[Federal Register Volume 67, Number 112 (Tuesday, June 11, 2002)]
[Notices]
[Pages 40019-40032]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-14339]


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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 May 17, 2002, through May 30, 2002. The 
last

[[Page 40020]]

biweekly notice was published on May 28, 2002 (67 FR 36924).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the 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 July 11, 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.

[[Page 40021]]

    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 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].

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

    Date of amendment request: May 7, 2002.
    Description of amendment request: The proposed amendment would 
relocate the Boration System Technical Specification (TS) requirements 
to the Technical Requirements Manual (TRM). Additional TS changes to 
retain boron dilution analysis restrictions would be made as a result 
of the relocation of the Boration System TS requirements to the TRM. 
The proposed amendment would also revise the TS Limiting Condition for 
Operation, action requirements, and surveillance requirements 
associated with the Emergency Core Cooling, Containment Spray and 
Cooling, and Auxiliary Feedwater Systems. The proposed changes would 
remove redundant testing requirements that are already addressed by the 
Inservice Testing Program, which is required pursuant to TS 4.0.5. The 
proposed changes would also increase the allowed outage time and 
shutdown time for an inoperable train (subsystem) of the Emergency Core 
Cooling System, consistent with standard industry guidelines and other 
Millstone Unit No. 2 TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The staff's review is 
presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes will not alter the way any structure, 
system, or component functions, and will not alter the manner in 
which the plant is operated. The proposed changes to the TSs do not 
impact any system or component that could cause an accident. The 
ability of the equipment associated with the proposed changes to 
mitigate the design-basis accidents will not be affected. In 
addition, the design-basis accidents will remain the same postulated 
events described in the Millstone Unit No. 2 Final Safety Analysis 
Report, and the consequences of those events will not be affected. 
Therefore, the proposed changes will not increase 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 not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
unusual operator actions. The proposed changes will not alter the 
way any structure, system, or component functions, and will not 
alter the manner in which the plant is operated. There will be no 
adverse effect on plant operation or accident mitigation equipment. 
The response of the plant and the operators following an accident 
will not be different. In addition, the proposed changes do not 
introduce any new failure modes. Therefore, the proposed changes 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes to the TSs do not impact any system or 
component that could cause an accident and will not result in any 
change in the operational characteristics of the associated accident 
mitigation equipment. The equipment associated with the proposed TS 
changes will continue to be able to mitigate the design-basis 
accidents as assumed in the safety analysis. In addition, the 
proposed changes will not affect equipment design and there are no 
changes being made to the TS-required safety limits or safety system 
settings. Therefore, the proposed changes will not result in a 
reduction in a margin of safety.

    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: April 24, 2002.
    Description of amendment request: Entergy Operations, Inc. requests 
revision of the River Bend Station, Unit 1 licensing basis and 
Technical Specifications to utilize the alternative accident source 
term described in NUREG-1465.
    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:

    This proposed amendment to the River Bend Technical 
Specifications (TS) revises those specifications affected by the 
implementation of the alternative source term concepts in accordance 
with NUREG 1465. In addition, based on the alternative source term, 
changes are proposed to selected specifications associated with 
handling irradiated fuel in the primary containment or Fuel Building 
and CORE ALTERATIONS. The alternative source term changes affect the 
definitions, and the specifications for the Control Room Fresh Air 
System, Standby Gas Treatment System, Fuel Building Ventilation 
System and leakage rates for Primary Containment and the Personnel 
Airlocks seal air systems.
    Entergy Operations, Inc. [Entergy] has evaluated whether or not 
a significant hazards consideration is involved with the proposed 
amendment by focusing on the three standards set forth in 10 CFR 
50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The alternative source term does not require modification of the 
facility; rather, once the occurrence of an accident has been 
postulated the new source term is an input to evaluate the potential 
consequences. The implementation of the alternative source

[[Page 40022]]

