[Federal Register Volume 67, Number 33 (Tuesday, February 19, 2002)]
[Notices]
[Pages 7410-7427]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-3750]


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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 Pub. L. 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 January 25, 2002 through February 7, 2002. 
The last biweekly notice was published on February 5, 2002 (67 FR 
5323).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC's Public Document 
Room (PDR), located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland. The filing of requests for a 
hearing and petitions for leave to intervene is discussed below.
    By March 21, 2002, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714, which is 
available at the NRC's PDR, located at One White Flint North, 11555 
Rockville Pike (first floor), Rockville, Maryland. Publicly available 
records will be accessible from the Agencywide Documents Access and 
Management Systems (ADAMS) Public Electronic Reading Room on the 
internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or an Atomic 
Safety and Licensing Board, designated by the Commission or by the 
Chairman of the Atomic Safety and Licensing Board Panel, will rule on 
the request and/or petition; and the Secretary or the designated Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first

[[Page 7411]]

prehearing conference scheduled in the proceeding, but such an amended 
petition must satisfy the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland, by the above date. A copy of the petition should 
also be sent to the Office of the General Counsel, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and to the attorney 
for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. Publicly available records 
will be accessible from the Agencywide Documents Access and Management 
Systems (ADAMS) Public Electronic Reading Room on the internet at the 
NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not 
have access to ADAMS or if there are problems in accessing the 
documents located in ADAMS, contact the NRC PDR Reference staff at 1-
800-397-4209, 304-415-4737 or by e-mail to [email protected].

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

    Date of amendment request: November 30, 2001.
    Description of amendment request: A change is proposed to 
Surveillance Requirement (SR) 3.0.3 to allow a longer period of time to 
perform a missed surveillance. The time is extended from the current 
limit of ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is less'' to ``* * * up to 24 hours or up to the 
limit of the specified Frequency, whichever is greater.'' In addition, 
the following requirement would be added to SR 3.0.3: ``A risk 
evaluation shall be performed for any Surveillance delayed greater than 
24 hours and the risk impact shall be managed.''
    The Nuclear Regulatory Commission (NRC) staff issued a notice of 
opportunity for comment in the Federal Register on June 14, 2001, (66 
FR 32400), on possible amendments concerning missed surveillances, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on September 28, 2001, (66 FR 
49714). The licensees affirmed the applicability of the following NSHC 
determination in its application dated November 30, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

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

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. A 
missed surveillance will not, in and of itself, introduce new 
failure modes or effects and any increased chance that a standby 
system might fail to perform its safety function due to a missed 
surveillance would not, in the absence of other unrelated failures, 
lead to an accident beyond those previously evaluated. The addition 
of a requirement to assess and manage the risk introduced by the 
missed surveillance will further minimize possible concerns. Thus, 
this change does not create the possibility of a new or different 
kind of

[[Page 7412]]

accident from any accident previously evaluated.

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

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

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

Arizona Public Service Company, et al., Docket No. STN 50-529, Palo 
Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona

    Date of amendment request: December 21, 2001.
    Description of amendment request: The amendment would revise the 
operating license and the Technical Specifications (TSs) to support 
replacement of the steam generators and the subsequent increased power 
to a level of 3990 MWt, a 2.94 percent increase.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

a. Evaluation of the Probability of Previously Evaluated Accidents

    Plant Structures, Systems and Components (SSCs) have been 
verified to be capable of performing their intended design functions 
at uprated power conditions. Where necessary, a small number of 
minor modifications will be made prior to implementation of uprated 
power operations so that surveillance test acceptance criteria 
continues to be met. The analysis has concluded that operation at 
uprated power conditions will not adversely affect the capability or 
reliability of plant equipment. Current technical specification 
surveillance requirements ensure frequent and adequate monitoring of 
system and component operability. All systems will continue to be 
operated within current operating requirements at uprated 
conditions. Therefore, no new structure, system or component 
interactions have been identified that could lead to an increase in 
the probability of any accident previously evaluated in the Updated 
Final Safety Analysis Report (UFSAR).

b. Evaluation of the Consequences of Previously Evaluated Accidents

    The radiological consequences were reviewed for all design basis 
accidents (DBAs) (i.e., both LOCA [loss-of-coolant accident] and 
non-LOCA accidents) previously analyzed in the UFSAR. The analyses 
showed that the resultant radiological consequences for both LOCA 
and non-LOCA accidents remained within regulatory and Standard 
Review Plan (SRP) limits at uprated power conditions.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The configuration, operation and accident response of the PVNGS 
[Palo Verde Nuclear Generating Station] Unit 2 SSCs are unchanged by 
operation at uprated power conditions or by the associated proposed 
TS changes. Analyses of transient events have confirmed that no 
transient event results in a new sequence of events that could lead 
to a new accident or different scenario.
    The effect of operation at uprated power conditions on plant 
equipment has been evaluated. No new operating mode, safety-related 
equipment lineup, accident scenario, or equipment failure mode was 
identified as a result of operating at uprated conditions. In 
addition, operation at uprated power conditions does not create any 
new failure modes that could lead to a different kind of accident. 
Minor plant modifications, to support implementation of uprated 
power conditions, will be made as required to existing SSCs. The 
basic design function of all SSCs remains unchanged and no new 
equipment or systems have been installed that could potentially 
introduce new failure modes or accident sequences.
    Based on these analyses, it is concluded that no new accident 
scenarios, failure mechanisms or limiting single failures are 
introduced as a result of the proposed changes. The proposed changes 
do not have an adverse effect on any safety-related system or design 
basis function. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    A comprehensive analysis was performed to evaluate the effects 
of power uprate on PVNGS Unit 2. This analysis identified and 
defined the major input parameters to the NSSS [nuclear steam supply 
system], reviewed NSSS design transients, and reviewed the 
capabilities of the NSSS and BOP [balance-of-plant] fluid systems, 
NSSS/BOP interfaces, NSSS and BOP control systems, and NSSS and BOP 
SSCs. NSSS accident analyses were re-performed or reviewed to 
confirm that acceptable results were maintained and that the 
radiological consequences remained within regulatory and SRP limits. 
The nuclear and thermal hydraulic performance of nuclear fuel was 
also reviewed to confirm acceptable results. The analyses confirmed 
that all NSSS and BOP SSCs are capable, some with minor 
modifications, to safely support operations at uprated power 
conditions.
    The margin of safety of the reactor coolant pressure boundary is 
maintained under uprated power conditions. The design pressure of 
the reactor pressure vessel and reactor coolant system will not be 
challenged as the pressure mitigating systems were confirmed to be 
sufficiently sized to adequately control pressure under uprated 
power conditions.
    Reanalysis of containment structural integrity under DBA 
conditions indicates that the calculated peak containment pressure 
(Pa) increases from 52.0 psig to 58.0 psig, but remains less than 
the containment internal design pressure of 60 psig. The proposed 
value for Pa has been rounded up from the actual calculated value of 
57.85 psig.
    Radiological consequences of the following accidents were 
reviewed: Main Steam Line Break, Locked Reactor Coolant Pump (RCP) 
Rotor, CEA Ejection, Small Steam Line Break Outside Containment, 
Steam Generator Tube Rupture, LBLOCA [large break loss of coolant 
accident], SBLOCA [small break loss of coolant accident], Waste Gas 
Decay Tank Rupture, Liquid Waste Tank Failure, and Fuel Handling 
Accident. The resultant radiological consequences for each of these 
accidents remained within regulatory and SRP limits at uprated power 
conditions.
    The analyses supporting operation at power uprate conditions 
have demonstrated that all systems and components are capable of 
safely operating at uprated power conditions. All DBA acceptance 
criteria will continue to be met. Therefore, it is concluded that 
the proposed changes do not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three

