spent Fuel Cask Laydown Area and the Cask Transfer Bay was performed 
per the requirements of EDR-1 [(Reference 2) Generic Licensing 
Topical Report EDR-I (P)-A, ``EDERER's Nuclear Safety Related eXtra 
Safety And Monitoring (X-SAM) CRANES,'' Revision 3, Amendment 3, 
dated October 8, 1982] Appendix B and C (Attachments 2 and 3 
[attachments to this application]). The results of the evaluation 
confirmed the design bases for the Spent Fuel Pool and the Spent 
Fuel Building are maintained.
    As such, use of a single-failure-proof handling system precludes 
the possibility of a heavy load drop which could cause an accident 
outside of the existing design bases.
    Additionally, the proposed changes retain existing restrictions 
on the travel of non-single-failure-proof cranes over fuel 
assemblies in the Spent Fuel Pool. These retained restrictions 
continue to support the existing safety analysis of Section 15.2.2, 
``Fuel Handling Accident'' of the UFSAR, Reference (9).
    Therefore, operation of the facility in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Section 5.1.2 of NUREG-0612 identifies that each of the four 
alternative sets of criteria for the handling of heavy loads near or 
over the Spent Fuel Pool, including over fuel assemblies, provides a 
level of safety that is essentially equivalent to the level of 
safety provided by any of the other three alternative sets of 
criteria.
    The proposed change corresponds to the application of the first 
of the four alternative sets of criteria, which is described in 
NUREG-0612 Section 5.1.2(1), implementation of a single-failure-
proof handling system.
    Additionally, the proposed change includes the retention of 
existing crane travel restrictions for the Fuel Handling Cranes, 
therefore, maintaining the existing margin of safety concerning the 
operation of those other cranes.
    Therefore, operation of the facility in accordance with the 
proposed amendment will not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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. Robert K. Gad, III, Ropes & Gray, One 
International Plaza, Boston, Massachusetts 02110-2624.
    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: December 20, 2001.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TS) to eliminate the use of the term 
``unreviewed safety question.'' The change is proposed by the licensee 
to reflect changes in the NRC's regulations in 10 CFR 50.59 as noticed 
in the Federal Register on October 4, 1999. The proposed changes in the 
license amendment request are consistent with an NRC approved Technical 
Specifications Task Force Standard TS Traveler (TSTF-364).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would implementation of the changes proposed in this LAR 
[license amendment request] involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. This LAR makes an administrative change to the Technical 
Specifications [TS] made necessary as part of Duke's implementation 
of revised NRC regulations. The changes proposed to these TS have no 
substantive impact on the Catawba licensing bases, nor Duke's 
ability to conservatively evaluate changes to these licensing bases. 
Therefore, the proposed changes have no impact on any accident 
probabilities or consequences.
    2. Would implementation of the changes proposed in this LAR 
create the possibility of a new or different kind of accident from 
any accident previously evaluated?
    No. This LAR makes administrative changes that have no impact on 
any accident analyses.
    3. Would implementation of the changes proposed in this LAR 
involve a significant reduction in a margin of safety?
    No. The proposed changes are administrative, an implementation 
of the revised 10CFR50.59 regulation. Implementation of the revised 
10CFR50.59 regulation provides the necessary regulatory requirements 
to ensure that nuclear plants' margin of safety is preserved.

    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: Richard J. Laufer, Acting.

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: December 20, 2001.
    Description of amendment request: The amendments would revise 
Technical Specification 5.6.5.b to eliminate the revision number and 
dates of the topical reports that contain the analytical methods used 
to determine the core operating limits. This proposed change is 
consistent with TSTF (Technical Specification Task Force)-363.

[[Page 10011]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would implementation of the changes proposed in this LAR 
[license amendment request] involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. This LAR makes an administrative change to TS 5.6.5.b, Core 
Operating Limits Report (COLR), affecting a list of documents that 
are separately reviewed and approved by the NRC. The changes 
proposed to TS 5.6.5.b have no substantive impact on the Catawba 
licensing bases. Only NRC-approved methodologies will be used to 
generate the core operating limits. Based on these considerations, 
it has been determined that the proposed changes have no impact on 
any accident probabilities or consequences.
    2. Would implementation of the changes proposed in this LAR 
create the possibility of a new or different kind of accident from 
any accident previously evaluated?
    No. This LAR makes administrative changes that have no impact on 
any accident analyses.
    3. Would implementation of the changes proposed in this LAR 
involve a significant reduction in a margin of safety?
    No. The analytical methodologies used to generate the core 
operating limits are unchanged by this LAR. As such, this LAR has no 
affect on margins of safety. Future changes to these methodologies 
will remain subject to NRC review and approval. Therefore, this 
proposed amendment does not involve a reduction in any margin of 
safety.

