[Federal Register Volume 67, Number 131 (Tuesday, July 9, 2002)]
[Notices]
[Pages 45560-45576]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-16956]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from June 14, 2002 through June 27, 2002. The 
last biweekly notice was published on June 25, 2002 (67 FR 42814).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. The filing of requests for a 
hearing and petitions for leave to intervene is discussed below.
    By July 25, 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

[[Page 45561]]

2.714, \1\ which is available at the Commission's PDR, located at One 
White Flint North, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System's (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing 
or petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
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    \1\ The most recent version of Title 10 of the CODE OF FEDERAL 
REGULATIONS, published January 1, 2002, inadvertently omitted the 
last sentence of 10 CFR 2.714(d) and subparagraphs (d)(1) and (2), 
regarding petitions to intervene and contentions. Those provisions 
are extant and still applicable to petitions to intervene. Those 
provisions are as follows: ``In all other circumstances, such ruling 
body or officer shall, in ruling on--
    (1) A petition for leave to intervene or a request for hearing, 
consider the following factors, among other things:
    (i) The nature of the petitioner's right under the Act to be 
made a party to the proceeding.
    (ii) The nature and extent of the petitioner's property, 
financial, or other interest in the proceeding.
    (iii) The possible effect of any order that may be entered in 
the proceeding on the petitioner's interest .
    (2) The admissibility of a contention, refuse to admit a 
contention if:
    (i) The contention and supporting material fail to satisfy the 
requirements of paragraph (b)(2) of this section; or
    (ii) The contention, if proven, would be of no consequence in 
the proceeding because it would not entitle petitioner to relief.''
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    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland, by the above date. Because of continuing 
disruptions in delivery of mail to United States Government offices, it 
is requested that petitions for leave to intervene and requests for 
hearing be transmitted to the Secretary of the Commission either by 
means of facsimile transmission to 301-415-1101 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
and because of continuing disruptions in delivery of mail to United 
States Government offices, it is requested that copies be transmitted 
either by means of facsimile transmission to 301-415-3725 or by e-mail 
to [email protected]. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the attorney for 
the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. Publicly available records 
will be accessible from the Agencywide Documents Access and Management 
System's (ADAMS) Public Electronic Reading Room on the Internet at the 
NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not 
have access to ADAMS or if there are problems in accessing the 
documents located in ADAMS, contact the NRC PDR Reference staff at 1-
800-397-4209, 304-415-4737 or by e-mail to [email protected].

[[Page 45562]]

Consumers Energy Company, Docket No. 50-155, Big Rock Point Nuclear 
Plant, Charlevoix, County, Michigan

    Date of amendment request: June 11, 2002.
    Description of amendment request: The amendment request changes the 
Defueled Technical Specifications by adding applicability statements to 
the requirements for storage and inspection of spent fuel and for the 
program requirements for spent fuel pool water chemistry.
    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:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The requested license amendment involves the addition of 
applicability statements to the program and activity requirements 
for the storage and inspection of spent fuel activities and 
requirements and the SFP [spent fuel pool] water chemistry. These 
applicability statements make requirements applicable whenever 
irradiated fuel is stored in the SFP. Once irradiated fuel has been 
completely removed from the SFP and transferred to a certified dry 
fuel storage container under a general 10 CFR Part 72 license, these 
program requirements for the SFP are no longer necessary. The 
program requirements consist of the specification, establishment, 
implementation, and maintenance of fuel configuration, fuel cooling, 
and water chemistry for the SFP to minimize the potential effects of 
decay heat and corrosion.
    The corresponding program requirements for fuel storage in dry 
containers are specified in the container's certificate of 
conformance and safety analysis report. The corresponding program 
requirements currently include:
    1. Analysis of fuel assemblies to determine maximum temperatures 
within the fuel assemblies to the temperature at the edge of the 
assemblies,
    2. Design of passive heat removal components to remove heat via 
convection, conduction, and radiation, and
    3. Specifications for canister vacuum drying pressure and helium 
backfill pressure that would ensure that a sufficiently inert 
environment is produced within the canister to inhibit corrosion.
    The program requirements associated with fuel storage in the SFP 
do not contribute to accident prevention or mitigation following the 
complete removal of irradiated fuel. The corresponding program 
features for fuel storage in dry storage containers are specified 
and containers are specified and controlled under other applicable 
license documents. These changes do not significantly increase the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any other accident previously evaluated.
    The requested amendment involves the addition of applicability 
statements that will have the effect of making a program requirement 
associated with the SFP inapplicable when the SFP is no longer used 
for irradiated fuel storage. The corresponding program requirements 
are adequately specified in applicable license documents. The 
elimination of this program requirement following complete removal 
of irradiated fuel from the SFP does not result in any new or 
different accident initiators from those already assumed in 
accidents previously evaluated, nor does it exacerbate any such 
accidents. Therefore, these changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The safety margins produced as a result of the specification of 
program requirements for fuel storage in the SFP are adequately 
maintained in corresponding program requirements associated with 
fuel storage in dry storage containers. These corresponding program 
requirements are specified in the dry storage container's 
certificate of compliance and safety analysis report. Therefore, 
this change does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's significant hazards 
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: David A. Mikelonis, Esquire, Consumers 
Energy Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: Robert A. Gramm.

Dominion Nuclear Connecticut, Inc., et al., Docket Nos. 50-336 and 50-
423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London 
County, Connecticut

    Date of amendment request: May 13, 2002.
    Description of amendment request: The proposed amendment modifies 
the Millstone Nuclear Power Station, Unit No. 2 (MP2) and Unit No. 3 
(MP3) Technical Specifications (TSs) to change selected MP2 and MP3 
radiological-related TSs. These changes are due to the revision to Part 
20 of Title 10 of the Code of Federal Regulations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. 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.
    These changes do not have an impact on the acceptance criteria for 
any design-basis accident described in the respective MP2 or MP3 
Updated Final Safety Analysis Report (UFSAR).
    The changes have no impact on plant equipment operation. Since the 
changes are administrative or editorial in nature they cannot affect 
the likelihood or consequences of accidents. Therefore, the proposed 
changes will not increase the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The revisions to the Occupational Radiation Exposure Report, 
Radioactive Effluent Controls Program, and High Radiation Area 
Specifications in accordance with TSTF travelers 152, 258, and 308 will 
have no effect on plant operation. Since the proposed changes are 
solely administrative or editorial in nature, they do not affect plant 
operation in any way.
    The proposed changes do not involve a physical alteration of the 
plant or change the plant configuration (no new or different type of 
equipment will be installed). The proposed changes do not require any 
new or unusual operator actions. The changes do not alter the way any 
structure, system, or component functions and do not alter the manner 
in which the plant is operated. The changes do not introduce any new 
failure modes. Therefore, the proposed changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    Since the proposed changes are solely administrative or editorial 
changes to the TSs, they do not affect plant operation in any way. The 
proposed changes to each unit's TSs will revise them to reflect the 
requirements of the current 10 CFR Part 20, standardize terminology, 
provide clearer guidance, clarify inconsistencies, remove extraneous 
information, and result in minor format changes that will not result in 
any technical changes to current requirements.
    The proposed changes have no effect on any safety analyses 
assumptions and therefore do not impact any margins of safety. The 
proposed changes do not impact any acceptance criteria for the design-
basis accidents described in the

[[Page 45563]]

respective MP2 or MP3 UFSAR and do not impact the consequences of 
accidents previously evaluated. Therefore, the proposed changes will 
not result in a reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, CT 06385.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: May 29, 2002.
    Description of amendment request: The amendments would revise the 
Technical Specifications 5.5.2 to allow, on a one-time basis, extension 
of the interval governing the conduct of containment integrated leak 
rate test (ILRT) from ten to fifteen years. The amendments represent a 
one-time exception to the ten-year frequency of the performance-based 
Type A tests as delineated by Regulatory Guide 1.163, ``Performance-
Based Containment Leak-Test Program,'' September 1995. The amendments 
will allow conduct of each respective unit's ILRT within fifteen years 
from the last ILRT performed for each unit.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The following discussion is a summary of the evaluation of the 
changes contained in these proposed amendments against the 10 CFR 
50.92(c) requirements to demonstrate that all three standards are 
satisfied. A no significant hazards consideration is indicated if 
operation of the facility in accordance with the proposed amendments 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated, or
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. Involve a significant reduction in a margin of safety.