[[Page 7413]]

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the request for an amendment involves no 
significant hazards consideration.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Section Chief: Stephen Dembek.

Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-317, Calvert 
Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

    Date of amendment request: January 31, 2002.
    Description of amendment request: The proposed amendment would 
allow a one-time five-year extension, for a total of 15 years, for the 
performance of the next Unit 1 integrated leak rate test (ILRT). The 
proposed amendment would also exempt Unit 1 from the requirement to 
perform a post-modification containment ILRT associated with the steam 
generator replacement.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:


    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated:

[Extension of Type A integrated leakage rate testing:]

    This proposed one-time extension of the Type A test interval 
does not increase the probability of an accident since there are no 
design or operating changes involved and the test is not an accident 
initiator. The proposed extension of the test interval does not 
involve a significant increase in the consequences of an accident 
since research documented in NUREG-1493 has found that, generically, 
fewer than three percent of the potential containment leak paths are 
not identified by Type B and C testing. Calvert Cliffs, through 
testing and containment inspections, also provides a high degree of 
assurance that the Containment will not degrade in a manner 
detectable only by a Type A test. Inspections required by the 
Maintenance Rule (10 CFR 50.65) and by the American Society of 
Mechanical Engineers Boiler and Pressure Vessel Code are performed 
to identify containment degradation that could affect leak 
tightness.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

[Exemption from post-modification testing:]

    The steam generator replacement activities do not affect the 
containment structure or the actual containment liner. Access for 
the replacement steam generators as well as removal of the old steam 
generators will be through the equipment hatch. However, the outer 
shell of the steam generators, the inside containment portions of 
the main steam line, the feedwater lines, the auxiliary feedwater 
lines, and the steam generator blowdown lines are all part of the 
primary reactor containment boundary that will be impacted by the 
replacement activities.
    Calvert Cliffs Nuclear Power Plant Technical Specification 
5.5.16 states, ``A program shall be established to implement the 
leakage testing of the Containment as required by 10 CFR 50.54(o) 
and 10 CFR part 50, Appendix J, Option B. This program shall be in 
accordance with the guidelines contained in Regulatory Guide 1.163, 
`Performance-Based Containment Leak-Test Program' dated September 
1995, including errata.'' Regulatory Guide 1.163, ``Performance-
Based Containment Leak-Test Program,'' endorses NEI [Nuclear Energy 
Institute] 94-01, Revision 0 for methods acceptable to comply with 
the requirements of Option B. Prior to returning the Containment to 
operation, NEI 94-01 requires leakage rate testing (Type A testing 
or local leakage rate testing), following repairs and modification 
that affect the containment leakage integrity.
    The affected area of the primary containment boundary is also 
part of the pressure boundary of an American Society of Mechanical 
Engineers (ASME) Class 2 component/piping system and, as such, the 
planned replacement of the steam generators are subject to the 
repair and replacement requirements of ASME Section XI. The ASME 
Section XI surface examination, volumetric examination, and system 
pressure test requirements are more stringent than the Appendix J, 
Option B testing requirements. The acceptance criteria for ASME 
Section XI system pressure testing of welded joints is ``zero 
leakage.'' In addition, the test pressure for the system pressure 
test will be approximately 17 times that of Appendix J, Option B 
test.
    The objective of the Type A test is to assure the leak-tight 
integrity of the area affected by the modification. Although the 
leak test is in a direction reverse to that of the design basis 
accident environment, the ASME Section XI inspection and testing 
requirements more than fulfill the intent of the requirements of 
Appendix J, Option B with the exception of secondary side access 
manways. Section 9.2.1, NEI 94-01, Revision 0 allows reverse testing 
if justified. Section XI pressure test applies a sealing pressure to 
the secondary manway due to the inward door swing configuration. 
Hence, a Type B local leak rate test will be performed for the 
secondary manways. For all other affected components, reverse 
testing is justified since the acceptance criteria for ASME Section 
XI system pressure testing of welded joints is ``zero leakage,'' and 
the test pressure of the system pressure test will be approximately 
17 times that of a Type A test. Hence, the probability or 
consequences of design bases accidents previously evaluated are 
unchanged.
    Therefore, the proposed revision to Technical Specification 
5.5.16 to eliminate the requirement to perform post-modification 
containment integrated leakage rate testing following replacement of 
Unit 1 steam generators will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.