term has been evaluated in revisions to the analyses of the limiting 
design basis accidents at River Bend Station. Based on the results 
of these analyses, it has been demonstrated that, even with the 
requested Technical Specification changes, the dose consequences of 
these limiting events are within the regulatory guidance currently 
approved by the NRC for use with the alternative source term. This 
guidance is presented in Regulatory Guide 1.183, 10CFR50.67 and 
Standard Review Plan Section 15.0.1, ``Radiological Consequences 
Analyses Using Alternative Source Terms.''
    Because the equipment affected by the revised operational 
conditions is not considered an initiator to any previously analyzed 
accident, inoperability of the equipment cannot increase the 
probability of any previously evaluated accident. The proposed 
requirements bound the conditions of the current design basis fuel 
handling accident analysis which concludes that the radiological 
consequences are within the acceptance criteria of NUREG 0800, 
Section 15.7.4 and General Design Criteria 19. As noted above, with 
the alternative source term implementation, the acceptance criteria 
are also being revised. The results of the revised Fuel Handling 
Accident demonstrate that the dose consequences are within the NRC 
regulatory guidance. This guidance is presented in Regulatory Guide 
1.183, 10CFR50.67 and Standard Review Plan Section 15.0.1, 
``Radiological Consequences Analyses Using Alternative Source 
Terms.''
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes using the alternative source term dose 
methodology are analytical in nature and do not physically alter the 
facility or of any equipment within the facility. Similarly, the 
alternative source term does not create any new initiators or 
precursors of a new or different kind of accident. The proposed 
changes to the Technical Specifications, while they revise certain 
performance requirements, do not involve any physical modifications 
to the plant.
    The proposed changes related to shutdown controls based on the 
alternative source term do not create the possibility of a new or 
different kind of accident from any previous analyzed.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The changes above are associated with the implementation of a 
new licensing basis for River Bend Station. Approval of the basis 
change from the original source term in accordance with TID-14844 to 
the new alternative source term of NUREG-1465 is requested by this 
submittal. The results of the accident analyses prepared in support 
of this submittal are subject to revised acceptance criteria. These 
analyses have been performed using conservative methodologies as 
outlined in the regulatory guidance and conservatively represent the 
requested Technical Specification changes. Safety margins and 
analytical conservatisms have been evaluated and are well 
understood. The analyzed events have been carefully selected and 
margin has been retained to ensure that the analyses adequately 
bound all postulated event scenarios. The dose consequences of these 
limiting events are within the acceptance criteria also found in the 
latest regulatory guidance. This guidance is presented in Regulatory 
Guide 1.183, 10CFR50.67 and Standard Review Plan Section 15.0.1, 
``Radiological Consequences Analyses Using Alternative Source 
Terms.''
    The proposed changes continue to ensure that the doses at the 
exclusion area and low population zone boundaries as well as control 
room, are within the corresponding regulatory limits. In a similar 
way, the results of the existing analyses demonstrated that the dose 
consequences were within the applicable NRC-specified regulatory 
limit. Specifically, the margin of safety for these accidents is 
considered to be that provided by meeting the applicable regulatory 
limit for Alternate Source Term methodologies, which, for most 
events, is conservatively set at, or below, the 10CFR50.67 limit. 
With respect to the control room personnel doses, the margin of 
safety is the difference between the 10CFR100 limits and the 
regulatory limit defined by 10CFR50, Appendix A, General Design 
Criterion (GDC) 19.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, Entergy concludes that the proposed 
amendment(s) present no significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

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

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

    Date of amendment request: May 14, 2002.
    Description of amendment request: Entergy Operations, Inc. is 
proposing that the River Bend Station, Unit 1 Operating License be 
amended to reflect a 1.7 percent increase in the licensed 100% reactor 
core thermal power level (an increase in reactor power level from 3,039 
megawatts thermal to 3,091 megawatts thermal). These changes result 
from increased accuracy of the feedwater flow measurement to be 
achieved by utilizing high accuracy ultrasonic flow measurement 
instrumentation. The basis for this change is consistent with the 
revision, issued in June 2000, to appendix K to part 50 of title 10 of 
the Code of Federal Regulations, allowing operating reactor licensees 
to use an uncertainty factor of less than 2 percent of rated reactor 
thermal power in analyses of postulated design basis loss-of-coolant 
accidents.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The comprehensive analytical efforts performed to support the 
proposed change included a review of the Nuclear Steam Supply System 
(NSSS) systems and components that could be affected by this change. 
All systems and components will function as designed, and the 
applicable performance requirements have been evaluated and found to 
be acceptable.
    The comprehensive analytical efforts performed to support the 
proposed uprate conditions included a review and evaluation of all 
components and systems that could be affected by this change. 
Evaluation of accident analyses confirmed the effects of the 
proposed uprate are bounded by the current dose analyses. All 
systems will function as designed, and all performance requirements 
for these systems have been evaluated for the uprate conditions and 
found acceptable. Because the integrity of the plant will not be 
affected by operation at the new power level conditions, it is 
concluded that all structures, systems, and components required to 
mitigate a transient remain capable of fulfilling their intended 
functions. The reduced uncertainty in the flow input to the power 
calorimetric measurement allows the current safety analyses to be 
used, with small changes to the core operating limits, to support 
operation at a core power of 3,091 megawatts thermal (MWt). As such, 
all Final Safety Analysis Report (FSAR) Chapter 15 accident analyses 
continue to demonstrate compliance with the relevant event 
acceptance criteria. Those analyses performed to assess the effects 
of mass and energy releases remain valid. The source terms used to 
assess radiological consequences have been reviewed and determined 
to either bound operation at the new power level condition, or new 
analyses were performed to verify all acceptance criteria continue 
to be met.

[[Page 40023]]