First Standard

    The proposed amendments will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed extension to the Type A testing intervals 
cannot increase the probability of an accident previously evaluated 
since extension of the intervals is not a physical plant 
modification that could alter the probability of accident 
occurrence, nor is it an activity or modification by itself that 
could lead to equipment failure or accident initiation. The proposed 
extension to the Type A testing intervals does not result in a 
significant increase in the consequences of an accident as 
documented in NUREG-1493. The NUREG notes that very few potential 
containment leakage paths are not identified by Type B and Type C 
tests. It concludes that reducing the Type A testing frequency to 
once per twenty years leads to an imperceptible increase in risk.
    Catawba and McGuire provide a high degree of assurance through 
testing and inspection that the containments will not degrade in a 
manner detectable only by Type A testing. Recent Type A tests for 
the Catawba and McGuire units identified containment leakage within 
acceptance criteria, indicating a very leak tight containment. 
Inspections required by the ASME Code are also performed in order to 
identify indications of containment degradation that could affect 
leak tightness. Separately, Type B and Type C testing, required by 
TS [Technical Specifications], identify any containment opening from 
design penetrations, such as valves, that would otherwise be 
detected by a Type A test. These factors establish that an extension 
to the Type A test intervals will not represent a significant 
increase in the consequences of an accident.

Second Standard

    The proposed amendments will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed revisions to the Catawba and McGuire TS add 
a one-time extension to the current interval for Type A testing. The 
current test interval of ten years, based on past performance, would 
be extended on a one-time basis to fifteen years from the last Type 
A test. The proposed extension to Type A test intervals does not 
create the possibility of a new or different type of accident since 
there are no physical changes being made to the plants and there are 
no changes to the operation of the plants that could introduce a new 
failure mode.

Third Standard

    The proposed amendments will not involve a significant reduction 
in a margin of safety. The proposed revisions to the Catawba and 
McGuire TS add a one-time extension to the current interval for Type 
A testing. The current test interval of ten years, based on past 
performance, would be extended on a one-time basis to fifteen years 
from the last Type A test. The proposed extension to Type A test 
intervals will not significantly reduce the margin of safety. The 
NUREG-1493 generic study of the effects of extending containment 
leakage testing intervals found that a twenty-year interval resulted 
in an imperceptable increase in risk to the public. NUREG-1493 found 
that, generically, the design containment leakage rate contributes 
about 0.1 percent of the overall risk and that decreasing the Type A 
testing frequency would have a minimal effect on this risk, since 95 
percent of the Type A detectable leakage paths would already be 
detected by Type B and Type C testing. Similar proposed changes have 
been previously reviewed and approved by the NRC, and they are 
applicable to Catawba and McGuire.
    Based upon the preceding discussion, Duke Energy Corporation has 
concluded that the proposed amendments do not involve a significant 
hazards consideration.

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

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: The proposed change will revise 
Appendix 3B and Section 6.2.1.2 of the Updated Safety Analysis Report 
pertaining to the method of analysis. The proposed change will replace 
the current vendor THREED code for room pressure-temperature analyses 
due to High Energy Line Breaks (HELB) with GOTHIC (Generation of 
Thermal-Hydraulic Information for Containments). The proposed change 
will allow Entergy Operations, Inc. (EOI) to update the analysis and to 
evaluate additional changes to the plant.
    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. Will the operation of the facility in accordance with these 
proposed changes involve a significant increase in the probability 
or consequence of an accident previously evaluated?
    Response: The proposed change involves no increase in the 
probability of the accidents previously evaluated since no physical 
change to the plant will be made. The change of the High Energy Line 
Break (HELB) analysis method does not affect the probability of the 
analyzed event occurring.

[[Page 45564]]

The line break locations have not been affected and remain as 
originally designed.
    This submittal is required due to the change of HELB analysis 
code from the vendor code THREED to the modern industry standard 
analysis code GOTHIC. This is a change in the methodology for 
determining the effects of the mass and energy release in the plant 
as a result of currently postulated events. The change in the 
evaluation methodology has been benchmarked and reviewed to confirm 
the results remain consistent with the current analysis. The changes 
to the model used for the additional analysis allow the use of new, 
more physically realistic models for Containment and Auxiliary 
Building pressure/temperature responses and will demonstrate 
continued qualification of the equipment in these buildings. Mass 
and energy releases for some cases have also been recalculated to 
credit pipe friction, which was only credited for certain cases 
previously.
    With these new results the equipment has been reviewed and 
remains qualified per current programs established at RBS [River 
Bend Station]. Therefore, the plant will continue to function as 
designed and thus there will be no impact on consequences.
    2. Will the operation of the facility in accordance with these 
proposed changes create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    Response: No physical change to the plant will be made. The HELB 
locations were identified by reviewing all the possible break 
locations in each Auxiliary and Containment Building volume 
containing high-energy lines. The locations of the breaks remain the 
same as the previous HELB analyses. The HELB analyses have been 
evaluated for the current plant configuration. The new HELB analysis 
has been benchmarked against the previous accepted methods and found 
to correlate with the previous analysis. Therefore the results can 
be used to predict plant responses to events. The proposed change 
uses improved methods for mass and energy release calculation and 
pressure / temperature responses to determine the EQ [equipment 
qualification] qualification envelopes. Therefore, no new or 
different interaction would be created.
    3. Will the operation of the facility in accordance with these 
proposed changes involve a significant reduction in a margin of 
safety?
    Response: The operation of the facility in accordance with the 
proposed changes will not involve a significant reduction in a 
margin of safety.
    The GOTHIC code has been successfully benchmarked versus the 
vendor THREED code, which was used in the original design 
calculations. The HELB analysis results with the benchmarking GOTHIC 
model are consistent with the THREED results. Therefore, the use of 
GOTHIC code will not involve a reduction in an identified margin of 
safety. Given that GOTHIC code is an improved methodology and it has 
been extensively qualified against the solved analytical problems 
and testing results, the use of GOTHIC code will produce more 
accurate pressure/temperature responses for the HELB analyses. The 
use of the GOTHIC code has been approved for pressure/temperature 
responses analysis at various other plants including Joseph M. 
Farley Nuclear Plant, Units 1 and 2, and Waterford [Steam Electric 
Station, Unit] 3.
    The results with the revised methods will be used to show that 
safety equipment meets the EQ requirements. The peak temperatures 
and pressures in the HELB GOTHIC benchmark model are within the 
existing EDC [environmental design criteria] envelopes. Therefore, 
the pressure/temperature responses from the HELB benchmark analyses 
have no impact on the equipment qualification.
    The methodology in the original design calculations is very 
conservative. The mass and energy releases without crediting 
friction introduce excessive amount of high-energy fluid into the 
break rooms, which is unrealistic. Some HELB calculations have 
credited both the frictional flows and the additional zone to 
eliminate excessive conservatism in the pressure/temperature 
responses. There is no reduction in a margin of safety and the 
design room differential pressure limits continue to be [met].
    The use of this method by EOI RBS is consistent with the 
guidance given in NRC [U.S. Nuclear Regulatory Commission] Generic 
Letter 83-11 and Supplement 1, addressing the performance of safety 
analyses by licensees. EOI has implemented this guidance for the 
GOTHIC methodology consistent with the intended application. The 
GOTHIC methodology has been verified and validated by the software 
vendor. In addition, this methodology is controlled by EOI 
procedures and under the EOI quality assurance program. This 
includes EOI and RBS specific verification and validation of this 
application of GOTHIC and review of the calculations performed.

    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 Nuclear Operations, Inc. (ENO), Docket No. 50-003, Indian Point 
Nuclear Generating Station, Unit 1, Buchanan, New York

    Date of application for amendment: May 30, 2002.
    Description of amendment request: The proposed changes will modify 
the Indian Point Generating Station, Unit 1 (IP1), Technical 
Specifications (TSs) and Provisional Operating License No. DPR-5. IP1 
is completely enclosed within the protected area for Indian Point 
Nuclear Generating Station, Unit 2 (IP2). IP1 depends on the IP2 TSs 
and processes for the implementation of certain regulatory 
requirements. The requested changes will simplify the IP1 TSs to 
facilitate the IP2 transition to the Improved TSs. The IP1 TSs will be 
reformatted, reordered and repaginated for consistency and clarity. ENO 
also proposes that certain changes supersede requirements of the 
``Order Approving Decommissioning Plan and Authorizing Decommissioning 
of Facility'' \2\ (the Order) to ensure compliance with the current 
requirements of 10 CFR Part 50.59, ``Changes, tests, and experiments.'' 
and 10 CFR Part 50.82, ``Termination of license,'' for evaluating 
whether changes can be made to IP1 without NRC approval.
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    \2\ U.S. Nuclear Regulatory Commission (NRC) letter to 
Consolidated Edison, ``Order to Authorize Decommissioning and 
Amendment No. 45 to License No. DPR-5 for Indian Point Unit 1 (TAC 
No. M59664),'' dated January 31, 1996.
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    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    The NSB [Nuclear Services Building] sewage effluent line 
radiation monitor is not required to function to mitigate any 
postulated accident. The design or operation of the radiation 
monitor on the existing sewage effluent discharge line will not be 
changed by deleting operability and surveillance requirements for 
the NSB sewage effluent radiation monitor from the IP1 TS. The 
nuclear services building sewage effluent line is neither an 
accident initiator nor mitigator.
    The other proposed changes do not result in a change to the 
design or operation of any plant structure, system or component. 
Therefore any assumptions of the operability or performance of any 
structure, system or component in accident evaluations are 
unchanged.
    The proposed fire protection TS 2.11 involves deleting 
requirements from the IP1 TS that are solely applicable to IP2. Any 
assumptions of the operability or performance of any structure, 
system or component in IP2 accident evaluations, including the Fire 
Plan, are unchanged. Therefore, there is no increase in the 
probability or in the consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    The proposed TS change involves the deletion of operability and 
surveillance requirements for radioactive effluent monitoring of the 
NSB sewage effluent from the IP1 TS. The proposed TS changes do not