[Extension of Type A integrated leakage rate test interval:]

    This proposed one-time extension to the interval for the Type A 
test does not involve any design or operational changes that could 
lead to a new or different kind of accident from any accident 
previously evaluated. The test itself is not changing and will be 
performed after a longer interval. The proposed change does not 
involve a physical alteration of the plant (no new or different type 
of equipment will be installed) or a change in the methods governing 
normal plant operation.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

[Exemption from post-modification testing:]

    The proposed revision does not involve a physical change to the 
plant and there are no changes to the operation of the plant that 
could introduce a new failure mode. As described above in Item 1, 
the objective of the Appendix J, Option B test is to assure that 
leak-tight integrity of the area affected by the modification. The 
ASME Section XI inspection and testing requirements are more 
stringent than the Appendix J, Option B testing requirements.
    Therefore, the proposed revision to Technical Specification 
5.5.16 to eliminate the requirement to perform post-modification 
containment integrated leakage rate testing following replacement of 
Unit 1 steam generators will not create the possibility of a new or 
different [kind] of accident from any previously evaluated.
    3. Would not involve a significant reduction in the margin of 
safety.

[Extension of Type A integrated leakage rate test interval:]

    The generic study of the increase in the Type A test interval, 
NUREG-1493, concluded there is an imperceptible increase in the 
plant risk associated with extending the test interval out to 20 
years. Further, the extended test interval would have a minimal 
effect on this risk since Type B and C testing detect 97 percent of 
potential leakage paths. For the requested change in the Calvert 
Cliffs Integrated Leakage Rate Test interval, it was determined that 
the risk contribution of leakage will increase 0.07 percent (based 
on change in offsite dose). This change is considered very small and 
does not represent a significant reduction in the margin of safety.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

[Exemption from post-modification testing:]

    As described above in Item 1, the ASME Section XI surface 
examination, volumetric examination, and system pressure test 
requirements are more stringent than the Appendix J, Option B 
testing requirements. The acceptance criteria for ASME Section XI

[[Page 7414]]

system pressure testing of welded joints is ``zero leakage.'' In 
addition, the test pressure for the system pressure test will be 
approximately 17 times that of Appendix J, Option B test.
    Therefore, the proposed revision to Technical Specification 
5.5.16 to eliminate the requirement to perform post-modification 
containment integrated leakage rate testing following replacement of 
Unit 1 steam generators 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 proposed to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Joel Munday (Acting).

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

    Date of amendment request: November 6, 2001, as supplemented 
December 27, 2001.
    Description of amendment request: The proposed amendment would: 1) 
Increase the allowable nominal average fuel assembly enrichment from 
4.5 w/o U-235 to 4.85 w/o U-235 for all regions of the spent fuel pool, 
the new fuel storage racks (dry), and the reactor core; 2) Allow fuel 
to be located under the cell blockers in 40 empty Region B storage 
cells; and, 3) Credit spent fuel pool soluble boron for reactivity 
control during normal conditions to maintain spent fuel pool 
Keff 0.95.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Previously evaluated Final Safety Analysis Report (FSAR) Chapter 
14 accidents are a fuel handling accident either in the spent fuel 
pool (SFP) or in containment, and a spent fuel cask drop accident. 
Since there are no changes to plant equipment, nor any changes in 
how fuel is moved, there are no changes to the probability of a fuel 
handling accident in the spent fuel pool or containment.
    Since there are no changes to plant equipment, nor any changes 
in how a shielded cask would be moved, there are no changes to the 
probability of a spent fuel cask drop accident.
    The consequences of a fuel drop accident in either containment 
or the spent fuel pool are not affected, since none of the inputs to 
these fuel drop accidents is affected. There are no physical 
hardware changes made to the plant. The limiting fuel burnup is not 
changed, nor is there any change in the source term of radioactivity 
present in the fuel. Allowing fuel to be stored in the 40 Region B 
locations currently empty, does not alter the existing FSAR 
conclusion that a dropped fuel assembly or consolidated storage box 
could not strike more than one fuel assembly in the storage rack. 
This is still true since the fuel stored in these 40 locations is 
stored at the same elevation as fuel in any other storage locations. 
The FSAR states that the worst fuel handling incident that could 
occur in the SFP is the drop of a fuel assembly to the pool floor, 
with resultant failure of 14 fuel rods when the assembly rotates and 
impacts a protruding structure. Radiological consequences for both 
the failure of 14 rods and the entire fuel assembly are presented in 
the FSAR. The storage of fuel in the 40 currently blocked locations 
does not affect this FSAR sequence of events for the dropped fuel 
assembly in the SFP accident. The amount of soluble boron 
concentration necessary in the SFP to ensure that Keff is 
maintained  0.95 on a 95/95 bases is increased from 800 
ppm to 1400 ppm. However, this increase in required SFP soluble 
boron concentration does not increase any dose consequences from the 
fuel drop accident in the SFP. The increase in soluble boron 
concentration from 800 ppm to 1400 ppm is a result of crediting an 
additional 600 ppm of SFP soluble boron under normal conditions.
    The consequences of a spent fuel cask drop accident in the SFP 
is not affected, since none of the inputs to the spent fuel cask 
drop accident is affected. There are no physical hardware changes 
made to the plant. The limiting fuel burnup is not changed, nor is 
there any change in the source term of radioactivity present in the 
fuel. The amount of soluble boron concentration necessary in the SFP 
to ensure that Keff is maintained  0.95 on a 
95/95 bases is increased from 800 ppm to 1400 ppm. However, this 
increase in required SFP soluble boron concentration does not 
increase any dose consequences from the spent fuel cask drop 
accident in the SFP. The increase in soluble boron concentration 
from 800 ppm to 1400 ppm is a result of crediting an additional 600 
ppm of SFP soluble boron under normal conditions.
    With regard to the proposed change in the design features 
section of Technical Specifications (TS), which would allow higher 
enrichments in the new fuel storage (dry) vault, there are no FSAR 
Chapter 14 accident conditions currently analyzed, therefore there 
can be no change in probability or consequences of an existing 
accident.
    With regard to the proposed change in the design features 
section of TS, which would allow higher enrichments in the reactor 
core, enrichment by itself is not a parameter which will affect the 
probability or consequences of an accident previously analyzed. The 
effects of enrichment on other reactor core parameters such as 
shutdown margin, MTC [moderator temperature coefficient] and power 
distributions is considered by meeting the exi[s]ting TS 
requirements for these parameters. Also, the reactor core 
radioactive source term is not affected since the exiting design 
basis analysis bounds use of the proposed enrichment. Therefore, a 
change in the maximum enrichment limit will not impact any safety 
analyses because the important inputs to these analyses are 
protected by Technical Specifications. Since there are no changes to 
these existing reactor core TS parameter limits, there will be no 
effect on the probability or consequences of an accident previously 
analyzed.
    Therefore, based on the above analysis, the proposed changes do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The changes to be made primarily affect nuclear criticality 
analysis and do not create a new or different kind of accident. 
Changes in allowed enrichment, boraflex credit, soluble boron 
credit, and allowing fuel to be stored in 40 additional locations 
are all impacts to the SFP criticality analysis. The SFP criticality 
analysis is part of the basic design of the system and is not an 
accident. The ability to maintain the SFP Keff 
 0.95, as well as within the 10 CFR 50 App. A GDC62 
criteria [General Design Criterion (GDC)-62, ``Prevention of 
criticality in fuel storage and handling,'' of Appendix A, ``General 
Design Criteria for Nuclear Power Plants,'' to 10 CFR part 50] of 
sub-critical have been evaluated. Criticality impacts are more 
appropriately discussed under the margin of safety criterion.
    Since there are no changes to the plant equipment, there is no 
possibility of a new or different kind of accident being initiated 
or affected by equipment issues. There are no changes in how fuel is 
moved or qualified for storage, so a new accident cannot be 
initiated from fuel handling related procedures.
    Higher SFP soluble boron concentrations are required than 
previously required to compensate for the positive reactivity 
insertions from postulated accident conditions (i.e., dropped cask). 
However, merely increasing the amount of SFP soluble boron required 
for compensating for the existing analyzed accident does not create 
the potential for a new or different kind of accident.
    With regard to the proposed change in the design section of TS, 
which would allow higher enrichments in the new fuel storage (dry) 
vault, no new or different kind of accident conditions are created. 
The existing new fuel storage analysis previously submitted to the 
NRC is not altered, and already bounds enrichments up to 5.0 w/o U-
235.
    With regard to the proposed change in the design features 
section of TS, which would allow higher enrichments in the reactor 
core, the higher enrichment fuel in the reactor core does not 
require any new or different plant