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

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: December 20, 2001.
    Description of amendment request: The proposed amendments would 
revise the Technical Specification 5.6.5.b to eliminate the revision 
number and dates of the topical reports that contain the analytical 
methods used to determine the core operating limits. This proposed 
change is consistent with TSTF (Technical Specification Task Force)-
363. This notice supersedes in its entirety the previous notice issued 
on February 5, 2002 (67 FR 5326) for the Oconee December 20, 2001, 
application, which contained the incorrect licensee analysis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would implementation of the changes proposed in this LAR 
[license amendment request] involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. This LAR makes an administrative change to TS 5.6.5.b, Core 
Operating Limits Report (COLR), affecting a list of documents that 
are separately reviewed and approved by the NRC. The changes 
proposed to TS 5.6.5.b have no substantive impact on the Oconee 
licensing bases. Only NRC-approved methodologies will be used to 
generate the core operating limits. Based on these considerations, 
it has been determined that the proposed changes have no impact on 
any accident probabilities or consequences.
    2. Would implementation of the changes proposed in this LAR 
create the possibility of a new or different kind of accident from 
any accident previously evaluated?
    No. This LAR makes administrative changes that have no impact on 
any accident analyses.
    3. Would implementation of the changes proposed in this LAR 
involve a significant reduction in a margin of safety?
    No. The analytical methodologies used to generate the core 
operating limits are unchanged by this LAR. As such, this LAR has no 
affect on margins of safety. Future changes to these methodologies 
will remain subject to NRC review and approval. Therefore, this 
proposed amendment does not involve a reduction in any 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: Richard J. Laufer, Acting.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York

    Date of amendment request: January 8, 2002.
    Description of amendment request: The amendment would delete the 
Technical Specification (TS) requirements governing the reactor vessel 
material surveillance program; would change the TS Sections 4.2, 
``Inservice Inspection and Testing,'' 5.2.C, ``Design Features--
Containment,'' and 6.4, ``Administrative Controls--Training,'' to 
correct errors; and would change TS Section 6.1, ``Responsibility,'' 
and 6.2, ``Organization,'' to reflect the organizational changes 
resulting from the license transfer to Entergy Nuclear Operations, Inc. 
(ENO).
    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 license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    The proposed change to TS Section 3.1.B involves deleting 
specific TS requirements that duplicate the requirements of 10 CFR 
[Code of Federal Regulations] 50.60, 10CFR50 Appendix G, and 10CFR50 
Appendix H. The proposed change does not result in a change to the 
design or operation of any plant structure, system or component. 
Therefore any assumptions of the operability or performance of any 
structure, system or component in accident evaluations are 
unchanged.
    The proposed change to TS 4.2.1 simply corrects an improper 
reference to the CFR. There are no physical changes to IP2 or to the 
operation of any system, structure, or component.
    The proposed change to TS 5.2.C makes the design feature 
description consistent with TS Limiting Condition for Operation 
3.3.B wherein the requirements for the method of post-accident 
iodine removal are specified. Making the Design Feature consistent 
with the appropriate LCO has no effect on the assumptions and the 
results of the accident analyses.
    TS sections 6.1, 6.2, and 6.4 are administrative controls. 
Changing an administrative control has no affect on accident 
analyses.
    Therefore, there will be no increase in the probability or in 
the consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    The proposed change to TS Section 3.1.B does not affect the 
effectiveness of ENO's implementation of the requirements of 
10CFR50.60 that ensure the reactor vessel continues to be protected 
against non-ductile failure.
    There is no change to any system, structure, or component as a 
result of any of the proposed changes.

[[Page 10012]]

    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The proposed TS changes simplify the methods of controlling the 
schedule for the reactor vessel surveillance specimen withdrawal 
schedule in that a duplicative control is removed. The effectiveness 
of ENO compliance with 10CFR50.60 and 10CFR50 Appendices G and 
Appendix H is not adversely affected by this change. The level of 
regulatory control for the reactor vessel pressure/temperature 
limits is not changed.
    The effectiveness of IP2's [Indian Point 2] inservice testing 
program is not affected by the correction of the improper CFR 
reference in TS 4.2.1. ENO is required to comply with 10CFR55 at 
IP2. The effectiveness of ENO's compliance with 10CFR55 is not 
affected by deleting the improper CFR citation from TS 6.4. 
Similarly, ENO's compliance with the IP2 license and the all 
applicable laws and regulations is not affected by the proposed 
changes to the TS sections for responsibility and organization.
    The change to the Design Features to properly identify the 
method specified in TS 5.2.B for post-accident iodine removal does 
not affect the margin of safety.
    This change does not affect any design function for or the 
operation of any plant structure, system, or component.
    Therefore, the change [* * *] does not result in a change to any 
of the safety analyses or [a] margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Joel Munday, Acting.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York

    Date of amendment request: January 8, 2002.
    Description of amendment request: The amendment would incorporate 
the use of a more conservative equation to calculate the power range 
high neutron trip setpoint when one or more main steam safety valves 
are inoperable during four loop operation--Technical Specification (TS) 
3.4, ``Steam and Power Conversion System.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated.
    The proposed change to the setpoints will cause a reactor trip 
on high neutron flux for a decreased heat removal event at an 
earlier (more conservative) condition. The consequences of an 
accident with the proposed setpoints are less severe than those 
predicted with the use of the current setpoints.
    The main steam line code safety valves, in conjunction with the 
high neutron flux reactor trip mitigate the consequences of 
decreased heat removal and uncontrolled rod cluster assembly bank 
withdrawal events. The systems acting together do not initiate or 
cause any accident. Therefore, the probability of analyzed accidents 
is unchanged by the proposed TS change.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    There is no change to either the design or operation of the main 
steam line code safety valves. This proposed change only changes the 
high neutron flux trip setpoints in response to the inoperable main 
steam line code safety valves. This feature currently exists both in 
the plant and in the TS.
    Therefore, the proposed change does not create a new accident 
initiator or precursor, or create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in [a] margin of 
safety.
    The current TS setpoints have been determined to be non-
conservative and insufficient to guarantee safety. The proposed 
change would impose limits that were anticipated in the original TS 
and are conservative with respect to the current TS. Therefore, the 
margin of safety as defined in the TS (protection of the secondary 
system from overpressurization so that it is available for decay 
heat removal) will be restored to that intended with the original 
TS.
    The ability to keep the core cooled in spite of the 
inoperability of some main steam line code safety valves is enhanced 
by the proposed change. Therefore, operation of the facility in 
accordance with the proposed amendment would not involve a 
significant reduction in [a] margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Joel Munday, Acting.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York