[[Page 45565]]

affect the design or operation of any plant structure, system, or 
component.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    This change to TS 1.0 does not affect a design function for or 
the operation of any plant structure, system, or component. The 
change does not affect the method of ENO's compliance with any 
regulation.
    The proposed TS change involving IP1 TS 2.11 statement governs 
the protection of IP2 safe shutdown systems from fire. Effective 
protection of IP2 safe shutdown systems from fire is mandated by IP2 
License Condition 2.K. The effectiveness of ENO compliance with IP2 
License Condition 2.K is not affected by this change. In addition, 
this change does not affect a design function or the operation of 
any plant structure, system, or component.
    The proposed changes to TS sections 3.1 and 3.2 involve 
eliminating the duplication of requirements in the IP1 TS and 
incorporating the requirements by reference to the IP2 TS. A single 
ENO organization operates both IP1 and IP2. The effective 
organizational requirements to ensure compliance with all ENO IP1 
and IP2 site requirements are mandated by the IP2 TS. The 
effectiveness of ENO's safety management of the Indian Point site is 
not affected by this change. In addition, this change does not 
affect a design function or the operation of any plant structure, 
system, or component.
    The proposed TS change to sections 4.1 and 5.2 involves 
eliminating the reference in the IP1 TS to the specific applicable 
section number of the IP2 TS. A single organization operates both 
IP1 and IP2. The applicable IP2 TS is obvious by the activity title. 
The effectiveness of ENO's safety management of the Indian Point 
site is not affected by this change. In addition, this change does 
not affect a design function or the operation of any plant 
structure, system, or component.
    Effective compliance with the 10CFR20 requirements for radiation 
protection and monitoring radioactive effluent releases is mandated 
by other IP1 and IP2 TS and license provisions. The effectiveness of 
ENO compliance with 10CFR20 requirements is not adversely affected 
by the elimination of TS requirements for the radiation protection 
plan and radioactive effluent monitoring on the nuclear services 
building sewage effluent line.
    The proposed TS change involves requirements for the site 
Meteorological Monitoring and Radiological Environmental Monitoring 
programs. However, IP2 TS provisions mandate effective compliance 
for meteorological and radiological environmental monitoring. The 
effectiveness of ENO compliance with 10CFR50.47, 10CFR100, and 
10CFR20 requirements is not adversely affected by this change. In 
addition, this change does not affect a design function or the 
operation of any plant structure, system, or component. IP2 TS 
provisions mandate effective compliance with requirements for 
radiation protection. The effectiveness of ENO's compliance with 10 
CFR 20 is not adversely affected by this change or the change to the 
section for sealed sources. In addition, this change does not affect 
a design function or the operation of any plant structure, system, 
or component.
    The proposed TS change involves the location of routine and 
event reporting requirements. However, other IP2 TS provisions 
mandate effective compliance with reporting requirements. In 
addition, this change does not affect a design function or the 
operation of any plant structure, system, or component.
    The effectiveness of ENO's compliance with 10CFR50.59 is not 
adversely affected by the clarification and relocation of the 
applicability of the FSAR [Final Safety Analysis Report]. In 
addition, this change does not affect a design function or the 
operation of any plant structure, system, or component.
    Therefore, the change does not result in a change to any of the 
safety analyses or any margin of safety.

ENO also requests that the expiration date of IP1 Provisional Operating 
License No. DPR-5 be changed from ``midnight, October 14, 2002,'' to 
``midnight, September 28, 2013,'' the current expiration date for 
Facility Operating License No. DPR-26 for IP2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    In its Safety Evaluation and Environmental Assessment for its 
January 31, 1996, Order Approving Decommissioning Plan and 
Authorizing Decommissioning of Facility, the NRC evaluated the 
acceptability of the possession-only license and safety issues 
related to SAFSTOR of Indian Point Nuclear Generating Unit No. 1 
until September 28, 2013. The requested change does not involve any 
activity that could change the assumptions of the prior Safety 
Evaluation and Environmental Assessment.
    Therefore, the proposed license amendment does not involve a 
significant increase in the probability or in the consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    In its Safety Evaluation and Environmental Assessment for its 
January 31, 1996, Order Approving Decommissioning Plan and 
Authorizing Decommissioning of Facility, the NRC evaluated the 
acceptability of the possession-only license and safety issues 
related to SAFSTOR of Indian Point Nuclear Generating Unit No. 1 
until September 28, 2013. The requested change does not involve any 
activity that could change the assumptions of the prior Safety 
Evaluation and Environmental Assessment.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    In its Safety Evaluation and Environmental Assessment for its 
January 31, 1996, Order Approving Decommissioning Plan and 
Authorizing Decommissioning of Facility, the NRC evaluated the 
acceptability of the possession-only license and safety issues 
related to SAFSTOR of Indian Point Nuclear Generating Unit No. 1 
until September 28, 2013. The requested change does not involve any 
activity that could change the assumptions of the prior Safety 
Evaluation and Environmental Assessment.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: May 30, 2002.
    Description of amendment request: The proposed amendment would 
revise the Facility Operating License and Technical Specifications 
(TSs) to increase the licensed core thermal power level to 3067.4 
megawatts (MWt), which is a 1.4% increase above the currently 
authorized power level of 3025 MWt. The proposed power uprate involves 
the improvement in the core power uncertainty allowance originally 
required for the emergency core cooling system (ECCS) evaluations 
performed in accordance with Appendix K, ``ECCS Evaluation Models,'' to 
Part 50 of Title 10 of the Code of Federal Regulations. In addition, 
changes would be made in TS Sections 2.2, 3.3, 3.4, 3.7, and the 
applicable TS Bases would be revised to account for the change in power 
level.
    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:


[[Page 45566]]