[[Page 7415]]

equipment, and does not change the manner in which currently 
installed equipment is operated. There are no changes to normal core 
operation, and the unit will meet all applicable design criteria and 
will operate within the existing reactor core TS limits. No new 
failure modes have been created for any system, component or piece 
of equipment, and no new single failure mechanisms are introduced. 
Therefore, allowing higher enrichments in the reactor core will not 
create a new or different kind of accident condition.
    Therefore, based on the above analysis, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The margin of safety relevant to the SFP are:
     To ensure that the SFP Keff remains 
 0.95 on a 95/95 basis to ensure the criticality safety 
of the SFP.
     To ensure that the spent fuel in the SFP remains 
adequately cooled so that the fission product barriers remain 
intact.
    A criticality analysis has been performed to ensure that the 
spent fuel pool Keff remains  0.95 on a 95/95 
basis under all normal and postulated accident conditions. Thus the 
margin of criticality safety is not changed. Most of the changes in 
the criticality analysis are of an input nature, such as a change in 
allowed enrichment. The only change in methodology is the crediting 
of soluble boron for normal conditions. The approach used is 
consistent with WCAP-14416-NP-A. The NRC has previously approved for 
other plants similar applications for soluble boron credit for 
normal conditions. The criticality analysis has been performed to 
ensure that the spent fuel pool Keff remains less than 
1.00 on a 95/95 basis even with 0 ppm soluble boron concentration in 
the SFP. This ensures compliance with GDC62.
    The only change that could affect the SFP cooling analysis is 
allowing 40 additional fuel assemblies to be stored in the SFP. The 
current design basis heat load analysis already bounds the storage 
of these fuel assemblies. This ensures that the spent fuel in the 
SFP remains adequately cooled so that the fission product barriers 
remain intact. The current design basis heat load analysis bounds 
the increased fuel storage.
    With regard to the proposed change in the design section of TS, 
which would allow higher enrichments in the new fuel storage (dry) 
vault, there is no significant reduction in the margin of safety. 
The existing new fuel storage analysis previously submitted and 
approved by the NRC is not altered, and already bounds enrichments 
up to 5.0 w/o U-235, to ensure that Keff of the new fuel 
storage racks is maintained  0.95.
    With regard to the proposed change in the design features 
section of TS, which would allow higher enrichments in the reactor 
core, enrichment by itself is not a parameter which will affect the 
margin of safety. The margins of safety, such as fuel DNB 
protection, fuel melt protection and RCS boundary protection, are 
met by complying with the safety analysis and associated TS limits. 
The effects of enrichment on other reactor core parameters such as 
shutdown margin, MTC and power distributions is considered by 
meeting the existing TS requirements for these parameters. 
Therefore, a change in the maximum core enrichment limit will not 
impact any margins of safety because the important inputs to the 
safety analyses are protected by Technical Specifications. Since 
there are no change[s] to these existing reactor core TS parameter 
limits, there will be no effect on the margin of safety.
    Therefore, based on the above analysis, the proposed changes do 
not involve a reduction in a margin of safety.