    Date of amendment request: January 8, 2002.
    Description of amendment request: The amendment would revise the 
Technical Specification (TS) Section 3.7.C, ``Gas Turbine Generators,'' 
and Section 4.6, ``Emergency Power System Periodic Tests,'' by changing 
the requirement to maintain a minimum amount of fuel oil stored on site 
from 54,200 gallons to 94,870 gallons.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated.
    The Gas Turbine Generators only provide a Licensing Basis Event 
mitigating function. There is no previously evaluated accident or 
event that is initiated by the Gas Turbine Generators or their 
associated fuel storage system. The ability of the Gas Turbine 
Generators to provide power, as a backup to the Emergency Diesel 
Generators, is enhanced by the proposed change to increase the 
amount of fuel stored on site and dedicated to Gas Turbine Generator 
operation. The increase in minimum load has an insignificant affect 
because the Gas Turbine Generators are capable of loads far in 
excess of the proposed minimum load.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    There is no physical change to the plant. The currently existing 
fuel oil storage facilities will be used. The only change is to 
increase the minimum amount of fuel oil that must be maintained at 
the plant.
    Therefore, the proposed change does not create a new accident 
initiator or precursor, or create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not

[[Page 10013]]

involve a significant reduction in [a] margin of safety.
    The proposed limit for Gas Turbine Generator fuel oil storage 
ensures compliance with the current licensing basis that the Gas 
Turbine Generators be able to power all the loads required by 10 CFR 
[Code of Federal Regulations] 50 Appendix R to place the plant into 
a safe shutdown condition following a fire and maintain safe 
shutdown for three days. The increase in the minimum load rating 
ensures that each Gas Turbine Generator will support operation of 
additional components to enhance operational flexibility in response 
to an event.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant reduction in [a] 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Joel Munday, Acting.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York

    Date of amendment request: January 8, 2002.
    Description of amendment request: The amendment would delete the 
Technical Specification (TS) requirements governing the Fuel Storage 
Building Air Filtration System. The proposed changes affect TS 3.8, 
``Refueling, Fuel Storage and Operations with the Reactor Vessel Head 
Bolts Less Than Fully Tensioned,'' and TS 4.5.F, ``Fuel Storage 
Building Air Filtration System.''
    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 license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    The fuel storage building air filtration system is not involved 
in the initiation of any accident nor does it function to prevent 
any accident. The fuel storage building air filtration system was an 
accident mitigating system. Therefore there is no affect on the 
probability of occurrence of a fuel handling accident in the fuel 
storage building.
    The fuel storage building air filtration system was designed to 
provide an accident mitigation function by filtering the 
radionuclides that might have been released from a damaged fuel 
assembly in the event of a fuel handling accident. The charcoal 
adsorber was the primary component that supported this filtration 
function. However, based on the recent IP2 [Indian Point 2] analyses 
to show compliance with 10CFR [Code of Federal Regulations] 50.67, 
it has been shown that the doses to the public and to control room 
operators due to a fuel handling accident remain well within 
regulatory limits even assuming no credit for either isolation or 
filtration. Therefore, the charcoal filtration function is not 
required in the event of a fuel handling accident.
    There would be no change to the radiological consequences of the 
fuel handling accident in the fuel storage building analysis as a 
result of the proposed change. The proposed changes ensure that the 
assumptions of the fuel handling accident analysis for the release 
of radioactivity from a damaged fuel assembly in the fuel storage 
building are maintained.
    Therefore, there will be no increase in the probability or in 
the consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    The fuel storage building air filtration system is not an 
accident initiator. It was designed as an accident mitigation system 
to filter the radionuclides that may be released from a damaged fuel 
assembly during a fuel handling accident. The fuel storage building 
air filtration system does not affect any accident initiator.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The margin of safety is defined by 10CFR50.67 and 10CFR50 
Appendix A Criterion 19. The radiological consequences of a fuel 
handling accident in the fuel storage building have been shown to be 
well within the regulatory requirements even when assuming no credit 
for the fuel storage building air filtration system operation.
    The proposed change ensures that the assumptions of the current 
fuel handling analysis for the release of radioactivity from a 
damaged fuel assembly are maintained.
    Therefore, the change does not result in a change to any of the 
safety analyses or [a] margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear. Operations, Inc., 440 Hamilton Avenue, White Plains, 
NY 10601.
    NRC Section Chief: Joel Munday, Acting.