    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on any safety-related system or component and do not 
challenge the performance or integrity of any safety related system.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Operation at the uprated power condition does not involve a 
significant reduction in a margin of safety. Analyses of the primary 
fission product barriers have concluded that all relevant design 
criteria remain satisfied, both from the standpoint of the integrity 
of the primary fission product barrier and from the standpoint of 
compliance with the required acceptance criteria. The calculated 
loads on all affected structures, systems and components have been 
shown to remain within design criteria for all design basis event 
categories. No NRC [U.S. Nuclear Regulatory Commission] acceptance 
criterion is exceeded.
    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: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of amendment request: April 19, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.6.6 surveillance requirement (SR) 
to verify each spray nozzle on the containment spray ring headers at 
the top of containment dome is unobstructed. The current TS 3.6.6.8 
requirement is to verify each spray nozzle every 10 years. The proposed 
requirement is to revise the frequency to ``Following maintenance that 
could result in nozzle blockage OR Following fluid flow through the 
nozzles.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change revises the Frequency for Technical 
Specifications (TS) Surveillance Requirement (SR) 3.6.6.8 for 
verifying each spray nozzle is unobstructed from ``10 years'' to 
``Following maintenance that could result in nozzle blockage OR 
Following fluid flow through the nozzles.''
    Analyzed events are initiated by the failure of plant 
structures, systems, or components. The Containment Spray (CS) 
system is not considered as an initiator of any analyzed event. The 
proposed change does not have a detrimental impact on the integrity 
of any plant structure, system, or component that initiates an 
analyzed event. No active or passive failure mechanisms that could 
lead to an accident are affected. The proposed change will not alter 
the operation of, or otherwise increase the failure probability of 
any plant equipment that initiates an analyzed accident. Therefore, 
the proposed change does not involve a significant increase in the 
probability of an accident previously evaluated.
    The initial conditions of Design Basis Accident (DBA) and 
transient analyses in the Byron/Braidwood Stations' UFSAR assume the 
CS system is operable.
    The operability of the CS system in accordance with the proposed 
TS is consistent with the initial assumptions of the accident 
analyses and is based upon meeting the design basis of the plant. 
Since plant safety can be ensured at the proposed Frequency, we are 
proposing to revise the CS system testing provisions to require 
nozzle testing only after activities that could result in nozzle 
blockage, i.e., following maintenance that could result in nozzle 
blockage or following fluid flow through the nozzles. Nozzle 
blockage is considered unlikely during periods without maintenance 
or without fluid flow through the nozzles, since the nozzles are of 
a passive design and the system is kept in a normally dry state, 
thus minimizing corrosion susceptibility. In addition, the location 
of the nozzles at the top of the containment dome limits the 
possibility of the introduction of foreign material from sources 
external to the CS system. The proposed Frequency will continue to 
provide confidence that an unobstructed flow path is available, and 
will preclude the need for unnecessary testing when no activities 
have occurred that would introduce debris to the spray ring headers, 
or when no other active degradation mechanism is present. 
Operability of the CS system will not be affected. Therefore, the 
proposed change does not involve a significant increase in the 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change does not involve the use or installation of 
new equipment. Installed equipment is not operated in a new or 
different manner. No new or different system interactions are 
created, and no new processes are introduced. The current foreign 
material exclusion practices have been reviewed and judged 
sufficient to provide high confidence that debris will not be 
introduced during times when the CS system boundary is breached. The 
design of the CS system at Braidwood and Byron Stations precludes 
borated water from reaching the spray nozzles, except during a CS 
actuation. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed change does not introduce any new setpoints at 
which protective or mitigative actions are initiated. No current 
setpoints are altered by this change. The design and functioning of 
the CS system is unchanged. Since the system is not susceptible to 
corrosion induced obstruction nor is the introduction of foreign 
material from external sources likely, and the design of the CS 
system at Braidwood and Byron Stations precludes borated water from 
reaching the spray nozzles except during a CS actuation, the 
proposed testing Frequency is sufficient to provide high confidence 
that the CS system will continue to function as designed. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.
    Therefore, based on the above evaluation, we have concluded that 
the proposed change does not involve any significant hazards 
consideration.

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

[[Page 40024]]

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

    Date of amendment request: February 15, 2002.
    Description of amendment request: Exelon proposed to increase the 
trip setpoints for Items 3.b and 3.c in Table 3.3.2-2, for the Reactor 
Water Cleanup System (RWCS) steam leak detection temperature isolation 
actuation instrumentation in the technical specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by Section 50.91(a) the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The RWCS is not required for safety purposes nor is it 
required to operate after a design-basis accident. The RWCS 
instrumentation and controls are not required for safe operation of 
the reactor. They provide a means of monitoring parameters and 
protecting the system. The increase in the isolation setpoint and 
allowable value for the RWCS pump room high ambient temperature and 
high differential temperature will not make any physical changes 
(modification) to the plant equipment. Therefore, the proposed 
changes to the RWCS setpoints will not increase the probability of 
an accident previously evaluated.
    This license amendment request (LAR) does not increase the 
consequences of an accident previously evaluated in the Updated 
Final Safety Analysis Report (UFSAR). This proposed change has no 
impact on the high-energy line break or loss-of-coolant accident 
(LOCA) accident analyses. This LAR does not adversely affect 
mitigating systems, structures or components (SSCs), and does not 
adversely affect the initial conditions of any accidents. Affected 
equipment will remain within the limitations of the Environmental 
Qualification Program. Redundancy and diversity of mitigating 
systems are unchanged as a result of this LAR. This LAR does not 
affect onsite or offsite radiological consequences of any accident 
previously evaluated in the UFSAR.
    Therefore, this LAR 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?
    No. The increase in the RWCS pump room high ambient temperature 
and high differential temperature settings proposed by this LAR does 
not change any SSC. This LAR does not create new operating or 
failure modes. Existing instruments are not accident initiators in 
any failure mode and changing settings does not change the 
instrument's functions. Therefore, this LAR 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?
    No. This LAR will allow the plant to operate at higher ambient 
temperatures in the RWCS pump rooms during normal operation. This 
change does not create additional heat loads or change the way any 
of the equipment is operated. No safety-related setpoints are 
associated with the RWCS system. The RWCS system instrumentation and 
controls are not required for safe operation of the reactor. They 
provide a means of monitoring parameters and protecting the system. 
Therefore, a change to the TSs for RWCS pump room high ambient 
temperature and high differential temperature limits to the new 
setpoints is not considered a significant reduction in a margin of 
safety.