    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The evaluations associated with this proposed change to core 
power level have demonstrated that all applicable acceptance 
criteria for plant systems, components, and analyses (including the 
Final Safety Analysis Report Chapter 14 safety analyses) will 
continue to be met for the proposed 1.4% increase in licensed core 
thermal power for IP3 [Indian Point Unit 3]. The subject increase in 
core thermal power will not result in conditions that could 
adversely affect the integrity (material, design, and construction 
standards) or the operational performance of any potentially 
affected system, component or analysis. Therefore, the probability 
of an accident previously evaluated is not affected by this change. 
The subject increase in core thermal power will not adversely affect 
the ability of any safety-related system to meet its intended safety 
function. Further, the radiological dose evaluations in support of 
this power uprate effort show that the current FSAR [Final Safety 
Analysis Report] Chapter 14 radiological analyses are unaffected, 
and that the current dose analyses of record bound plant operation 
with the subject increase in licensed core thermal power level.
    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 evaluations of this proposed amendment show that all 
applicable acceptance criteria for plant systems, components, and 
analyses (including FSAR Chapter 14 safety analyses) will continue 
to be met for the proposed 1.4% power increase in IP3 licensed core 
thermal power. The subject increase in core thermal power will not 
result in conditions that could adversely affect the integrity 
(material, design, and construction standards) or operational 
performance of any potentially affected system, component, or 
analyses. The subject increase in core thermal power will not 
adversely affect the ability of any safety-related system to meet 
its safety function. Furthermore, the conditions associated with the 
subject increase in core thermal power will neither cause initiation 
of any accident, nor create any new credible limiting single 
failure. The power uprate does not result in changing the status of 
events previously deemed to be non-credible being made credible. 
Additionally, no new operating modes are proposed for the plant as a 
result of this requested change.
    Therefore, the subject increase in core thermal power level will 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The evaluations associated with this proposed change show that 
all applicable acceptance criteria for plant systems, components, 
and analyses (including FSAR Chapter 14 safety analyses) will 
continue to be met for this proposed 1.4% increase in IP3 licensed 
core thermal power. The subject increase in core thermal power will 
not result in conditions that could adversely affect the integrity 
(material, design, and construction standards) or operational 
performance of any potentially affected system, component, or 
analysis. The subject power uprate will not adversely affect the 
ability of any safety-related system to meet its intended safety 
function. For example, most IP3 analyses already add a 2% 
uncertainty allowance to the nominal power level to account solely 
for power measurement uncertainty. These analyses have not been 
revised for the 1.4% uprate power level conditions because the sum 
of increased core power level (1.4%) and the improved power 
measurement accuracy (uncertainty less than 0.6%) is already bounded 
by the currently analyzed 2% uncertainty allowance.
    Therefore, the subject increase in core thermal power will 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: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: June 3, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.4.9, ``Pressurizer,'' to increase 
the pressurizer water level limit when the plant is in Mode 3 (Hot 
Standby). The current pressurizer water level limit is applicable for 
Modes 1, 2, and 3, and will remain unchanged for Modes 1 and 2. The 
proposed amendment would also revise TS 3.8.4, ``DC Sources--
Operating,'' to remove the notes that refer to the one-time amendment 
allowing the online replacement of station batteries 31 and 32. The 
notes are no longer applicable since the batteries have been replaced.
    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.
    Pressurizer water level is an assumed initial condition for 
certain accident analyses. Plant initial conditions are not accident 
initiators and do not have an effect on the probability of the 
accident occurring. The proposed change only revises the specified 
limit on water level in the pressurizer, so that this change would 
not affect accident probability.
    The specific accidents for which pressurizer water level is an 
assumed initial condition are a loss of load and a loss of normal 
feedwater. The limiting accident analysis results occur at full 
power conditions when the available core thermal power is maximized. 
The proposed change does not affect the specified pressurizer level 
limit at any power level from zero to full power. That is, the 
pressurizer level limit is not being changed in Modes 1 and 2. The 
proposed change does revise the specified pressurizer water level 
limit in Mode 3 (Hot Standby) but this does not affect accident 
analysis results because the limiting analyses will remain those 
that are postulated to occur in Mode 1 with the plant at full power.
    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 change does not involve physical changes to 
existing plant equipment or the installation of any new equipment. 
The design of the pressurizer, the pressurizer level control system 
and the pressurizer safety valves is not being changed and the 
ability of these systems, structures, and components to perform 
their design or safety functions is not being affected. The proposed 
change revises the specified limit on pressurizer water level in 
Mode 3 (Hot Standby) to allow operators greater flexibility in 
performing a plant cooldown. The method used in performing the plant 
cooldown is not being changed. This proposed change does not create 
new failure modes or malfunctions of plant equipment nor is there a 
new credible failure mechanism.
    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.
    Pressurizer level is an initial condition assumed in certain 
accident analyses involving an insurge in the pressurizer and an 
increasing reactor coolant system (RCS) pressure. These analyses 
demonstrate that the design pressure for the RCS is not exceeded for 
the limiting analyses based on the plant at full power. The proposed 
change does not affect the existing Technical

[[Page 45567]]

Specification requirement for Mode 1 (Power Operation) or Mode 2 
(Plant Startup) and therefore does not affect the assumptions or 
results of these accident analyses. The margin for RCS design 
pressure demonstrated by these analysis results is not being 
reduced. The proposed change only applies to the pressurizer level 
limit in Mode 3 (Hot Standby) when there is substantially lower 
thermal energy available to cause rapid expansion of reactor coolant 
and an insurge to the pressurizer. Protection of the RCS pressure 
boundary is still maintained by the pressurizer safety valves, which 
are not being modified by the proposed change in pressurizer water 
level.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: June 5, 2002.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to implement the alternate 
source term methodology for the fuel-handling accident analysis. 
Specifically, the proposed amendment would revise TS 3.9.3, 
``Containment Penetrations,'' to: (1) Permit the equipment hatch 
opening and the personnel air lock doors to be capable of being closed 
during movement of irradiated fuel, (2) allow use of administrative 
controls for unisolating containment penetrations during movement of 
irradiated fuel, (3) delete the containment purge and containment 
pressure relief requirements and associated surveillances with the 
reactor subcritical for less than 550 hours, and (4) eliminate the TS 
applicability ``during core alterations.'' In this regard, the proposed 
amendment would adopt TS Task Force (TSTF) Standard TS Change Travelers 
TSTF-68, ``Containment Personnel Airlock Doors Open During Fuel 
Movement,'' TSTF-312, ``Administratively Control Containment 
Penetrations,'' and, in part, TSTF-51, ``Revise Containment 
Requirements During Handling Irradiated Fuel and Core Alterations.'' 
The proposed amendment would also relocate the requirements in TS 
3.7.13, ``Fuel Storage Building Emergency Ventilation System,'' and TS 
3.3.8, ``Fuel Storage Building Emergency Ventilation System Actuation 
Instrumentation,'' to the licensee-controlled Technical Requirements 
Manual.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change involves the reanalysis of a fuel handing 
accident (FHA) in containment and in the fuel storage building. The 
new analysis, based on the Alternate Source Term (AST) in accordance 
with 10 CFR [Code of Federal Regulations] 50.67, will replace the 
existing analysis based on methodologies and acceptance criteria in 
place when Indian Point 3 was originally licensed. As a result of 
the new analysis, changes to the Technical Specifications are 
proposed which take credit for the new analysis results.
    The proposed changes to the technical specifications modify 
requirements regarding containment closure during movement of 
irradiated fuel assemblies in containment and relocate requirements 
for the fuel storage building emergency ventilation system from the 
technical specifications to a licensee controlled document. The 
proposed changes do not involve physical modifications to plant 
equipment and do not change the operational methods or procedures 
used for moving irradiated fuel assemblies. As such, there are no 
accident initiators affected by the proposed amendment. The revised 
requirements apply only when the plant is in a refueling condition 
(Mode 6), and specifically only when irradiated fuel is being moved. 
Previously evaluated accidents with the plant in other conditions 
ranging from cold shutdown (Mode 5) through power operation (Mode 1) 
are not affected. The AST methodology is used to evaluate a[n] FHA 
that is postulated to occur during fuel movement activities in the 
containment building and the fuel storage building. The analysis 
follows the guidance of the NRC Regulatory Guide 1.183 and uses the 
acceptance criteria of the NRC Standard Review Plan (NUREG 0800) for 
offsite doses and General Design Criteria 19 for control room 
personnel. The analysis demonstrates that the dose consequences meet 
regulatory acceptance criteria. The accident analysis conservatively 
assumes that the containment building and the fuel storage building, 
including ventilation filtration systems for those building[s] does 
not diminish or delay the assumed fission product release. The 
analysis does take credit for, and technical specifications enforce, 
the presence of 23 feet of water over the irradiated fuel while fuel 
movement activities are being performed. The analysis also takes 
credit for, and the technical specification bases enforce a fuel 
decay time of at least 84 hours. In addition, administrative 
controls are put in place to provide for closure of containment 
openings in the event of a[n] FHA. Use of an alternate analysis 
method does not affect fuel parameters or the equipment used to 
handle the fuel. The proposed changes to the technical 
specifications reflect assumptions made in the analysis.
    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 amendment involves the use of an alternate analysis 
methodology for the evaluation of the dose consequences from a[n] 
FHA that is postulated to occur in either the containment building 
or the fuel storage building (FSB). The analysis demonstrates that 
containment closure conditions and operation of the containment 
purge filtration system are not required to maintain dose 
consequence within regulatory limits following a postulated FHA in 
containment. Therefore the new analysis supports proposed changes to 
requirements for containment closure during movement of irradiated 
fuel assemblies in containment. The analysis results also 
demonstrate that operation of the fuel storage building emergency 
ventilation system is not required to maintain dose consequences 
within regulatory limits following a postulated FHA in the FSB. The 
containment closure components (e.g., equipment hatch, personnel 
airlock doors, and various containment penetrations) and filtration 
systems are not accident initiators. The proposed changes do not 
involve the addition of new systems or components nor do they 
involve the modification of existing plant systems. The proposed 
changes do not affect the way in which a[n] FHA is postulated to 
occur.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The existing dose analysis methodology and assumptions 
demonstrates that the dose consequences of a[n] FHA are within 
regulatory limits for whole body and thyroid doses as established in 
10 CFR 100. The alternate dose analysis methodology and assumptions 
also demonstrates that the dose consequences of a[n] FHA are within 
regulatory limits. The limits applicable to the alternate analysis 
are established in 10 CFR 50.67 in conjunction with the TEDE (total 
effective dose equivalent) acceptance directed in Regulatory Guide 
1.183. The acceptance criteria for both dose analysis methods have 
been developed for the