Exelon Generation Company, LLC, Docket No. 50-237, Dresden Nuclear 
Power Station, Unit 2, Grundy County, Illinois

    Date of amendment request: September 5, 2001.
    Description of amendment request: The proposed amendments would 
revise the battery terminal voltage on float charge for the alternate 
battery.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. The 
proposed change is to a SR 3.8.4.1 acceptance criterion that will 
continue to ensure equipment operability. By continuing to ensure 
equipment operability, the probability or consequences of an 
accident previously evaluated are not increased. Float charge is the 
condition in which the charger is supplying the continuous charge 
required to overcome the internal losses of a battery and maintain 
the battery in a fully charged state. The voltage requirements are 
based on the nominal design voltage of the battery and are 
consistent with the initial voltages assumed in the battery sizing 
calculations. The 125 VDC alternate battery continues to provide 
reliable DC power for operation of the required equipment. The 
number of cells in the alternate battery was increased from sixty to 
sixty-three and the acceptable float voltage needed to be revised to 
reflect the additional cells. The addition of the three cells has 
been evaluated and documented in calculations. These calculations 
demonstrate that the batteries are appropriately sized to supply the 
required loads following a loss of offsite power. The ability of the 
battery to perform its intended function remains unchanged. In 
addition, the proposed change has no impact on any initial condition 
assumptions for accident scenarios. Onsite or offsite dose 
consequences resulting from an accident previously evaluated are not 
affected by this proposed amendment request.
    Accordingly, there is no significant change 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 license amendment provides a change in a TS 
Surveillance Requirement that continues to ensure equipment 
operability. The increase in terminal voltage specifically supports 
the increase in the number of cells for the battery. The operation 
of the safety-related equipment and components remains unchanged. As 
such, the relationship between the 125 VDC power

[[Page 10014]]

system and plant transient response is maintained. The change in the 
acceptance criterion ensures that the equipment remains operable.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change does not involve a significant reduction in 
the margin of safety. The proposed change continues to ensure 
equipment operability. The increase in terminal voltage specifically 
supports the increase in the number of cells for the alternate 
battery. Since the change maintains the necessary level of system 
reliability, it does not involve a significant reduction in the 
margin of safety. The change in acceptance criterion is to reflect 
the increase in battery cells from sixty to sixty-three. This 
acceptance criterion ensures that the equipment remains operable.

    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, Vice President, 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: January 10, 2002.
    Description of amendment request: The proposed amendment would 
revise Surveillance Requirement (SR) 3.0.3 to extend the delay period, 
before entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period 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 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 January 10, 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 proposes to determined that the amendment requests 
involve no significant hazards consideration.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: December 10, 2001.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TSs) to incorporate the Nuclear 
Regulatory Commission (NRC)-approved generic change TSTF-287, Revision 
5, to the ``Standard Technical Specifications for General Electric 
Plants (BWR/4),'' NUREG-1433, Revision 1. Specifically, the proposed 
changes would: (a) insert a note in the Limiting Condition for 
Operation (LCO) in TS 3.7.3 to state that the control room habitability 
envelope boundary may be opened intermittently under administrative 
control; (b) insert a new LCO Action B in TS 3.7.3 to allow 24 hours to 
restore the control room habitability envelope boundary to

[[Page 10015]]

operable status if two control room emergency outside air supply 
(CREOAS) subsystems should become inoperable due to an inoperable 
control room habitability envelope boundary in Modes 1, 2 and 3; (c) 
re-label the existing LCO Actions b, c, d, and e to c, d, e, and f 
respectively; and (d) revise the existing LCO Action D to require 
immediate entry into LCO 3.0.3 when two CREOAS subsystems are 
inoperable for situations other than when the inoperability is due to 
an inoperable control room habitability envelope boundary. Minor 
formatting and editorial changes are also made.
    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?
    The proposed change relaxes the Required Actions of LCO 3.7.3 by 
allowing 24 hours to restore an inoperable control room habitability 
envelope pressure boundary to OPERABLE status. Required Actions and 
their associated Completion Times are not initiating events for any 
accidents previously evaluated. The accident analyses do not assume 
that required CREOAS equipment is out of service prior to the 
analyzed event. Consequently, this change in Required Actions does 
not significantly increase the probability of occurrence of any 
accident previously evaluated. The Required Actions in the proposed 
change have been developed to provide assurance that appropriate 
remedial actions are taken in response to the degraded condition, 
considering the operability status of the CREOAS system and the 
capability of minimizing the risk associated with continued 
operation. As a result, the consequences of any accident previously 
evaluated are not significantly increased. Therefore, this 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 does not involve a physical modification or 
alteration of plant equipment (no new or different type of equipment 
will be installed) or a change in the methods of governing normal 
plant operation. The Required Actions and associated Completion 
Times in the proposed change have been evaluated to ensure that no 
new accident initiators are introduced. Thus, this 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 a 
margin of safety?
    The relaxed Required Actions do not involve a significant 
reduction in the margin of safety. The proposed change has been 
evaluated to minimize the risk of continued operation with the 
control room habitability envelope pressure boundary inoperable. The 
operability status of the CREOAS system, a reasonable time for 
repairs or replacement of required features, and the low probability 
of a design basis accident occurring during the repair period have 
been considered in the evaluation. Therefore, this 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Joel T. Munday, Acting.

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

    Date of amendment request: December 14, 2001.
    Description of amendment request: A change is proposed to 
Surveillance Requirement (SR) 3.0.3 to allow a longer period of time to 
perform a missed surveillance. The time is extended from the current 
limit of ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is less'' to ``* * * up to 24 hours or up to the 
limit of the specified Frequency, whichever is greater.'' In addition, 
the following requirement 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 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 December 14, 2001.
    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

[[Page 10016]]

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 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: Richard J. Laufer, Acting.