    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Mr. Edward Cullen, Vice President & General 
Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett 
Square, PA 19348.
    NRC Section Chief: James W. Clifford.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook 
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: April 24, 2002.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) to relocate to the Seabrook 
Station Technical Requirements (SSTR) Manual, specific pressure, 
differential pressure and flow values, as well as specific test 
methods, contained in Surveillance Requirements (SRs) 4.6.2.1, 
``Containment Spray System,'' and 4.7.1.2.1b, ``Auxiliary Feedwater 
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. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes to relocate the specific pump pressure and 
flow criteria TS SRs to the SSTR are administrative in nature and do 
not adversely affect accident initiators or precursors, or alter the 
design assumptions, conditions, and configuration of the facility or 
the manner in which it is operated. The proposed changes do not 
alter or prevent the ability of structures, systems, or components 
to perform their intended function to mitigate the consequences of 
an initiating event within the acceptance limits assumed in the 
Updated Final Safety Analysis Report (UFSAR).
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes do not alter the design assumptions, 
conditions, or configuration of the facility or the manner in which 
it is operated. The proposed changes have no adverse impact on 
component or system interactions. Since there are no changes to the 
design assumptions, parameters, conditions and configuration of the 
facility, or the manner in which the plant is operated and 
surveilled, the proposed changes do not create the possibility of a 
new or different accident from any previously analyzed.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    There is no adverse impact on equipment design or operation and 
there are no changes being made to the TSs themselves that would 
adversely affect any current margin of safety. The proposed changes 
are administrative in nature and impose alternative procedural and 
programmatic controls on these parameter limits.
    Therefore, relocation of the specific pump pressure and flow 
criteria do not involve a significant reduction in a margin of 
safety.

    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: William J. Quinlan, Esq. Assistant General 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment requests: April 10, 2002.
    Description of amendment requests: The proposed license amendments 
would revise several of the Required Actions in the DCPP Technical 
Specifications (TS) that require suspension of operations involving

[[Page 40025]]

positive reactivity additions or suspension of operations involving 
reactor coolant system (RCS) boron concentration reductions. In 
addition, this license amendment request (LAR) proposes to revise 
several Limiting Condition for Operation (LCO) Notes that preclude 
reductions in RCS boron concentration when a reactor coolant pump(s) 
and/or a residual heat removal pump(s) are removed from operation. The 
proposed changes would allow small, controlled, safe insertions of 
positive reactivity, but limit the introduction of positive reactivity 
to ensure that compliance with the required shutdown margin or 
refueling boron concentration limits will still be satisfied.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since there are 
no hardware changes. The reactor trip system instrumentation and 
reactivity control systems will be unaffected. Protection systems 
will continue to function in a manner consistent with the plant 
design basis. All design, material, and construction standards that 
were applicable prior to the request are maintained.
    The probability and consequences of accidents previously 
evaluated in the Final Safety Analysis Report Update (FSAR) are not 
adversely affected because the changes to the Required Actions and 
LCO Notes assure the limits on SDM [shutdown margin] and refueling 
boron concentration continue to be met, consistent with the analysis 
assumptions and initial conditions included within the safety 
analysis and licensing basis. The activities covered by this LAR are 
routine operating evolutions. The proposed changes do not reduce the 
capability to borate the RCS.
    The equipment and processes used to implement RCS boration or 
dilution evolutions are unchanged and the equipment and processes 
are commonly used throughout the applicable modes under 
consideration. There will be no degradation in the performance of or 
an increase in the number of challenges imposed on, safety-related 
equipment assumed to function during an accident. There will be no 
change to normal plant operating parameters or accident mitigation 
performance.
    The proposed changes will not alter any assumptions or change 
any mitigation actions in the radiological consequence evaluations 
in the FSAR.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no hardware changes or any changes in the method by 
which any safety-related plant system performs its safety function. 
This amendment will not affect the normal method of plant operation 
or change any operating limits. The proposed changes permit the 
conduct of normal operating evolutions when additional controls over 
core reactivity are imposed by the TS. The proposed changes do not 
introduce any new equipment into the plant or alter the manner in 
which existing equipment will be operated. The changes to operating 
procedures are minor, with clarifications provided that required 
limits must continue to be met. No performance requirements or 
response time limits will be affected. These changes are consistent 
with assumptions made in the safety analysis and licensing basis 
regarding limits on SDM and refueling boron concentration.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this LAR. There will be no adverse effect or challenges imposed 
on any safety-related system as a result of this LAR.
    This LAR does not alter the design or performance of the reactor 
protection system, nuclear instrumentation system, or solid state 
protection system used in the plant protection systems.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do not alter the limits on SDM or refueling 
boron concentration. These limits continue to assure that core 
parameters remain within the bounds of the accident analysis. The 
nominal trip setpoints specified in the TS and the safety analysis 
limits assumed in the transient and accident analyses are unchanged. 
None of the acceptance criteria for any accident analysis is 
changed.
    The proposed changes do not affect the manner in which safety 
limits or limiting safety system settings are determined, nor will 
there be any effect on those plant systems necessary to assure the 
accomplishment of protection functions. Also, the proposed changes 
do not impact the overpower limit, departure from nucleate boiling 
ratio limits, heat flux hot channel factor (FQ), nuclear 
enthalpy rise hot channel factor 
(F[]H), loss of coolant 
accident peak cladding temperature, peak local power density, or any 
other margin of safety. The radiological dose consequence acceptance 
criteria will continue to be met.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment requests: April 15, 2002.
    Description of amendment requests: The proposed license amendments 
would approve changes in the implementation of the DCPP Control of 
Heavy Loads Program and other analyses, design and procedure changes 
required to implement a dry cask Independent Spent Fuel Storage 
Installation (ISFSI) at DCCP.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    With the Holtec International (Holtec) HI-STORM 100 System and 
the associated design and handling procedures, most cask drops and 
other events, which could damage other spent fuel, have been 
precluded through redundant handling systems, control system 
upgrades, and mechanical stops/electrical interlocks that preclude 
crane movement over spent fuel, meeting PG&E's commitments to the 
guidelines of NUREG-0612, ``Control of Heavy Loads at Nuclear Power 
Plants.'' For those remaining cases where a cask drop is still 
credible, the impact-limiter design ensures the deceleration of the 
contained spent fuel remains below fuel design limits, preventing 
damage to the contained fuel assemblies (and associated structures), 
and meeting the analysis guidance of NUREG-0612. As a result of this 
design approach, a cask-handling accident that results in a 
significant offsite radiological release is not considered credible.
    Other Diablo Canyon Power Plant (DCPP) licensing-basis events, 
such as the drop of a spent fuel assembly, have not been affected by 
these changes and remain bounding events for potential radiological 
consequences.
    Revision of the DCPP Control of Heavy Loads Program ensures that 
PG&E's commitments to NUREG-0612 guidelines will protect the new 
fuel storage locations and the new transfer cask/multi-purpose 
canister (MPC) loading/unloading activities.