[[Page 45568]]

purpose of evaluating design basis accidents to demonstrate adequate 
protection of public health and safety. An acceptable margin of 
safety is inherent in both types of acceptance criteria.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: June 7, 2002.
    Description of amendment request: The proposed amendment would 
change the requirements associated with handling irradiated fuel and 
performing core alterations. Specifically, the changes would eliminate 
operability requirements for secondary containment when handling 
recently irradiated fuel and during core alterations. The amendment 
would also revise the requirements associated with equipment whose 
performance is not credited in the new calculations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does not involve a significant increase in the probability or 
consequences of an accident previously analyzed?
    Response: No.
    The proposed TS [Technical Specifications] changes do not modify 
the design or operation of equipment used to move spent fuel or to 
perform core alterations. Because the equipment affected by the 
change is not an initiator to any previously analyzed accident, the 
proposed change cannot increase the probability of any previously 
analyzed accident.
    The conservative re-analysis of the fuel handling accident 
concludes that radiological consequences are within the acceptance 
criteria in Regulatory Guide 1.183 and 10 CFR 50.67. The results of 
the core alteration events, other than the fuel handling accident, 
remain unchanged from the original design-basis, which showed that 
these events do not result in fuel cladding damage or radioactive 
release. The radiological analysis uses the same FHA [fuel handling 
accident] source activity previously accepted in the design-basis 
FHA analysis. The same source activity is used with the guidance in 
the Regulatory Guide 1.183, Appendix B and the passive release/
transport path, which does not take the dose mitigation credit of 
engineered safeguards including secondary containment and CREVAS 
[Control Room Emergency Ventilation] Systems.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability of occurrence or 
consequences of an accident previously analyzed.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously analyzed?
    Response: No.
    The proposed post-FHA activity transport path is passive in 
nature and it does not take the credit of dose mitigation functions 
previously credited in the design-basis FHA analysis. The proposed 
changes do not introduce any new modes of plant operation and do not 
involve physical modifications to the plant.
    Therefore, this proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously analyzed.
    3. Does not involve a significant reduction in [a] margin of 
safety?
    Response: No.
    The proposed changes revise the FitzPatrick TS to establish 
operational conditions where specific activities represent 
situations during which significant radioactive releases can be 
postulated. These new operational conditions are consistent with the 
proposed design-basis accident analysis and are established such 
that the radiological consequences are less than the regulatory 
allowable limits. Safety margins and analytical conservatisms are 
retained to ensure that the analysis adequately bounds all 
postulated event scenarios. The selected assumptions and release 
models provide an appropriate and prudent safety margin against 
unpredicted events in the course of an accident and compensates for 
large uncertainties in facility parameters, accident progression, 
radioactive material transport and atmospheric dispersion. The 
proposed TS applicability statements continue to ensure that the 
TEDE [Total Effective Dose Equivalent] at the control room and the 
exclusion area and low population zone boundaries are below the 
corresponding regulatory allowable limits in 10 CFR 50.67(b)(2).
    Therefore, these changes do not involve a significant reduction 
in [a] margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: Richard J. Laufer.

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

    Date of application for amendments: May 24, 2002.
    Description of amendment request: Exelon Generation Company, LLC, 
the licensee, is proposing changes to the Peach Bottom Atomic Power 
Station, Units 2 and 3 (PBAPS), Operating Licenses and Technical 
Specifications associated with an increase in the licensed power level. 
The changes involve a proposed 1.62 percent increase in the licensed 
reactor core thermal power level (an increase in reactor power level 
from 3,458 megawatts thermal to 3,514 megawatts thermal). These changes 
result from increased accuracy of the feedwater flow and temperature 
measurements to be achieved by utilizing high accuracy ultrasonic flow 
measurement instrumentation. This results in a more accurate 
determination of reactor core thermal power level. 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 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 and found acceptable.
    The primary loop components (reactor vessel, reactor internals, 
control rod drive housings, piping and supports, recirculation 
pumps, etc.) continue to comply with their applicable structural 
limits and will continue to perform their intended design functions. 
Thus, there is no increase in the probability of a structural 
failure of these components.
    All of the [Nuclear Steam Supply System] NSSS systems will still 
perform their intended design functions during normal and accident 
conditions. The balance of plant [(BOP)] systems and components 
continue to

[[Page 45569]]

meet their applicable structural limits and will continue to perform 
their intended design functions. Thus, there is no increase in the 
probability of a structural failure of these components. All of the 
NSSS/BOP interface systems will continue to perform their intended 
design functions. The safety relief valves and containment isolation 
valves meet design sizing requirements at the uprated power level.
    Because the integrity of the plant will not be affected by 
operation at the uprated condition, 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 core thermal power uncertainty 
measurement allows most of the current safety analyses to be used, 
with small changes to the core operating limits, to support 
operation at a core power of 3514 megawatts thermal (MWt). Other 
analyses performed at a nominal power level have either been 
evaluated or re-performed for the 1.62% increased power level. The 
results demonstrate that the applicable analysis acceptance criteria 
continue to be met at the 1.62% uprate conditions. As such, all 
PBAPS Updated Final Safety Analysis Report (UFSAR) Chapter 14 
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 bound operation at the 1.62% uprated 
condition.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. No new accident scenarios, failure mechanisms, or 
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. As 
appropriate, all evaluations have been performed using methods that 
have either been reviewed and approved by the NRC, or that are in 
compliance with regulatory review guidance and standards.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

Exelon Generation Company, LLC, Docket No. 50-254, Quad Cities Nuclear 
Power Station, Unit 1, Rock Island County, Illinois

    Date of amendment request: May 30, 2002.
    Description of amendment request: The proposed change revises the 
safety limit minimum critical power ratio for Unit 1 Cycle 18 for two 
loop operation and single loop operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. Limits have been established consistent with NRC 
approved methods to ensure that fuel performance during normal, 
transient, and accident conditions is acceptable. The proposed 
change conservatively establishes the safety limit for the minimum 
critical power ratio (SLMCPR) for Quad Cities Nuclear Power Station 
(QCNPS), Unit 1, Cycle 18 such that the fuel is protected during 
normal operation and during any plant transients or anticipated 
operational occurrences.
    Changing the SLMCPR does not increase the probability of an 
evaluated accident. The change does not require any physical plant 
modifications, physically affect any plant components, or entail 
changes in plant operation. Therefore, no individual precursors of 
an accident are affected.
    The proposed change revises the SLMCPR to protect the fuel 
during normal operation as well as during any transients or 
anticipated operational occurrences. Operational limits will be 
established based on the proposed SLMCPR to ensure that the SLMCPR 
is not violated during all modes of operation. This will ensure that 
the fuel design safety criteria (i.e., that at least 99.9% of the 
fuel rods do not experience transition boiling during normal 
operation and anticipated operational occurrences) is met. Since the 
operability of plant systems designed to mitigate any consequences 
of accidents has not changed, the consequences of an accident 
previously evaluated are not expected to increase.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated? 
Creation of the possibility of a new or different kind of accident 
would require the creation of one or more new precursors of that 
accident. New accident precursors may be created by modifications of 
the plant configuration, including changes in allowable modes of 
operation. The proposed change does not involve any modifications of 
the plant configuration or allowable modes of operation. The 
proposed change to the SLMCPR assures that safety criteria are 
maintained for QCNPS, Unit 1, Cycle 18.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    Does the proposed change involve a significant reduction in a 
margin of safety?
    The value of the proposed SLMCPR provides a margin of safety by 
ensuring that no more than 0.1% of the rods are expected to be in 
boiling transition if the MCPR limit is not violated. The proposed 
change will ensure the appropriate level of fuel protection. 
Additionally, operational limits will be established based on the 
proposed SLMCPR to ensure that the SLMCPR is not violated during all 
modes of operation.
    This will ensure that the fuel design safety criteria (i.e., 
that at least 99.9% of the fuel rods do not experience transition 
boiling during normal operation as well as anticipated operational 
occurrences) are met.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

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

[[Page 45570]]