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

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

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

    Date of amendment request: January 31, 2002.
    Brief description of amendment request: The proposed amendment 
would revise the technical specifications by replacing the peak linear 
heat rate safety limit with a peak fuel centerline temperature safety 
limit.
    Date of publication of individual notice in Federal Register: 
February 11, 2002 (67 FR 6279).
    Expiration date of individual notice: The comment period expires on 
February 25, 2002, and the hearing period expires on March 13, 2002.

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

    Date of amendment request: January 31, 2002.
    Brief description of amendment request: The proposed amendment 
would replace the Technical Specification (TS) Safety Limit 2.1.1.2, 
``Peak Linear Heat Rate,'' (PLHR) with a Peak Fuel Centerline 
Temperature Safety Limit and update the Index accordingly. The 
associated TS Bases changes are also made to appropriately reflect the 
proposed new Safety Limit.
    Date of publication of individual notice in Federal Register: 
February 11, 2002 (67 FR 6281).
    Expiration date of individual notice: The comment period expires on 
February 25, 2002, and the hearing period expires on March 13, 2002.

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].

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

    Date of application for amendment: August 21, 2001, as supplemented 
January 11, 2002.
    Brief description of amendment: The amendment revises the actions 
taken for an inoperable battery charger, revises the battery charger 
testing criteria, and relocates certain safety-related battery 
surveillance requirements from the Technical Specifications to a 
licensee-controlled program.
    Date of issuance: February 15, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 142.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 14, 2001 (66 
FR 57118). The letter of January 11, 2002, provided clarification and 
did not affect the NRC staff's proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated February 15, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 19, 2000, as 
supplemented on September 24, 2001.
    Brief description of amendment: The amendment revises the Oyster 
Creek Technical Specifications, Section 3.17, to remove reference to 
the current licensing basis control room calculated dose consequences 
and substitute the associated regulatory dose limits that apply for 
control room habitability in

[[Page 10017]]

accordance with General Design Criterion (GDC) 19 of 10 CFR part 50 and 
Standard Review Plan Section 6.4. Concurrent with this requested 
change, AmerGen Energy Company (AmerGen) recalculated control room 
relative concentration (X/Q) values using the ARCON96 methodology to 
demonstrate its capability to meet GDC-19 dose requirements. The NRC 
staff finds acceptable the use of the diffuse source X/Q values 
calculated by AmerGen because they appear to be more limiting than 
assuming a point source release through a building penetration.
    Date of Issuance: February 7, 2002.
    Effective date: February 7, 2002 and shall be implemented within 30 
days of issuance.
    Amendment No.: 225.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44163). The September 24, 2001, letter provided clarifying information 
within the scope of the original application and did not change the 
staff's initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated February 7, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 20, 2000, as 
supplemented August 2 and September 28, 2001.
    Brief description of amendment: The Technical Specification (TS) 
changes deleted the specification for hydrogen monitoring 
instrumentation from TS Sections 3.5.5.2, 3.6, and Tables 3.5-3 and 
4.1-4, corrected a typographical error in Item 8 of Table 4.1-4, 
deleted the specifications for hydrogen recombiners in TS Section 
4.4.4, and changed the Bases for TS Section 4.12.2 to delete its 
reference to hydrogen purge and hydrogen recombiners.
    Date of issuance: February 8, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days including the designation of hydrogen monitoring 
instrumentation as Category 3 variables as defined in Regulatory Guide 
1.97.
    Amendment No.: 240.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 14, 2001 (66 
FR 57118). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 8, 2002.
    No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-318, Calvert 
Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

    Date of application for amendment: November 19, 2001.
    Brief description of amendment: The amendment increases the allowed 
outage time of one train of the control room emergency ventilation 
system from 10 to 14 days (for the loss of the emergency power supply 
only). This is a one-time change to support corrective maintenance and 
inspections of the 1A Diesel Generator during the Unit 1 refueling 
outage.
    Date of issuance: February 13, 2002.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 223.
    Renewed License No. DPR-69: Amendment revised the Technical 
Specifications.
    Date of initial notice in Federal Register: January 8, 2002 (67 FR 
926). The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 13, 2002.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone 
Nuclear Power Station, Unit No. 3, New London County, Connecticut

    Date of application for amendment: June 29, 2000, as supplemented 
on October 16, 2000, and January 25, April 4, and September 21, 2001.
    Brief description of amendment: The amendment modifies Technical 
Specification (TS) Sections 3.3.2, ``Instrumentation--Engineered Safety 
Features Actuation System Instrumentation;'' 3.7.7, ``Plant Systems--
Control Room Emergency Ventilation System;'' 3.7.8, ``Plant Systems--
Control Room Envelope Pressurization System;'' 3.7.9, ``Plant Systems--
Auxiliary Building Filter System;'' 3.9.1.1, ``Refueling Operations--
Boron Concentration;'' 3.9.1.2, ``Refueling Operations--Boron 
Concentration;'' 3.9.2, ``Refueling Operations--Instrumentation;'' 
3.9.4, ``Refueling Operations--Containment Building Penetrations;'' 
3.9.9, ``Refueling Operations--Containment Purge and Exhaust Isolation 
System;'' 3.9.10, ``Refueling Operations--Water Level--Reactor 
Vessel;'' and 3.9.12, ``Refueling Operations--Fuel Building Exhaust 
Filter System.'' The changes are associated with the revised fuel 
handling accident analyses, and integrity of the Control Room and the 
Fuel Building boundaries. Several administrative changes were also made 
to reflect Millstone Unit 3 terminology, remove unnecessary 
information, and eliminate confusion by providing consistency between 
LCO's (limiting conditions for operations), Action Requirements, and 
Surveillance Requirements. The TS changes are associated with the 
revised containment fuel handling accident analysis which results in an 
increase in the consequences of a containment fuel handling accident 
since the current analysis of a containment fuel handling accident does 
not assume the release of any radioactive material from containment. 
The revised analysis assumes a release of radioactive material because 
it assumes both personnel access hatch doors are open and at least one 
hatch door is closed within 10 minutes of a fuel handling accident 
inside containment.
    Date of issuance: February 20, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 203.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 29, 2000 (65 
FR 71136). The letters dated October 16, 2000, January 25 April 4, and 
September 21, 2001, provided clarifying information and did not change 
the staff's initial proposed no significant hazards consideration 
determination or expand the scope of the application as published in 
the Federal Register. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated February 20, 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: March 22, 2001, as supplemented 
by letter dated October 11, 2001.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) surveillance requirement (SR) for the 
methodology and frequency for the chemical analyses of the ice 
condenser ice bed. Also, these amendments add a new TS SR to address 
sampling requirements for ice