[[Page 40026]]

    The addition of restraint structures and use of impact limiters 
preclude adverse effects from seismic events and/or cask drops or 
tipovers, assuring that the fuel, MPC, transfer cask, and other 
potentially affected 10 CFR 50 structures remain within their design 
bases. The addition and installation of this equipment will be done 
after necessary evaluation and analysis is performed, to ensure the 
equipment does not introduce any unacceptable effect (e.g., seismic 
interaction).
    The proposed design of the dry cask system, the handling system, 
and associated procedural controls provide assurance that (1) 
operational errors and mishandling events, and (2) support system 
malfunctions will not result in an increase in the probability or 
consequence of an accident previously analyzed.
    The proposed changes to use the Holtec HI-STORM 100 system have 
been evaluated for seismic events and tornado missile impacts and it 
has been determined that these changes will not result in an 
increase in the probability or consequences of an accident 
previously evaluated.
    The Fire Protection Program will ensure that the combustible 
materials are properly controlled such that the total combustibles 
meet the current program commitments.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The engineering design measures and the handling procedures 
preclude the possibility of new or different kinds of accidents. 
Damage to 10 CFR 50 SSCs [structures, systems and components] from 
the cask handling and associated activities, and events resulting 
from possible damage to contained fuel, have been carefully 
considered in the following safety analyses. Both the types of 
accidents and the results remain within the envelope of existing 
analyses, as demonstrated by the PG&E and Holtec analyses.
    In Supplement No. 2 to the Safety Evaluation of DCPP (Reference 
7.18 [of the April 15, 2002, license amendment request]), the NRC 
reviewed and accepted Amendment 27 of the original DCPP Final Safety 
Analysis Report (FSAR) analysis of a cask-drop accident. Amendment 
22 to Facility Operating License No. DPR-80 and Amendment 21 to 
Facility Operating License No. DPR-82 allowed expansion of the spent 
fuel pool (SFP) storage capacity. In the safety evaluation for these 
amendments, the NRC reviewed the cask-drop accident and noted that 
the licensee had proposed administrative controls that would 
preclude the movement of a spent-fuel shipping cask in an exclusion 
zone over, and in the vicinity of, stored spent fuel that could 
result in a cask drop or tipping accident damaging stored spent 
fuel.
    Supplement No. 27 to the Safety Evaluation Report for DCPP Unit 
1 (Reference 7.19 [of the April 15, 2002, license amendment 
request]) and in Supplement No. 31 to the Safety Evaluation Report 
for Unit 2 (Reference 7.20 [of the April 15, 2002, license amendment 
request]) included the review and acceptance of the DCPP Control of 
Heavy Loads Program.
    The rupture of MPC dewatering, vacuum, forced helium dehydration 
or related closure system lines or the malfunction of equipment 
during cask handling operations resulting in radiological 
consequences are bounded by the DCPP Final Safety Analysis Report 
(FSAR) Update fuel-handling accident analysis.
    Other design considerations, such as SFP [spent fuel pool] 
thermal, water chemistry and clarity, criticality, and structural, 
were evaluated and determined not to introduce the possibility of a 
new or different kind of accident from any previously evaluated.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    With the Holtec HI-STORM 100 System, and the associated design 
and handling procedures, most cask drops and other events have been 
completely precluded through redundant load-handling systems, 
providing defense-in-depth as described in NUREG-0612, and meeting 
PG&E's commitments to the guidance of NUREG-0612. In those remaining 
cases where a cask drop is still credible, impact limiter design 
ensures that the deceleration of the contained spent fuel remains 
below fuel design limits, preventing damage to the contained fuel 
assemblies (and associated structures), and meeting the analysis 
guidelines of NUREG-0612. As a result of this design approach, the 
margin of safety has been maintained through the elimination of 
certain drops and the associated structural challenges.
    Other DCPP licensing-basis events, such as the drop of a spent 
fuel assembly, have not been affected by these changes and remain 
bounding events.
    Revision of DCPP Control of Heavy Loads Program to incorporate 
the additional restrictions on heavy loads movement will not affect 
the procedures or methodology used and will, therefore, not affect 
margins.
    The addition of restraint structures and use of impact limiters 
preclude adverse effects from seismic events and/or cask drops or 
tipovers, assuring that the fuel, MPC, transfer cask, and other 
potentially affected 10 CFR 50 structures remain within their design 
bases. Since design-basis criteria are fully satisfied, there is no 
impact on the margin of safety.
    The Fire Protection Program will continue to ensure that the 
combustible materials are properly controlled such that the total 
combustibles meet the current program commitments. Thus, there are 
no significant reductions in margin of safety associated with these 
changes.
    Other design considerations, such as SFP thermal, water 
chemistry, criticality, and structural, were evaluated and 
determined to not involve a reduction in a margin of safety.
    Therefore, the proposed changes do not involve a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment 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.