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

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

    Date of amendment request: December 20, 2001.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TS) for Catawba Nuclear Station, Units 1 and 
2, based on a revised radiological dose consequence analysis of a 
postulated fuel handling accident and weir gate drop accident. The 
licensee has requested these amendments in accordance with the 
requirements of 10 CFR 50.67 which addresses the use of an alternate 
source term at operating reactors, and relevant guidance provided in 
Regulatory Guide (RG) 1.183.
    Specifically, the proposed changes would revise TS 3.7.10, Control 
Room Area Ventilation System (CRAVS), to require immediate suspension 
of movement of irradiated fuel with less than two trains of the CRAVS 
operable. This change is being requested to correct a non-conservatism 
in this TS.
    The proposed change to TS 3.7.11, Control Room Area Chilled Water 
System, would delete the applicability of the specification during core 
alterations and during movement of irradiated fuel. This system is not 
credited as a mitigation system for the postulated fuel handling 
accident or weir gate drop accident.
    The proposed change to TS 3.7.13, Fuel Handling Ventilation Exhaust 
System, would change the Limiting Condition for Operation to require 
two trains be operable during the movement of recently irradiated fuel 
in the fuel building, and to require that movement of recently 
irradiated fuel in the fuel building be suspended if one train becomes 
inoperable. Recently irradiated fuel is defined as fuel that has 
occupied part of a critical reactor core within the previous 72 hours. 
Operability of the Fuel Handling Ventilation Exhaust System would only 
be required during movement of recently irradiated fuel assemblies. 
This change is being requested to incorporate the concept of recently 
irradiated fuel and to correct a non-conservatism in this TS.
    The proposed change to TS 3.9.3, Containment Penetrations, would 
amend the applicability of this specification. Current TS requirements 
regarding closure of the containment equipment hatch, the personnel 
airlock and containment penetrations would only apply during movement 
of recently irradiated fuel assemblies. The applicability of this 
specification during core alterations would be deleted.
    The licensee is requesting these amendments to provide flexibility 
in scheduling outage tasks and to modify unnecessarily restrictive 
containment closure and fuel handling building ventilation system 
requirements. The revised analyses also incorporate updated atmospheric 
dispersion factors for the Control Room intake pathway.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Does operation of the facility in accordance with the proposed 
amendment involve a significant increase in the probability or 
consequences of an accident previously evaluated? No.
    An alternate source term calculation has been performed for 
Catawba Nuclear Station that demonstrates that offsite dose 
consequences of a postulated fuel handling accident or weir gate 
drop accident remain within the limits provided sufficient decay has 
occurred prior to the movement of irradiated fuel without taking 
credit for certain mitigation features such as ventilation filter 
systems and containment closure. Irradiated fuel that has not 
undergone the required decay period of 72 hours is defined to be 
recently irradiated fuel and the currently approved Technical 
Specification requirements are applicable when this recently 
irradiated fuel is being handled.
    The proposed amendment would allow core alterations and movement 
of sufficiently decayed irradiated fuel within the containment 
building with the equipment hatch, personnel air locks and 
containment penetrations open. Operation of the

[[Page 7416]]

Containment Purge Exhaust System (CPES) is not required during 
movement of sufficiently decayed fuel. The amendment also would 
allow movement of irradiated fuel assemblies within the fuel 
building without the Fuel Handling Ventilation Exhaust System 
(FHVES) in operation. Movement of the weir gate is permitted without 
the FHVES in operation provided the irradiated fuel that could be 
impacted by a drop of the weir gate has undergone a minimum decay 
period of 19.5 days.
    This amendment does not alter the methodology or equipment used 
directly in fuel handling operations and weir gate movement. Neither 
ventilation filter systems, the CPES nor the FHVES, is used to 
actually handle fuel. Neither of these systems is an ``accident 
initiator'' either in this sense or any other sense. Similarly, 
neither the equipment hatch, the personnel air locks, nor any other 
containment penetration, nor any component thereof is an accident 
initiator.
    Actual fuel handling operations and weir gate movement 
themselves are not affected by the proposed changes. Therefore, the 
probability of a Fuel Handling Accident and Weir Gate Drop is not 
affected with the proposed amendment. No other accident initiator is 
affected by the proposed changes.
    For the reasons above, the proposed amendment does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The Fuel Handling Accident in Containment has been analyzed 
without credit for filtration by the CPES. Likewise, the Fuel 
Handling Accident in the Fuel Building and the Weir Gate Drop has 
been analyzed without credit for filtration by the FHVES. The 
analyses of these design basis events were conducted with the 
Alternative Source Term Methodology in accordance with 10 CFR 50.67 
and Regulatory Guide 1.183. These analyses show that the resultant 
radiation doses are within the limits specified in 10 CFR 50.67 and 
R.G. 1.183.
    The TEDE [total effective dose equivalent] radiation doses from 
the analyses supporting this LAR [license amendment request] have 
been compared to equivalent TEDE radiation doses estimated with the 
guidelines of R.G. 1.183 Footnote 7. The new values are shown to be 
comparable to the results of the previous analyses.
    For the reasons above, the proposed amendment does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    Does operation of the proposed facility in accordance with the 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated? No.
    The proposed change does not involve addition or modification to 
any plant system, structure, or component. The proposed amendment 
would increase the time during which the equipment hatch and 
personnel air locks could be open during core alterations and 
movement of irradiated fuel. The proposed amendment does not involve 
any change in the operation of these containment penetrations. 
Having these penetrations open does not create the possibility of a 
new accident.
    The proposed amendment also would remove the requirements for 
operability of the CPES and FHVES during core alterations or 
movement of sufficiently decayed irradiated fuel. It does not alter 
the operation of these systems beyond their functional capabilities. 
Modification of the requirements of operability for these systems 
from the plant Technical Specifications does not create the 
possibility of a new accident.
    The requirements for CRAVS are being revised to immediately 
suspend movement of irradiated fuel if one CRAVS train becomes 
inoperable. This change does not have the potential to cause a new 
or different type of accident.
    The proposed amendment does not create the possibility of a new 
or different kind of accident than any previously evaluated.
    Does operation of the facility in accordance with the proposed 
amendment involve a significant reduction in the margin of safety? 
No.
    The assumptions and input used in the analysis are conservative 
as noted below. The design basis Fuel Handling Accidents and Weir 
Gate Drop have been defined to identify conservative conditions 
(concerning offsite power and single failure). The source term and 
radioactivity releases have been calculated pursuant to Regulatory 
Guide 1.183 and with conservative assumptions concerning prior 
reactor operation. The control room atmospheric dispersion factors 
have been calculated with conservative assumptions associated with 
the release. The conservative assumptions and input noted above 
ensure that the radiation doses cited in this License Amendment 
Request are the upper bound to radiological consequences of a Fuel 
Handling Accident either in Containment or the Fuel Building and the 
Weir Gate Drop. The analyses show that there is a significant margin 
between the TEDE radiation doses calculated for the postulated Fuel 
Handling Accident and the Weir Gate Drop accident using the 
Alternative Source Term and the acceptance limits of 10 CFR 50.67 
and Regulatory Guide 1.183.
    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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn , Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: Richard J. Laufer, Acting.