[[Page 10018]]

additions to the ice bed. In addition, the amendments revise the 
current TS surveillance requirement acceptance criteria and 
surveillance frequency for the inspection of ice condenser ice basket 
flow channel areas.
    Date of issuance: February 11, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 195 and 188.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 11, 2001 (66 FR 
36339). The supplement dated October 11, 2001, provided clarifying 
information that did not expand the scope of the original Federal 
Register notice or the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated February 11, 2002.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: July 13, 2001.
    Brief description of amendment: The amendment deleted Technical 
Specification (TS) Tables 3.6-1, ``Non-Automatic Containment Isolation 
Valves Open Continuously or Intermittently for Plant Operation,'' and 
4.4-1, ``Containment Isolation Valves.'' The amendment also revised 
other TS sections that reference these tables. The removal of the 
tables is in accordance with the guidance in NRC Generic Letter 91-08, 
``Removal of Component Lists from Technical Specifications.''
    Date of issuance: February 12, 2002.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 223.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44166). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 12, 2002.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: July 16, 2001, as supplemented 
January 11, 2002.
    Brief description of amendment: The amendment updates the pressure-
temperature limit curves for Indian Point Nuclear Generating Unit No. 
2.
    Date of issuance: February 15, 2002.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 224.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41613). The January 11, 2002, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated February 15, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 23, 2001, as supplemented by 
letters dated September 21, and November 8, 2001.
    Brief description of amendment: This submittal requests a change to 
administrative Technical Specification (TS) 6.15. The change postpones 
the next Type A test performed after May 12, 1991, to no later than May 
11, 2006, resulting in an extended interval of 15 years for the 
performance of Integrated Leak Rate Test.
    Date of issuance: February 14, 2002.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment No.: 178.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44169). The September 21, and November 8, 2001, supplemental letters 
contained clarifying information that did not change the scope of the 
July 23, 2001, application nor the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
February 14, 2002.
    No significant hazards consideration comments received: No.
    Entergy Operations, Inc., Docket No. 50-382, Waterford Steam 
Electric Station, Unit 3, St. Charles Parish, Louisiana
    Date of amendment request: July 23, 2001, as supplemented December 
11, 2001.
    Brief description of amendment: As a follow-up response to a 
commitment identified in the Nuclear Regulatory Commission (NRC) staff 
letter dated December 22, 2000, ``Completion of Licensing Action for 
Generic Letter (GL) 96-06, Assurance of Equipment Operability and 
Containment Integrity During Design-Basis Accident Conditions,'' 
Entergy Operations Inc., (Entergy, the licensee) has proposed to revise 
their Waterford Steam Electric Station, Unit 3 (Waterford 3) Final 
Safety Analysis Report (FSAR) to resolve the ten containment 
penetrations susceptible to thermally induced overpressurization 
through an evaluation, detailed analysis, or installation of physical 
modifications prior to startup from the spring 2002 refueling outage. 
Entergy determined a change to Waterford 3's license basis, through 
procedural controls, risk analysis, and engineering analysis, for seven 
penetrations, as discussed in this license basis change request was 
necessary. Permanent resolution to the GL 96-06 issues for the 
remaining three penetrations will be satisfied through the installation 
of physical modifications.
    Date of issuance: February 19, 2002.
    Effective date: This license amendment is effective as of its date 
of issuance and shall be implemented prior to startup from Refuel 11 
scheduled for March 2002. Implementation of the amendment is the 
incorporation into the FSAR of the changes to the description of the 
facility as described in the licensee's application dated July 23, 
2001, as supplemented by letter dated December 11, 2001.
    Amendment No.: 179.
    Facility Operating License No. NPF-38: The amendment revised the 
FSAR.
    Date of initial notice in Federal Register: September 19, 2001 (66 
FR 48285). The December 11, 2001, supplement contained clarifying 
information that did not change the scope of the July 23, 2001, 
application nor the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated February 19, 2002.
    No significant hazards consideration comments received: No.