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

    Date of amendment request: April 8, 2002.
    Brief description of amendments: The proposed change would revise 
Technical Specification (TS) 3.4.16, ``RCS [Reactor Coolant System] 
Specific Activity,'' to lower the Limiting Condition For Operation and 
associated Surveillance Requirements for Dose Equivalent Iodine-131 in 
the Reactor Coolant System from a specific activity of 1.0 \Ci/gm to 
0.45 \Ci/gm. The change also includes approval of proposed changes to 
Technical Specification Bases for Main Steam Line Break post-accident 
radiological dose consequences analysis that was previously approved 
for implementing the Comanche Peak Steam Electric Station Steam 
Generator Alternate Repair Criteria.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to revise Technical Specification (TS) 
3.4.16 ``Reactor Coolant System Specific Activity'' to reduce the 
Limiting Condition For Operation (LCO) for Dose Equivalent I-131 in 
the reactor coolant from a specific activity of 1.0 \Ci/gm to 0.45 
\Ci/gm and the revised main steam line break (MSLB) radiological 
consequence analysis are used to determine post-accident dose. They 
are not related to any accident initiator. Therefore, this change 
cannot increase the probability of an accident.
    The revised MSLB offsite and control room radiological 
consequences analysis dose results are within 10 CFR Part 100 and 10 
CFR Part 50, Appendix A Criterion 19 limits and the NUREG-0800 SRP 
[Standard Review Plan] section 15.1.5 and section 6.4 guideline 
values.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 40027]]

    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to revise TS 3.4.16 ``Reactor Coolant System 
Specific Activity'' to reduce the LCO for Dose Equivalent I-131 in 
the reactor coolant from a specific activity of 1.0 \Ci/gm to 0.45 
\Ci/gm and the revised MSLB radiological consequence analysis do not 
involve any physical plant changes. The change does not involve 
changes in operation of the plant that could introduce a new failure 
mode for creating an accident or affect the mitigation of an 
accident.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to revise TS 3.4.16 ``Reactor Coolant System 
Specific Activity'' to reduce the LCO for Dose Equivalent I-131 in 
the reactor coolant from a specific activity of 1.0 \Ci/gm to 0.45 
\Ci/gm is a conservative change in that this reduced TS limit, when 
used in applicable plant radiological dose consequence analysis 
models with all other input parameters held constant, calculates 
decreased dose consequences to the thyroid. The change, with all 
other analysis input parameters held constant, increases the margin 
to acceptance limits. Therefore, this change does not result in a 
significant reduction in the margin provided by TS 3.4.16.
    The revised MSLB offsite and control room radiological 
consequences analysis dose results are within 10 CFR Part 100 and 10 
CFR Part 50, Appendix A Criterion 19 limits and the NUREG-0800 SRP 
section 15.1.5 and section 6.4 guideline values.
    Therefore the proposed change does not involve a reduction in a 
margin of safety.

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

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].

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].

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

    Date of application for amendments: January 31, 2002.
    Brief description of amendments: The amendments revised 
Surveillance Requirement (SR) 3.0.3 to extend the delay period, before 
entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period is extended from the current limit of 
``* * * up to 24 hours, or up to the limit of the specified Frequency, 
whichever is less'' to ``* * * up to 24 hours, or up to the limit of 
the specified Frequency, whichever is greater.'' In addition, the 
following requirement is added to SR 3.0.3: ``A risk evaluation shall 
be performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    Date of issuance: May 22, 2002.