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

    Date of amendment request: June 13, 2002.
    Description of amendment request: The amendment would revise the 
Improved Technical Specifications (ITS) 3.3.8 and associated bases, 
``Emergency Diesel Generator (EDG) Loss of Power Start (LOPS),'' by 
changing the completion time for required action D.2 from 12 to 36 
hours. The amendment also corrects a typographical error in ITS 3.3.8 
and clarifies the discussion in Bases Section B 3.3.8 for Actions D.1 
and D.2 to recognize the applicability of ITS 3.3.8 in MODES 5 and 6.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does not involve a significant increase in the probability 
or consequences of an accident previously analyzed.
    The proposed license amendment revises the Required Time to 
place the plant in MODE 5 if an inoperable loss of voltage Function 
for the emergency diesel generator (EDG) loss of power start (LOPS) 
cannot be restored to OPERABLE status, corrects a typographical 
error in the Section Number of ITS 3.3.8, and clarifies the wording 
of ITS Bases Section B 3.3.8 for Action D.1 and D.2 regarding the 
applicability of the specification during MODES 5 and 6.
    The EDG LOPS is intended to protect engineered safeguards 
equipment from damage due to sustained undervoltage conditions, and 
to ensure rapid restoration of power to the engineered safeguards 
electrical buses in the event of a loss of offsite power. The EDG 
LOPS is not an initiator of any design basis accident. The design 
functions of the EDG LOPS and the initial conditions for accidents 
that require an EDG LOPS will not be affected by the change. 
Therefore, the change will not increase the probability or 
consequences of an accident previously evaluated.
    (2) Does not create the possibility of a new or different kind 
of accident from any accident previously analyzed.
    The proposed amendment involves no changes to the design 
functions or operation of the EDG LOPS. Editorial corrections, 
clarification of the wording in Bases Section B 3.3.8, or changing 
the Required Completion Time for placing the plant in MODE 5 when an 
inoperable loss of voltage function cannot be restored will not 
introduce any new failure mechanisms, malfunctions or accident 
initiators. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Does not involve a significant reduction in the margin of 
safety.
    The proposed change corrects a typographical error, clarifies 
the wording of Bases Section B 3.3.8 for Actions D.1 and D.2, and 
revises the required Completion Time to place the plant in MODE 5. 
The revised Completion Time will allow the plant to be shutdown in 
an orderly fashion without challenging plant systems or plant 
cooldown limits. The proposed change does not change the design or 
operation of the EDG LOPS, and does not impact the ability of the 
EDG LOPS to perform its design functions. Thus, the proposed 
amendment will not result in a reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: R. Alexander Glenn, Associate General 
Counsel (MAC-BT15A), Florida Power Corporation, P.O. Box 14042, St. 
Petersburg, Florida 33733-4042.
    NRC Acting Section Chief: Kahtan N. Jabbour.

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

    Date of amendment request: June 7, 2002.
    Description of amendment request: The proposed amendments would 
delete requirements from the Technical Specifications (TSs) (and, as 
applicable, other elements of the licensing bases) to maintain a Post 
Accident Sampling System (PASS). Licensees were generally required to 
implement PASS upgrades as described in NUREG-0737, ``Clarification of 
TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory 
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Assess Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades was an outcome of the 
lessons learned from the accident that occurred at TMI, Unit 2. 
Requirements related to PASS were imposed by Order for many facilities 
and were added to or included in the TSs for nuclear power reactors 
currently licensed to operate. However, lessons learned and 
improvements implemented over the last 20 years have shown that the 
information obtained from PASS can be readily obtained through other 
means, or is of little use in the assessment and mitigation of accident 
conditions.
    The Nuclear Regulatory Commission (NRC) staff issued a notice of 
opportunity for comment in the Federal Register on December 27, 2001 
(66 FR 66949) on possible amendments to eliminate PASS, including a 
model safety evaluation and model no significant hazards consideration 
(NSHC) determination, using the consolidated line item improvement 
process. The NRC staff subsequently issued a notice of availability of 
the models for referencing in license amendment applications in the 
Federal Register on March 20, 2002 (67 FR 13027). The licensee affirmed 
the applicability of the NSHC determination in its application dated 
June 7, 2002. The NSHC determination is restated below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:

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

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

[[Page 45571]]

consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
accident previously evaluated.

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

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

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

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

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

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of amendment request: May 29, 2002.
    Description of amendment request: The proposed amendment would 
revise TS 3.8.1, ``AC Sources--Operating,'' to allow portions of 
Surveillance Requirement (SR) 3.8.1.5 to be performed with the units in 
Mode 1, 2, 3 or 4. This proposed amendment is consistent with changes 
made to NUREG-1431, Standard Technical Specifications, Westinghouse 
Plants, by Technical Specification Task Force (TSTF) Traveler, TSTF-
283, Revision 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    The standby emergency power sources are primarily a support 
system for systems required to be operable for accident mitigation. 
SR 3.8.1.5 demonstrates the standby emergency power source 
operation, during a loss of offsite power actuation test signal in 
conjunction with an Engineering Safeguards Feature (ESF) actuation 
signal. The proposed amendment only changes the allowed operating 
Modes in which portions of this surveillance may be performed. 
Performing portions of the surveillance in Mode 1, 2, 3, or 4 will 
require an assessment to determine that plant safety is maintained 
or will be enhanced.
    Therefore, the consequences of an accident previously evaluated 
will not be significantly increased as a result of the proposed 
change.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a new or different kind 
of accident from any accident previously evaluated.
    The possibility for a new or different type of accident from any 
accident previously evaluated is not created as a result of this 
amendment. These changes do not introduce any new or different 
normal operation or accident initiators. Performing the surveillance 
in Mode 1, 2, 3, or 4 will require an assessment to determine that 
plant safety is maintained or will be enhanced.
    Equipment important to safety will continue to operate as 
designed. The changes do not result in any event previously deemed 
incredible being made credible. The changes do not result in more 
adverse conditions or result in any increase in the challenges to 
safety systems. Therefore, operation of the Point Beach Nuclear 
Plant in accordance with the proposed amendment will not create the 
possibility of a new or different type of accident from any accident 
previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant reduction 
in a margin of safety.
    The standby emergency power sources are primarily a support 
system for systems required to be operable for accident mitigation. 
SR 3.8.1.5 demonstrates the standby emergency power source 
operation, during a loss of offsite power actuation test signal in 
conjunction with an ESF actuation signal. Performing the 
surveillance in Mode 1, 2, 3, or 4 will require an assessment to 
determine that plant safety is maintained or will be enhanced. There 
are no new or significant changes to the initial conditions 
contributing to accident severity or consequences. The proposed 
amendment will not otherwise affect the plant protective boundaries, 
will not cause a release of fission products to the public, nor will 
it degrade the performance of any other structures, systems or 
components (SSCs) important to safety. Therefore, allowing a portion 
of the surveillance to be performed in Mode 1, 2, 3, or 4, will not 
result in a significant reduction in the margin of safety.

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

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

    Date of amendment request: May 22, 2002.
    Description of amendment request: The proposed amendment revises 
Technical Specifications (TSs) 3/4.3.5, allowing the automatic 
operation of the atmospheric steam relief valves during Mode 2 to 
maintain secondary side pressure at or below an indicated steam 
generator pressure of 1225 psig during startup and shutdown of the 
reactors.

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

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change only provides another method of controlling 
the SG PORVs [steam generator power-operated relief valves] under 
specified operating conditions. The operating conditions in 
Specification 3/4.3.5 remain unchanged. No change is required to 
plant design since the proposed method of control is already part of 
the plant's configuration. The proposed method of control is the 
same method of control

[[Page 45572]]

normally required by the specification in Modes 1 and 2. The 
proposed method of control will not impact the accident analysis 
assumptions or results. 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 method of controlling the SG PORVs is the same 
method that these valves are controlled in Modes 1 and 2 by the 
specification under normal conditions. The proposed change will 
allow the setpoint of these valves to be adjusted to support startup 
and shutdown activities. The adjustment of the setpoint is 
restricted so that the accident analysis is not impacted. No change 
to the design of the valves or plant configuration is required to 
implement the proposed change. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change that will allow for an additional method of 
controlling the SG PORVs during startup and shutdown activities is 
consistent with the operating restrictions for the current method of 
valve control. The accident analysis assumptions and results will 
remain unaffected. 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis, & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: May 23, 2002.
    Description of amendment request: The proposed amendment revises 
the near-end of life (EOL) Moderator Temperature Coefficient (MTC) 
Surveillance Requirements by placing a set of conditions on core 
operation.
    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 probability or consequences of accidents previously 
evaluated in the UFSAR [updated final safety analysis report] are 
unaffected by this proposed change because there is no change to any 
equipment response or accident mitigation scenario. There are no 
additional challenges to fission product barrier integrity. 
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 change. 
The proposed change does 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.
    The margin of safety associated with the acceptance criteria of 
any accident is unchanged. The proposed change will have no affect 
on the availability, operability, or performance of the safety-
related systems and components. A change to a surveillance 
requirement is proposed, but the limiting conditions for operation 
required by the Technical Specifications are not changed.
    The Technical Specifications Bases are founded in part on the 
ability of the regulatory criteria to be satisfied assuming the 
limiting conditions for operation are met for the various systems. 
Conformance to the regulatory criteria for operation with the 
conditional exemption from the near-EOL MTC measurement is 
demonstrated and the regulatory limits are not exceeded. Therefore, 
the margin of safety as defined in the TS [technical specification] 
is not reduced and 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis, & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