[[Page 10019]]

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

    Date of application for amendments: July 6, 2001, as supplemented 
by letters dated October 25, 2001, and December 17, 2001.
    Brief description of amendments: The amendments revise technical 
specifications (TS) Section 3.3.1.1, ``Reactor Protection System 
Instrumentation,'' to modify the description for Reactor Protection 
System (RPS) Function 7.a, ``Scram Discharge Volume Water Level--
High.'' This change supports a planned upgrade to the scram discharge 
volume level instrumentation from Fluid Components International 
thermal switches to Magnetrol float switches. These float switches are 
more reliable than the existing thermal switches, which are highly 
sensitive to a steam environment, since they respond to actual water 
level increases within the scram discharge volume. These types of 
Magnetrol float switches are used successfully in various applications 
at Quad Cities.
    Date of issuance: February 11, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 203 and 199.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 11, 2002 (67 FR 
1520). The October 25, 2001, and December 17, 2001, supplements 
provided clarifying information that was within the scope of the 
original Federal Register notice and did not change the staff's initial 
proposed no significant hazards considerations determination. The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated February 11, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 13, 2001, as 
supplemented by letters December 17 and December 26, 2001, and January 
10, 2002.
    Brief description of amendments: The amendments revise the 
Technical Specifications to support the licensee's planned upgrade of 
the reactor water level instrumentation, including changes to 
surveillance requirements frequencies, functional testing, and 
allowable values.
    Date of issuance: February 12, 2002.
    Effective date: As of the date of issuance and shall be implemented 
for Unit 1 prior to reaching Startup (i.e., Mode 2) following refueling 
outage 17, scheduled for completion in November 2002, and for Unit 2 
prior to reaching Startup (i.e., Mode 2) following refueling outage 16, 
scheduled for completion in February 2002.
    Amendment Nos.: 204 and 200.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 17, 2001 (66 FR 
52800). The December 17 and December 26, 2001, and January 10, 2002, 
supplements provided clarifying information that was within the scope 
of the original Federal Register notice and did not change the staff's 
initial proposed no significant hazards considerations determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated February 12, 2002.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, 
Beaver Valley Power Station, Unit No. 1, Beaver County, Pennsylvania

    Date of application for amendment: June 29, 2001, as supplemented 
by letters dated October 4, and December 1, 2001.
    Brief description of amendment: The amendment revised the pressure-
temperature curves and the cold overpressure protection limits. The 
changes are supported by a new fluence determination based on 
evaluation of a surveillance capsule, and the use of the American 
Society of Mechanical Engineers Code Case N-640.
    Date of issuance: February 20, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No: 249.
    Facility Operating License No. DPR-66: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 17, 2001 (66 FR 
52801). The supplemental letters provided additional information but 
did not change the initial proposed no significant hazards 
consideration determination or expand the amendment beyond the scope of 
the initial notice. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated February 20, 2002.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit 2, Beaver County, Pennsylvania

    Date of application for amendment: March 28, 2001, as supplemented 
by letter dated September 25, 2001.
    Brief description of amendment: The amendment revised the technical 
specifications associated with crediting soluble boron for reactivity 
control in the spent fuel pool.
    Date of issuance: February 11, 2002.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment No: 128.
    Facility Operating License No. NPF-73: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41620). The September 25, 2001, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination or expand the scope of the original Federal 
Register notice. The Commission's related evaluation of the amendment 
is contained in a Safety Evaluation dated February 11, 2002.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit 2, Beaver County, Pennsylvania

    Date of application for amendment: June 28, 2001, as supplemented 
September 13, 2001, December 19, 2001, and January 21, 2002.
    Brief description of amendment: The amendment revised the technical 
specification (TS), 3.1.1.4, upper limit for the moderator temperature 
coefficient (MTC), from 0  x  10 -4 change in reactivity per 
degree Fahrenheit (k/k/ deg.F) to +0.2  x  10 -4 
k/k/ deg.F for power levels up to 70 percent of rated thermal 
power (RTP), and ramping linearly to 0  x  10 -4 k/
k/ deg.F from 70 percent to 100 percent RTP. The change is needed to 
address future core designs with higher energy requirements, associated 
with plant operation at higher capacity factors.
    Date of issuance: February 21, 2002.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment No: 129.
    Facility Operating License No. NPF-73. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55019). The September 13, 2001,

[[Page 10020]]