[[Page 40028]]

    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 253 and 229.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 19, 2002 (67 FR 
12600). The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated May 22, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 1, 2001, as supplemented by 
letters dated November 28, 2001, December 17, 2001, January 24, 2002, 
February 4, 2002 (two letters), April 25, 2002, May 10, 2002 and May 
28, 2002.
    Description of amendment request: The amendments changed the 
Technical Specifications (TS) to replace the current accident source 
term used in design basis radiological analyses with an alternative 
source term pursuant to 10 CFR 50.67, ``Accident Source Term.'' License 
Conditions were added to the Unit 2 Operating License.
    Date of issuance: May 30, 2002.
    Effective date: Unit 1, upon issuance. Unit 2, upon completion of 
Refueling Outage 15.
    Amendment Nos: 221 and 246.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
changed the Technical Specifications and added License Conditions to 
DPR-62 only.
    Date of initial notice in Federal Register: September 5, 2001 (66 
FR 46477). The supplements contained clarifying information only, and 
did not change the initial no significant hazards consideration 
determination or expand the scope of the initial Federal Register 
notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 30, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 24, 2002, as supplemented 
March 26, 2002.
    Brief description of amendment: The amendment revises the Technical 
Specifications to delete Required Action 3.3.1.1.J.2, which specifies 
that the oscillation power range monitor upscale trip function be 
restored to operable status within 120 days when it is determined to be 
inoperable.
    Date of issuance: May 24, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 146.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 59503). The March 26, 2002, supplemental letter provided additional 
clarifying information that was 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 the amendment is contained 
in a Safety Evaluation dated May 24, 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: September 20, 2001, as 
supplemented on January 25 and April 29, 2002.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.8, ``Refueling, Fuel Storage and Operations with 
the Reactor Vessel Head Bolts Less Than Fully Tensioned,'' TS Table 
4.1-2, ``Frequencies for Sampling Tests,'' and TS 5.4, ``Fuel 
Storage,'' to allow credit for soluble boron in the criticality 
analysis for the spent fuel pit (SFP). The amendment also incorporates 
changes to the SFP rack layout by dividing it into sub-regions and 
specifying requirements for fuel assembly burnup and soluble boron 
concentration for various loading configurations in these sub-regions.
    Date of issuance: May 29, 2002.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 227.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55012). The January 25 and April 29, 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 May 29, 2002.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
County, Pennsylvania

    Date of application for amendments: November 30, 2001.
    Brief description of amendments: These amendments revised 
Surveillance Requirement (SR) 3.0.3 to extend the delay period, before 
entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period is extended from the current limit of 
``* * * up to 24 hours or up to the limit of the specified Frequency, 
whichever is less'' to ``* * * up to 24 hours or up to the limit of the 
specified Frequency, whichever is greater.'' In addition, the following 
requirement is added to SR 3.0.3: ``A risk evaluation shall be 
performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    Date of issuance: May 23, 2002.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendments Nos.: 243, 247.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 19, 2002 (67 
FR 7417). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 23, 2002.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County, 
Pennsylvania

    Date of application for amendments: May 22, 2001.
    Brief description of amendments: The amendments allowed the 
relocation of the Technical Specification (TS) sections associated with 
the curie content limit for liquid and gaseous waste storage and the TS 
sections associated with the explosive gas concentration limits to 
licensee controlled documents. In addition, the amendments allow for 
revisions to the reporting requirements of TS 6.9.3, ``Annual 
Radioactive Release Report.''
    Date of issuance: May 21, 2002.
    Effective date: Effective as of the date of issuance and shall be 
implemented within 90 days.
    Amendment Nos.: 250, 130.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.

[[Page 40029]]