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 amendment request: May 24, 2002.
    Description of amendment request: The proposed amendments would 
allow Mode 2 (startup) operation with two, rather than three, 
intermediate range monitor channels per trip system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The intermediate range monitors (IRMs) monitor neutron flux 
levels in the reactor core during startup. The IRM detectors are 
capable of generating a trip signal during a continuous rod 
withdrawal error in the startup range. However, the IRMs perform no 
function related to the probability of occurrence of a previously 
evaluated accident. Also, the IRM trip signal is not necessary to 
mitigate the limiting control rod withdrawal error. The limiting 
case assumes the trip signal is generated from the safety-related 
average power range monitor (APRM). Therefore, the consequences of 
this previously evaluated abnormal operating transient are not 
increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change reduces the number of required operable IRM 
channels per trip system from three to two. However, the manner in 
which the actuation logic functions and the systems respond are 
unaffected by the proposed change. Furthermore, the IRMs will 
continue to perform their design function of core monitoring during 
startup and mitigating nonlimiting transient events postulated to 
occur during startup. Therefore, the proposed change cannot 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?
    The Bases for Units 1 and 2 Technical Specification Table 
3.3.1.1-1 state the ``IRMs are capable of generating trip signals 
that can be used to prevent fuel damage resulting from abnormal 
operating transients in the intermediate power (startup) range.'' 
The proposed change ensures the IRMs will still effectively mitigate 
these events. The most significant source of reactivity change is 
due to a control rod withdrawal error. With the proposed change, the 
IRMs will continue to

[[Page 45573]]

provide protection against rod withdrawal errors, and peak fuel 
energy depositions will remain below the 170 cal/gm threshold 
criterion defined in the Technical Specifications Bases. Therefore, 
the proposed change does not reduce a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) 
are satisfied. Therefore, the NRC staff proposes to determine that 
the amendment request involves no significant hazards consideration.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: John A. Nakoski.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, 
Vermont Yankee Nuclear Power Station, Vernon, Vermont

    Date of amendment request: March 19, 2002, as supplemented on 
June 3, 2002.
    Description of amendment request: The proposed Technical 
Specification changes involve the removal of the existing scram 
function and Group 1 isolation valve closure functions of the Main 
Steam Line Radiation Monitors (MSLRM). An explicit requirement for 
periodic functional test and calibration of the MSLRM is added to 
maintain operability of the mechanical vacuum pump (MVP) isolation 
function. This proposed no significant hazards consideration 
determination replaces in its entirety the notice published in the 
Federal Register on May 14, 2002 (67 FR 34495).
    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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The scram and Group 1 isolation functions of the MSLRMs do not 
serve as initiators for any of the accidents evaluated in the 
Updated Final Safety Analysis Report (UFSAR). The MSLRM scram 
function is not credited in the UFSAR, and the Group 1 isolation 
trip function of the MSLRMs was only assumed in one design-basis 
event which was the control rod drop accident. Because these 
functions are not initiators of accidents, their removal does not 
increase the probability of occurrence of previously evaluated 
accidents.
    There is no accident analysis that relies on the high radiation 
scram of the reactor protection system and its removal has no impact 
on the consequences of accidents previously evaluated. The results 
of the control rod drop accident analysis remain within approved 
guidelines, thus any potential increase in consequences would not be 
considered significant.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Create the possibility for a new or different kind of 
accident from any previously evaluated.
    The proposed changes to the plant involve limited changes to 
protective circuitry, but do not involve any plant hardware changes 
that could introduce any new failure modes. The changes will not 
affect non-MSLRM scram and isolation functions. In addition, the 
MSLRMs will remain active for other trip/isolation functions, and 
these monitors will still alarm in the control room to alert 
operators to off-normal conditions.
    Therefore, the removal of the Group 1 isolation valve closure 
and scram functions of the MSLRMs does not create the possibility of 
a new or different kind of accident than those previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change involves the elimination of the scram and 
Group I isolation signal from the MSLRMs. Operation under the 
proposed change will not change any plant operation parameters, nor 
any protective system setpoints other than removal of these 
functions. The effects of the control rod drop accident without the 
MSLRM scram and isolation signal results in doses which remain well 
within 10 CFR Part 100, ``Reactor Site Criteria,'' limits.
    Therefore, this change does not involve a significant reduction 
in the 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. David R. Lewis, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

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

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

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

    Date of amendment request: June 13, 2002.
    Brief description of amendment request: The proposed amendment 
would revise Technical Specifications Section 4.13.A, ``Inspection 
Requirements,'' to allow the use of the optimum eddy current probe 
size when performing steam generator tube inspections. The proposed 
amendment would also correct several grammatical errors.
    Date of publication of individual notice in Federal Register: 
June 25, 2002 (67 FR 42806).
    Expiration date of individual notice: July 25, 2002.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission 
has determined for each of these amendments that the application 
complies with the standards and requirements of the Atomic Energy 
Act of 1954, as amended (the Act), and the Commission's rules and 
regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with 
these actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 
51.22(b), no environmental impact statement or environmental 
assessment need be prepared for these amendments. If the Commission 
has prepared an environmental assessment under the special 
circumstances provision in 10 CFR 51.12(b) and has made a 
determination based on that assessment, it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the 
Commission's related letter, Safety Evaluation and/or Environmental 
Assessment as indicated. All of these items are available for public 
inspection at the Commission's Public Document Room, located at One 
White Flint North, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the 
Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to 
ADAMS or if there are problems in accessing the documents located in 
ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 
1-800-397-4209, 301-415-4737 or by email to [email protected].

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

    Date of application for amendment: April 17, 2001.
    Brief description of amendment: The amendment makes editorial 
and administrative corrections to Technical Specifications (TS) 
Section 3.3,

[[Page 45574]]

``Instrumentation,'' and eliminates minor discrepancies between TS 
Section 3.3 and other plant licensing basis documents.
    Date of issuance: June 25, 2002.
    Effective date: As of the date of issuance and shall be 
implemented within 30 days.
    Amendment No.: 152.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 26, 2001 
(66 FR 66463). The Commission's related evaluation of the amendment 
is contained in a Safety Evaluation dated June 25, 2002.
    No significant hazards consideration comments received: No.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, 
Units Nos. 1, 2, and 3, Maricopa County, Arizona