December 19, 2001, and January 21, 2002, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
February 21, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 1, 2001.
    Brief description of amendment: FirstEnergy Nuclear Operating 
Company submitted License Amendment Request (LAR) 00-0003 for Davis-
Besse Nuclear Power Station, Unit 1, for review and approval. This 
amendment request proposes to revise references in the Technical 
Specification (TS) to the American Society of Mechanical Engineers 
(ASME) Boiler and Pressure Vessel Code (ASME Code), Section XI, as the 
source of requirements for the inservice testing of ASME Code Class 1, 
2, and 3 pumps and valves. The TS will reference the ASME Code for 
Operation and Maintenance of Nuclear Power Plants (ASME OM Code).
    Date of issuance: January 31, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 250.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 30, 2001 (66 FR 
29375). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 31, 2002.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: May 22, 2001, as supplemented by 
letters dated November 15, 2001, February 12, 2002, and electronic 
transmission dated February 19, 2002.
    Brief description of amendment: This amendment revised Davis-Besse 
Nuclear Power Station (DBNPS) Technical Specifications in accordance 
with Framatone Technologies Incorporated Topical Report BAW-2303P, 
Revision 4, ``OTSG Repair Roll Qualification Report.'' The changes 
revise the existing DBNPS Once-Through Steam Generators (OTSGs) repair 
roll requirements to (1) use updated limiting tensile tube loads, (2) 
define new exclusion zones within the steam generator in which 
application of the repair roll is prohibited, (3) allow the repair roll 
to be used in the lower tubesheet area, (4) remove the limitation of 
only one repair roll per OTSG tube, and (5) replace the requirement 
that the repair roll be one inch in length with a requirement that the 
repair roll be installed in accordance with Topical Report BAW-2303P, 
Revision 4.
    Date of issuance: February 20, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 252.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41621). The supplemental information contained clarifying information 
and did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice. The Commission's related evaluation of the amendment 
is contained in a Safety Evaluation dated February 20, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 15, 2001, as supplemented by 
letter dated February 8, 2002.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) 3/4.9.4, ``Refueling Operations--Containment 
Penetrations,'' TS 3/4.9.12, ``Refueling Operations--Storage Pool 
Ventilation,'' and associated Bases. The changes will allow the 
containment equipment hatch cover to be removed during core alterations 
and movement of irradiated fuel inside containment provided the 
Emergency Ventilation System is operable with the ability to filter any 
radioactive release.
    Date of issuance: February 14, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 251.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 27, 2001 (66 FR 
34284). The supplemental information contained clarifying information 
and did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice. The Commission's related evaluation of the amendment 
is contained in a Safety Evaluation dated February 14, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 5, 2001.
    Brief description of amendment: This amendment revised Technical 
Specification 5.5.11, ``Technical Specifications (TSs) Bases Control 
Program,'' to reflect Technical Specification Task Force (TSTF) 
Standard Technical Specification Traveler, TSTF-364, Revision 0, 
``Revision to Technical Specification Bases Control Program to 
Incorporate Changes to 10 CFR 50.59.''
    Date of Issuance: February 21, 2002.
    Effective Date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 121.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 8, 2002 (67 FR 
927). The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 21, 2002.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: March 29, 2001, as supplemented 
October 30, 2001.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TS) and TS Bases to eliminate the requirements for 
certain engineered features operability during core alterations and 
movement of irradiated fuel which had decayed for at least 2 days.
    Date of issuance: February 11, 2002.
    Effective date: As of the date of issuance, to be implemented prior 
to Refueling Outage 8.
    Amendment No.: 101.
    Facility Operating License No. NPF-69: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 30, 2001 (66 FR 
29358). The licensee's October 30, 2001, supplement withdrew portions 
of the

[[Page 10021]]

original application, leaving the balance unchanged.
    The staff's related evaluation of the amendment is contained in a 
Safety Evaluation dated February 11, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 17, 2001, as supplemented 
July 26, and October 15, 2001.
    Brief description of amendments: The amendments update the reactor 
pressure vessel pressure-temperature limit curves for SSES-1 and 2.
    Date of issuance: February 7, 2002.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 200, 174.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 3, 2001 (66 FR 
50471). The July 26, and October 15, 2001, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
February 7, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 8, 2001, as supplemented on 
December 14, 2001.
    Brief description of amendments: The amendments add the following 
to the Technical Specifications: (1) The phrase, ``or if open, capable 
of being closed'' to Limiting Condition for Operation 3.9.4 for the 
equipment hatch, during core alterations or movement of irradiated fuel 
assemblies inside containment, and (2) the requirement to verify the 
capability to install the equipment hatch in a new Surveillance 
Requirement (SR) 3.9.4.2. The previous SR 3.9.4.2 was renumbered SR 
3.9.4.3, but not changed.
    Date of issuance: February 20, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance, including the incorporation 
of the changes to the Technical Specification Bases as described in the 
licensee's application dated November 8, 2001.
    Amendment Nos.: 93 and 93.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64307). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 20, 2002.
    No significant hazards consideration comments received: No.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland. Publicly available records will be 
accessible from the Agencywide Documents Assess and Management Systems 
(ADAMS) Public Electronic Reading Room on the internet at the NRC Web 
site, http://www.nrc.gov/NRC/ADAMS/index.html. If you do not have 
access to ADAMS or

[[Page 10022]]

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, 304-415-4737 or by e-mail to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By April 4, 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 which is available at the 
Commission's Public Document Room, located at One White Flint North, 
11555 Rockville Pike (first floor), Rockville, Maryland 20852, and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room). If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, by the above date. A copy of 
the petition should also be sent to the Office of the General Counsel, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-001, and to 
the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

    Date of amendment request: February 8, 2002, as supplemented 
February 10, 2002.
    Description of amendment request: The amendment adds a license 
condition allowing a one-time limited duration exception from the 
Technical Specification (TS) Surveillance Requirement (SR) to verify 
that the opening, closing, and frictional torque of the ice condenser 
inlet doors are within specified limits as required by TS SRs 
4.6.5.3.1.b.3, 4.6.5.3.1.b.4, and 4.6.5.3.1.b.5, respectively.
    Date of issuance: February 14, 2002.
    Effective date: As of the date of issuance, to be implemented 
immediately.
    Amendment No.: 265.
    Facility Operating License No. (DPR-58): Amendment revises the 
Operating License.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): No. The Commission's related evaluation of the 
amendment, finding of emergency circumstances, state consultation, and 
final NSHC determination are contained in a Safety Evaluation dated 
February 14, 2002.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: William D. Reckley, Acting Section Chief.

    Dated at Rockville, Maryland, this 25th day of February, 2002.
    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 02-5000 Filed 3-4-02; 8:45 am]
BILLING CODE 7590-01-P