    Date of initial notice in Federal Register: October 3, 2001 (66 FR 
50467). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 21, 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: November 8, 2000, as 
supplemented February 6, May 7, and November 21, 2001.
    Brief description of amendment: The amendment changed the technical 
specifications associated with the deletion of TS 3/4.4.1.6, ``Reactor 
Coolant Pump--Startup.''
    Date of issuance: May 30, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No: 131.
    Facility Operating License No. NPF-73: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81917). The February 6, May 7, and November 21, 2001, letters 
provided additional information that clarified the application but did 
not expand the scope of the application as originally noticed or change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 30, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 9, 2002, as supplemented 
April 25, 2002.
    Brief description of amendments: The amendments consist of changes 
to the Technical Specifications (TSs) in response to the application 
dated April 9, 2002, as supplemented April 25, 2002. In the April 25, 
2002, supplemental letter, the licensee requested that the portion of 
the original application dealing with the Unit 2 AB and CD train 
batteries for Unit 2 only be processed on an emergency basis. By letter 
dated April 26, 2002, the Nuclear Regulatory Commission issued 
Amendment No. 249 for Unit 2. The amendments revise the Surveillance 
Requirement (SR) for the Train AB, and CD batteries in TS 4.8.2.3.2.c.1 
for Unit 1 and SR TS 4.8.2.5.2.c.1 for the N train batteries in both 
Units 1 and 2. The amendments modify the requirements to verify that 
battery cells, cell plates and racks show no visual indication of 
physical damage or abnormal deterioration. The amendments would allow 
the operability of batteries exhibiting damage or deterioration to be 
determined by an evaluation. The amendments are consistent with an NRC-
approved change to the Standard Technical Specifications for 
Westinghouse plants (NUREG 1431, Revision 1) as documented in Technical 
Specification Task Force Standard Technical Specification.
    Date of issuance: May 30, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 7 days.
    Amendment Nos.: 269 and 250.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 25, 2002 ( 67 FR 
20552). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 30, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 28, 2002.
    Brief description of amendment: Amendment changes Technical 
Specification 3.0.3 to allow a longer time before entering a limiting 
condition for operation in the event of a missed surveillance and adds 
requirements to (1) perform a risk evaluation for any surveillance 
delayed greater than 24 hours and (2) manage the risk impact.
    Date of issuance: May 30, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 246.
    Facility Operating License No. DPR-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 30, 2002 (67 FR 
21290). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 30, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 14, 2001, as supplemented by 
letters dated January 15 and April 15, 2002.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 2.10.4(5)(a)(iii), ``DNBR [departure from nucleate 
boiling ratio] Margin During Power Operation Above 15% Rated Power,'' 
to decrease the minimum required reactor coolant system flow rate from 
206,000 gallons per minute (gpm) to 202,500 gpm. In addition, the Bases 
section for TS 2.10.4 has been revised to be consistent with the 
approved change to the TS.
    Date of issuance: May 24, 2002.
    Effective date: May 24, 2002 , and to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 209.
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2927). The January 15 and April 15, 2002, supplemental letters provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 24, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 10, 2001.
    Description of amendment request: The proposed amendments revised 
the Technical Specifications (TSs) to incorporate the Nuclear 
Regulatory Commission (NRC)-approved generic change Technical 
Specification Task Force-287, Revision 5, to the ``Standard Technical 
Specifications for General Electric Plants (BWR/4),'' NUREG-1433, 
Revision 1. Specifically, the changes: (a) Inserted 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) inserted a new LCO 
Action B in TS 3.7.3 to allow 24 hours to restore the control room 
habitability envelope boundary to 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-labeled the

[[Page 40030]]

existing LCO Actions B, C, D, and E to C, D, E, and F respectively; and 
(d) revised 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 
were also made.
    Date of issuance: May 17, 2002.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 203, 177.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 5, 2002 (67 FR 
10014). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 17, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 11, 2002.
    Brief description of amendments: The amendments revise TS Section 
1.1, Definitions, to change the definition of response time testing as 
it is applied to the Engineered Safety Features, and the Reactor 
Protective System, based on approved Technical Specification Task Force 
(TSTF) Traveler TSTF-368, Revision 0, ``Incorporate Combustion 
Engineering Owners Group (CEOG) Topical Report to Eliminate Pressure 
Sensor Response Time Testing.''
    Date of issuance: 1 May 22, 2002.
    Effective date: May 22, 2002, to be implemented within 60 days of 
issuance.
    Amendment Nos.: Unit 2-188; Unit 3-179.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 16, 2002 (67 FR 
18648). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 22, 2002.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of application for amendments: August 31, 2001, as 
supplemented by letter dated November 15, 2001, February 20 (two 
letters), dated February 21, and March 14, 2002.
    Brief description of amendments: The amendments revised the 
Technical Specifications to extend the completion times for the 
required actions associated with restoring an inoperable emergency 
diesel generator.
    Date of issuance: May 17, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 231 and 172.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: November 17, 2001 (66 
FR 52803). The supplements dated November 15, 2001, February 20 (two 
letters), February 21, and March 14, 2002, provided clarifying 
information that did not change the scope of the August 31, 2001, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 17, 2002.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch Nuclear 
Plant, Unit 2, Appling County, Georgia

    Date of application for amendment: May 21, 2001.
    Brief description of amendment: The amendment revised the Technical 
Specifications to eliminate the response time testing requirements for 
the reactor protection system signals of reactor high steam dome 
pressure and reactor vessel water level low.
    Date of issuance: May 17, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 173.
    Renewed Facility Operating License No. NPF-5: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31713). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 17, 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

[[Page 40031]]

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 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 email to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By July 11, 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

[[Page 40032]]

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).

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

    Date of application for amendment: May 29, 2002.
    Brief description of amendment: This amendment revises the 
Technical Specification (TS) 3/4.3.3.6 ``Accident Monitoring 
Instrumentation'' and associated Bases for Reactor Vessel Level and In 
Core Temperature monitoring to be consistent with NUREG-1431, Revision 
2, ``Standard Technical Specifications Westinghouse Plants.''
    Date of issuance: May 30, 2002.
    Effective date: May 30, 2002.
    Amendment No. 110.
    Facility Operating License No. NPF-63. Amendment revises the TS.
    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 May 30, 2002.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Thomas Koshy, Acting.

    Dated at Rockville, Maryland, this 3rd day of June, 2002.

    For the Nuclear Regulatory Commission.
Stuart A. Richards,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 02-14339 Filed 6-10-02; 8:45 am]
BILLING CODE 7590-01-P