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

    Date of amendment request: January 31, 2002.
    Description of amendment request: The proposed amendment revises 
Technical Specification 5.6.5, ``Core Operating Limits Report (COLR),'' 
to include an additional reference to Entergy Operations, Inc. 
(Entergy) Topical Report ENEAD-01-P, ``Qualification of Reactor Physics 
Methods for Pressurized Water Reactors of the Entergy System.'' This 
topical report documents a Nuclear Regulatory Commission (NRC)-approved 
methodology that can be utilized to determine core operating limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to add the Entergy Topical Report ENEAD-01-
P, ``Qualification of Reactor Physics Methods for Pressurized Water 
Reactors of the Entergy System,'' to the Core Operating Limits 
Report (COLR) references is administrative in nature. The topical 
report has been reviewed and approved by the NRC in [a] Safety 
Evaluation Report dated September 29, 1995 (0CNA099519). The 
physical design or operation of the plant is not impacted by this 
proposed change. The proposed change does not adversely impact 
transient analysis assumptions or results. The COLR-related safety 
analyses will continue to be performed utilizing NRC-approved 
methodologies, and specific reload changes will be evaluated under 
the provisions of 10 CFR 50.59.
    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.
    Adding a reference in the technical specifications to the NRC-
approved methodology in Entergy Topical Report ENEAD-01-P is 
administrative in nature. No physical alterations of plant 
configuration, changes to the plant operating procedures, or 
operating parameters are proposed. No new equipment is being 
introduced, and no equipment is being operated in a manner 
inconsistent with its design. The COLR-related safety analyses will 
continue to be performed utilizing NRC-approved methodologies. A 10 
CFR 50.59 review will continue to be performed to evaluate specific 
reload changes.
    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.

[[Page 7417]]

    The proposed change to reference Entergy Topical Report ENEAD-
01-P is administrative in nature. Existing technical specification 
operability and surveillance requirements are not reduced by the 
proposed change. The cycle-specific COLR limits for future reloads 
will continue to be developed based on NRC-approved methodologies 
and their corresponding physics parameter uncertainties. Technical 
specifications will continue to require that the core be operated 
within these limits and specify appropriate actions to be taken if 
the limits are violated. The COLR-related safety analyses will 
continue to be performed utilizing NRC-approved methodologies, and 
specific reload changes will be evaluated per 10 CFR 50.59.
    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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: January 31, 2002.
    Description of amendment request: The proposed amendment changes 
administrative Technical Specification 5.5.16 regarding Containment 
Integrated Leak Rate Testing (ILRT). The change clarifies the statement 
that the ILRT Program is in accordance with Regulatory Guide 1.163, 
``Performance-Based Containment Leak-Test Program,'' by noting an 
exception based on Nuclear Energy Institute 94-01, ``Industry Guideline 
for Implementing Performance-Based Option of 10 CFR 50, Appendix J.'' 
The effect of this change will be to allow a one-time extension of the 
interval (to 15 years) for performance of the next ILRT.
    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.
    [Appendix J [of] 10 CFR [part] 50] was amended to incorporate 
provisions for performance-based testing in 1995. The proposed 
amendment to Technical Specification (TS) 5.5.16 adds a one-time 
extension to the current interval for Type A testing (i.e., the 
integrated leak rate test). The current interval of ten years, based 
on past performance, would be extended on a one-time basis to 15-
years from the date of the last test. The proposed extension to the 
Type A test cannot increase the probability of an accident since 
there are no design or operating changes involved and the test is 
not an accident initiator. The proposed extension of the test 
interval does not involve a significant increase in the consequences 
since research documented in NUREG-1493, ``Performance Based 
Containment Leak Rate Test Program,'' has found that, generically, 
fewer than 3% of the potential containment leak paths are not 
identified by Type B and C testing. In addition, at ANO-1 [Arkansas 
Nuclear One, Unit 1,] the testing and containment inspections also 
provide a high degree of assurance that the containment will not 
degrade in a manner detectable only by a Type A test. Inspections 
required by the Maintenance Rule (10 CFR 50.65) and by the American 
Society of Mechanical Engineers Boiler and Pressure Vessel Code are 
performed to identify containment degradation that could affect 
leaktightness.
    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 extension to the interval for the Type A test does 
not involve any design or operational changes that could lead to a 
new or different kind of accident from any accidents previously 
evaluated. The test itself is not being modified, but is only 
intended to be performed after a longer interval. The proposed 
change does not involve a physical alteration of the plant (no new 
or different type of equipment will be installed) or a change in the 
methods governing normal plant operation.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The generic study of the increase in the Type A 
test interval, NUREG-1493, concluded there is an imperceptible 
increase in the plant risk associated with extending the test 
interval out to twenty years. Further, the extended test interval 
would have a minimal effect on this risk since Type B and C testing 
detect 97% of potential leakage paths. For the requested change in 
the ANO-1 ILRT interval, it was determined that the risk 
contribution of leakage will increase 0.19%. This change is 
considered very small and does not represent a significant reduction 
in the margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 
2, Ogle County, Illinois

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

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

Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 
2 and 3, York County, Pennsylvania

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

    Date of amendment request: November 30, 2001.
    Description of amendment request: A change is proposed to 
Surveillance Requirement (SR) 3.0.3 to allow a longer period of time to 
perform a missed surveillance. The time is extended from the current 
limit of ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is less'' to ``. . . up to 24 hours or up to the 
limit of the specified Frequency, whichever is greater.'' In addition, 
the following requirement would be added to SR 3.0.3: ``A risk 
evaluation shall be performed for any Surveillance delayed greater than 
24 hours and the risk impact shall be managed.''
    The Nuclear Regulatory Commission (NRC) staff issued a notice of 
opportunity for comment in the Federal Register on June 14, 2001, (66 
FR 32400), on possible amendments concerning missed surveillances, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on September 28, 2001, (66 FR 
49714). The licensees

[[Page 7418]]

affirmed the applicability of the following NSHC determination in its 
application dated November 30, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

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

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. A 
missed surveillance will not, in and of itself, introduce new 
failure modes or effects and any increased chance that a standby 
system might fail to perform its safety function due to a missed 
surveillance would not, in the absence of other unrelated failures, 
lead to an accident beyond those previously evaluated. The addition 
of a requirement to assess and manage the risk introduced by the 
missed surveillance will further minimize possible concerns. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.