    Date of application for amendments: December 13, 2001.
    Brief description of amendments: The amendments revise Item d of 
TS 5.5.11, ``Ventilation Filter Testing Program (VFTP),'' to lower 
the maximum allowable differential pressure across the engineered 
safety features ventilation systems units when tested at the 
specified system flow rates.
    Date of issuance: June 18, 2002.
    Effective date: June 18, 2002, and shall be implemented within 
60 days of the date of issuance.
    Amendment Nos.: Unit 1-142, Unit 2-142, Unit 3-142.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 5, 2002 (67 
FR 5325). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 18, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 19, 2001, as 
supplemented March 27, 2002
    Brief description of amendment: The amendment revises Technical 
Specification 5.5.16 to eliminate the requirement to perform post-
modification containment integrated leakage rate testing following 
replacement of the Unit 2 steam generators.
    Date of issuance: June 27, 2002.
    Effective date: As of the date of issuance to be implemented 
following the Unit 2 refueling and steam generator replacement 
outage in spring 2003.
    Amendment No.: 230.
    Renewed License No. DPR-69: Amendment revised the Technical 
Specifications.
    Date of initial notice in Federal Register: March 19, 2002 (67 
FR 12599).
    The March 27, 2002, supplemental letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation 
dated June 27, 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: February 21, 2002.
    Brief description of amendment: The amendment authorizes changes 
to the Updated Final Safety Analysis Report (UFSAR) and the 
Technical Requirements Manual to eliminate the chlorine detection 
function from the control center heating, ventilation and air 
conditioning system. Changes to the UFSAR are subject to the 
requirements of 10 CFR 50.59; however, the changes were submitted to 
the Nuclear Regulatory Commission for review and approval since they 
involve the elimination of an automatic action.
    Date of issuance: June 26, 2002.
    Effective date: As of the date of issuance and shall be 
implemented within 60 days.
    Amendment No.: 147.
    Facility Operating License No. NPF-43: Amendment revises the 
UFSAR and TRM.
    Date of initial notice in Federal Register: April 16, 2002 (67 
FR 18643). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 26, 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: May 24, 2001.
    Brief description of amendment: The amendment deletes License 
Condition 2.C.(11), which required inspection of the low-pressure 
turbine discs during the second refueling outage and specified that 
the frequency of subsequent inspections should be in accordance with 
the turbine manufacturer's recommendations. License Condition 
2.C.(11) is no longer applicable to Fermi 2.
    Date of issuance: June 26, 2002.
    Effective date: As of the date of issuance and shall be 
implemented within 30 days.
    Amendment No.: 148.
    Facility Operating License No. NPF-43: Amendment revises the 
License.
    Date of initial notice in Federal Register: December 12, 2001 
(66 FR 64288). The Commission's related evaluation of the amendment 
is contained in a Safety Evaluation dated June 26, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application of amendments: June 21, 2000, as 
supplemented by letters dated April 30 and May 20, 2002.
    Brief description of amendments: The amendments authorize 
changes to the Updated Final Safety Analysis Report Section 10.4.7, 
``Emergency Feedwater System.''
    Date of Issuance: June 11, 2002.
    Effective date: As of the date of issuance and shall be 
implemented within 30 days from the date of issuance.
    Amendment Nos.: 325/325/326.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-
55: Amendments authorized changes to the UFSAR.
    Date of initial notice in Federal Register: July 26, 2000 (65 FR 
46008). The supplement dated April 30 and May 20, 2002, provided 
clarifying information that did not change the scope of the June 21, 
2000, 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 June 11, 
2002.
    No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of application for amendment: April 16, 2001, as 
supplemented by letters dated November 8, 2001, and February 11, 
2002.
    Brief description of amendment: The amendment authorizes the 
licensee to modify the Final Safety Analysis Report (FSAR) to allow 
an unisolable drain line between the reactor core isolation cooling 
and the control rod drive/condensate pump rooms and identify the 
pump room doors and penetration seals that are not watertight. In 
addition, the change documents the minimum acceptable safe shutdown 
equipment.
    Date of issuance: June 19, 2002.
    Effective date: June 19, 2002, and shall be implemented in the 
next periodic update to the FSAR in accordance with 10 CFR 50.71(e).
    Amendment No.: 176.
    Facility Operating License No. NPF-21: The amendment revises the 
FSAR.
    Date of initial notice in Federal Register: May 16, 2001 (66 FR 
27175). The November 8, 2001 and February 11, 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 June 19, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 11, 2002.
    Brief description of amendment: The amendment revises Technical 
Specification 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

[[Page 45575]]

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: June 27, 2002.
    Effective date: June 27, 2002.
    Amendment No.: 212.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 14, 2002 (67 FR 
34485). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 27, 2002.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, 
Unit No. 1, Pope County, Arkansas

    Date of amendment request: March 13, 2002.
    Brief description of amendment: The amendment revises 
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: June 10, 2002.
    Effective date: As of the date of issuance and shall be 
implemented in conjunction with the implementation of Amendment No. 
215.
    Amendment No.: 217.
    Renewed Facility Operating License No. DPR-51: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: April 30, 2002 (67 
FR 21287). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 10, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 1, 2001.
    Brief description of amendments: These amendments revise 
Limerick Generating Station's Units 1 and 2 Technical Specifications 
by deleting Section 6.4, ``Training.''
    Date of issuance: June 14, 2002.
    Effective date: As of date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 160/122.
    Facility Operating License Nos. NPF-39 and NPF-85: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 
FR 55018). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 14, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 18, 2002.
    Brief description of amendment: The proposed amendment revises 
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: June 26, 2002.
    Effective date: As of the date of issuance and shall be 
implemented within 60 days of issuance.
    Amendment No.: 203.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 14, 2002 (67 FR 
34487). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 26, 2002.
    No significant hazards consideration comments received: No.

GPU Nuclear Inc., Docket No. 50-320, Three Mile Island Nuclear 
Station, Unit 2, Dauphin County, Pennsylvania

    Date of amendment request: February 8, 2002.
    Brief description of amendment request: The amendment would 
replace referenced control requirements for access to high radiation 
areas with the actual requirements of 10 CFR Part 20, and would 
replace the existing Three Mile Island Nuclear Station, Unit 2, 
Technical Specifications (TS) Section 6.11 with the wording 
contained in Three Mile Island Nuclear Station, Unit 1, TS Section 
6.12.
    Date of issuance: June 27, 2002.
    Effective date: As of the date of issuance and shall be 
implemented within 30 days.
    Amendment No.: 58.
    Facility Operating License No. DPR-73: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 2, 2002 (67 FR 
15623). The Commission's related evaluation of the amendment is 
contained in a safety evaluation dated June 27, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 18, 2001, as 
supplemented by letters dated January 30, and March 1, 2002.
    Brief description of amendment: The amendment revises (1) the 
reference point for reactor vessel level instrumentation 
specifications to use instrument ``zero'' instead of ``top of active 
fuel;'' (2) simplifies the safety limits and limiting safety system 
settings to eliminate specifications that are unnecessary, outdated, 
or redundant to other Technical Specifications (TSs); (3) changes 
the reactor coolant system pressure safety limit from 1335 psig to 
1332 psig to correct a minor calculation error; and (4) makes 
corresponding TS Bases changes.
    Date of issuance: June 11, 2002.
    Effective date: As of the date of issuance and shall be 
implemented within 60 days.
    Amendment No.: 128.
    Facility Operating License No. DPR-22: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 25, 2001 (66 FR 
38764). The supplements dated January 30 and March 1, 2002, provided 
additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did 
not change the staff's original proposed no significant hazards 
consideration determination. The Commission's related evaluation of 
the amendment is contained in a Safety Evaluation dated June 11, 
2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: January 10, 2002.
    Brief description of amendments: The amendments revise 
Surveillance Requirement (SR) 3.0.3 to extend the delay period 
before entering a Limiting Condition for Operation following a 
missed surveillance. The delay period 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: June 19, 2002.
    Effective date: June 19, 2002, shall be implemented within 30 
days from the date of issuance.
    Amendment Nos.: Unit 1-153; Unit 2-153.
    Facilit Operating License Nos. DPR-80 and DPR-82: 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 June 19, 2002.

[[Page 45576]]

    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 22, 2002, as supplemented by 
letters dated June 10, and June 14, 2002.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) TS 5.5.2.11.f.1.h, ``Steam Generator (SG) Tube 
Surveillance Program,'' to more clearly delineate the scope of the 
SG tube inspection required in the tubesheet region. This TS change 
will apply only to Cycle 12 (Unit 2) and Cycle 11 (Unit 3) 
operations.
    Date of issuance: June 17, 2002.
    Effective date: June 17, 2002, to be implemented within 30 days 
from the date of issuance.
    Amendment Nos.: Unit 2--189 ; Unit 3--180.
    Facility Operating License Nos. NPF-10 and NPF-15: The 
amendments revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (67 FR 38150 dated May 31, 2002). The notice 
provided an opportunity to submit comments on the Commission's 
proposed no significant hazards consideration determination. No 
comments have been received. The notice also provided for an 
opportunity to request a hearing by July 1, 2002, but indicated that 
if the Commission makes a final no significant hazards consideration 
determination any such hearing would take place after issuance of 
the amendment. The Commission's related evaluation of the amendment, 
finding of exigent circumstances, consultation with the State of 
California and final determination of no significant hazards 
consideration are contained in a Safety Evaluation dated June 17, 
2002. The June 10, and June 14, 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.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: March 25, 2002, as supplemented by 
the letter dated April 23, 2002.
    Brief description of amendments: The proposed change revised the 
Technical Specification (TS) 3.7.3, ``Feedwater Isolation Valves 
(FIVs) and Associated Bypass Valves,'' to adopt the NUREG-1431, 
``Standard Technical Specifications for Westinghouse Plants,'' 
Revision 2 version of the specification. The requirements of revised 
TS 3.7.3 added, among other things, operability and suitable 
surveillance requirements for Feedwater Control Valves and 
Associated Bypass Valves and allowed for the extended out-of-service 
time for one or more FIVs. In addition, a footnote which allowed a 
one-time extension for Condition A Completion Time, has been deleted 
because it is no longer applicable.
    Date of issuance: June 20, 2002.
    Effective date: As of the date of issuance and shall be 
implemented within 60 days from the date of issuance.
    Amendment Nos.: NPF-87, Amendment No. 97 and NPF-89, Amendment 
No. 97.
    Facility Operating License Nos. NPF-87 and NPF-89: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 14, 2002 (67 FR 
34492). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 20, 2002.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 1st day of July 2002.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 02-16956 Filed 7-8-02; 8:45 am]
BILLING CODE 7590-01-P