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

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

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

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: January 18, 2002.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) for St. Lucie Units 1 and 2 to 
remove the numerical working hour limits stated in the TS. Site 
personnel working hours currently are and will continue to be 
controlled by administrative procedures. The change is consistent with 
Technical Specifications Task Force (TSTF) Item TSTF-258, Rev. 4.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments are administrative in nature and they do 
not affect assumptions contained in plant safety analyses, the 
physical design and/or operation of the plant, nor do they affect 
Technical Specifications that preserve safety analysis assumptions. 
These proposed changes do not change the existing administrative 
controls on plant staff working hours. Any future changes to these 
procedures will be controlled under established procedure control 
processes that will ensure the administrative controls on work hours 
remain effective. Further, the proposed changes do not alter the 
design, function, or operation of any plant component. Therefore, 
operation of the facility in accordance with the proposed amendments 
would not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The changes being proposed are administrative in nature and do 
not introduce a new mode of plant operation or surveillance 
requirement, nor involve a physical modification to the plant. 
Therefore, the design, function, or operation of any plant component 
is not altered. The changes propose to relocate specific controls 
for plant staff working hours from the TS to existing administrative 
procedures. The specific controls for plant staff working hours are 
described in these procedures and require a deliberate decision-
making process to manage the potential for impaired personnel 
performance. Therefore, operation of the facility in accordance with 
the proposed amendments would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed changes conform closely to the industry and NRC 
approved TSTF-258 Rev. 4 and relate to the relocation of TS specific 
working hour limits and controls to administrative procedures that 
control working hours. The specific controls for working hours of 
reactor plant staff are described in procedures that require a 
deliberate decision-making process to manage the potential for 
impaired personnel performance. Furthermore, any future changes to 
these procedures will be controlled under established procedure 
control processes that will ensure the administrative controls on 
work hours remain effective. Therefore, operation of the facility in 
accordance with the proposed amendments would not involve a 
significant reduction in a margin of safety.

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

[[Page 7419]]

    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: January 25, 2002.
    Description of amendment request: The proposed amendments would 
revise the St. Lucie Plant, Units 1 and 2, Technical Specifications, 
Appendix B, ``Environmental Protection Plan (Non-Radiological)'' to 
incorporate the revised terms and conditions of the Incidental Take 
Statement included in the Biological Opinion issued by the National 
Marine Fisheries Service on May 4, 2001, as clarified by NMFS letter 
dated October 8, 2001. These amendments also incorporate administrative 
revisions necessary to change references to the National Pollutant 
Discharge Elimination System Permit to the Wastewater Permit, based on 
a change in administrative authority over these permits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The changes are administrative in nature and would in no way 
affect the initial conditions, assumptions, or conclusions of the 
St. Lucie Unit 1 or Unit 2 accident analyses. In addition, the 
proposed changes would not affect the operation or performance of 
any equipment assumed in the accident analyses. Based on the above 
information, we conclude that the proposed changes would not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    (2) Use of the modified specification would not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The changes are administrative in nature and would in no way 
impact or alter the configuration or operation of the facilities and 
would create no new modes of operation. We conclude that the 
proposed changes would not create the possibility of a new or 
different kind of accident.
    (3) Use of the modified specification would not involve a 
significant reduction in a margin of safety.
    The changes are administrative in nature and would in no way 
affect plant or equipment operation or the accident analysis. We 
conclude that the proposed changes would not result in a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
    Date of amendment request: January 25, 2002.
    Description of amendment request: The proposed amendment would 
revise Section 4.8.1.1.2.g.2 of the St. Lucie Units 1 and 2 Technical 
Specifications (TS), which currently requires a test of the diesel fuel 
oil system piping at elevated pressure once every 10 years. In lieu of 
hydrostatic testing, the diesel fuel oil systems will be included in 
the population of systems subjected to periodic system pressure testing 
at normal operating conditions required by the American Society of 
Mechanical Engineers Code for Class 3 systems in accordance with the 
inservice inspection program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because industry experience has shown that an inservice leak test 
conducted at normal operating temperature and pressure is just as 
effective at finding leakage as a hydrostatic test conducted at 110 
percent of the design pressure. Therefore, there is no increase in 
the probability or consequences of previously evaluated accidents. 
Also, note that the diesel generator fuel oil system is not 
specifically modeled in the St. Lucie probability safety assessment 
(PSA). Based on the St. Lucie PSA, the diesel generator failure 
probability is dominated by failure modes other than fuel oil pipe 
rupture. The total diesel generator failure probability is on the 
order of 1E-2, with the contribution from fuel oil pipe rupture on 
the order of 1E-5 (i.e., three orders of magnitude below the EDG 
[emergency diesel generator] failure probability).
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    The use of the modified specifications can not create the 
possibility of a new or different kind of accident from any 
previously evaluated since the proposed amendments provide an 
alternative method of leak detection for the required 10-year 
inservice inspection. They do not result in an operational condition 
different from that which has already been considered by TS. 
Therefore, the changes do not create the possibility of a new or 
different kind of accident or malfunction.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The alternative method of leak detection has no impact on the 
consequences of any analyzed accident and does not significantly 
change the failure probability of equipment that provides protection 
for the health and safety of the public. Therefore, there is no 
significant decrease 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: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: January 14, 2002.
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Nuclear Power Plant Technical Specifications (TS) 
3.10.f, ``Inoperable Rod Position Indicator Channels,'' to provide an 
allowed outage time (AOT) for the Individual Rod Position Indicator 
(IRPI) system of 24 hours with more than one IRPI per group inoperable. 
The TS did not previously have an explicit AOT for this condition. In 
addition, the proposed amendment would reformat TS 3.10.f using 
MicroSoft Word to more closely resemble the format of Improved Standard 
Technical Specification (ISTS) to improve clarity.

[[Page 7420]]

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

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The format changes are administrative in nature and therefore 
have no effect on the probability or consequences of an accident. 
The Individual Rod Position Indicator (IRPI) System is not an 
accident initiator. Therefore, any change to the system would not 
effect the probability of an accident previously evaluated. The risk 
of core damage/release of radioactivity would not increase with the 
other reactor condition monitors still functional along with the 
plant mode remaining the same.
    The proposed changes provide more time to troubleshoot and 
restore the system, which would keep the reactor in a steady state 
condition, rather than to challenge the plant with a reduction in 
power. The addition of hourly reactor temperature checks as well as 
placing the rod controls to manual are added to temporarily increase 
the surveillance on the reactor due to loss of the IRPI system 
during the IRPI AOT. Since IRPI's are not an accident initiator and 
compensatory measures have been added to ensure rod position is 
known incase one or more IRPIs are inoperable, this amendment does 
not involve an increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The format changes are administrative in nature and therefore 
have no effect on the probability or consequences of a new or 
different kind of accident from any accident previously evaluated. 
The primary function of the IRPI system is to monitor the position 
of each rod and send that information to the control room. A failure 
of this system will not result in an accident.
    The proposed changes do not involve a change to the physical 
plant or operations. Operations currently monitors power tilt, 
excore detectors, thermocouples, and rod movement when IRPIs become 
inoperable. The extra surveillance requirements added by the ISTS 
for the AOT are there to cover the loss of information when the 
IRPIs are OOS. The change to 24 hours for troubleshooting when more 
than one IRPI channel per group is inoperable would therefore not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    The format changes are administrative in nature and therefore 
are not involved in a significant reduction in the margin of safety. 
Margin of safety relates to actual rod position in relation to each 
other. This margin is controlled by rod misalignment requirements. 
The IRPIs provide indication of that position which the operators 
have other means of determining should the need arise. The 
implementation of this proposed amendment ensures continued close 
monitoring of rod position but also adds hourly documentation of the 
reactor coolant temperature requirement as well. The proposed change 
provides more time to troubleshoot and restore the system, which 
would keep the reactor in a steady state condition, rather than to 
challenge the plant with a reduction in power. Therefore, NMC 
concludes that there is not 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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: William D. Reckley, Acting Section Chief.

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant, 
Van Buren County, Michigan

    Date of amendment request: January 28, 2002.
    Description of amendment request: The proposed amendment would 
revise the Core Operating Limits Report (COLR) analytical methods 
referenced in Technical Specification (TS) 5.6.5.b. Specifically, the 
changes would add references to two NRC-approved Framatome ANP, Inc., 
reports: (1) EMF-2310(P)(A), Revision 0, ``SRP [Standard Review Plan] 
Chapter 15 Non-LOCA [loss-of-coolant accident] Methodology for 
Pressurized Water Reactors [PWRs],'' dated May 2001, and (2) EMF-
2328(P)(A), Revision 0, ``PWR Small Break LOCA Evaluation Model, S-
RELAP5 Based,'' dated March 2001. Existing references in TS 5.6.5.b 
describing Exxon Nuclear Company's large-break LOCA evaluation model 
would be deleted.
    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:

    Nuclear Management Company has evaluated whether or not a 
significant hazards consideration is involved with the proposed 
amendment by focusing on the three standards set forth in 10 CFR 
50.92, ``Issuance of Amendment.'' The following evaluation supports 
the finding that operation of the facility in accordance with the 
proposed change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed license amendment removes a safety analysis 
methodology and adds two new safety analysis methodologies in TS 
5.6.5.b. Accidents previously evaluated will be unaffected because 
they will continue to be analyzed using applicable methodologies 
approved by the Nuclear Regulatory Commission to ensure all required 
safety limits are met. The proposed amendment does not affect the 
acceptance criteria for loss-of-coolant accidents (LOCA) or non 
loss-of-coolant accidents. As such, the proposed amendment does not 
increase the probability or consequences of an accident. The 
proposed amendment does not involve operation of the required 
structures, systems or components (SSCs) in a manner or 
configuration different from those previously recognized or 
evaluated.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed amendment does not involve a physical alteration of 
any SSC or a change in the way any SSC is operated. The proposed 
amendment does not involve operation of any required SSCs in a 
manner or configuration different from those previously recognized 
or evaluated. No new failure mechanisms will be introduced by 
changes being requested.
    Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed amendment does not, by itself, introduce a failure 
mechanism. The proposed amendment does not involve any physical 
changes to the plant or manner in which the plant was operated. The 
proposed changes do not affect the acceptance criteria for loss-of-
coolant or non-loss-of-coolant accidents. All required safety limits 
will continue to be analyzed using methodologies approved by the 
Nuclear Regulatory Commission.
    Therefore, the proposed amendment would not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: William D. Reckley (Acting).

[[Page 7421]]

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

    Date of amendment request: January 4, 2002.
    Description of amendment request: The proposed amendment would add 
Technical Specification (TS) 3/4.3.10, ``Mechanical Vacuum Pump Trip 
Instrumentation.''
    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's review is presented below:

    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed amendment would add a new TS section for the 
instrumentation that provides the automatic tripping of the 
mechanical vacuum pumps when high radiation is detected in the main 
steamlines. The proposed change has no effect on any structures, 
systems, or components (SSCs) since the mechanical vacuum pump 
automatic trip function is already part of the existing plant 
design. The new TS requirements are being added because automatic 
tripping of the mechanical vacuum pumps is credited for mitigating 
the radiological consequences of a control rod drop accident (CRDA), 
and, as such, a TS LCO must be established to meet the requirements 
stated in 10 CFR 50.36(c)(2)(ii), Criterion 3. Since the proposed 
change only establishes TS requirements for an existing function, 
and there are no effects to any SSCs, there is no impact on the CRDA 
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?
    No. The proposed amendment does not change the design function 
or operation of any SSCs. Plant operation will not be affected by 
the proposed change and no new failure mechanisms, malfunctions, or 
accident initiators will be created. 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?
    No. The proposed amendment only establishes TS requirements for 
an existing function and will not change any plant operating 
parameters. The licensee's submittal stated that the proposed change 
does not affect the radiological consequences for a CRDA. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.

    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment requests: March 21, 2001 as superseded by letter 
dated October 24, 2001.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification (TS) 5.5.2.12 pertaining to ventilation 
filter testing program (VFTP). Specifically, the proposed amendments 
would: (a) in TS 5.5.2.12 lead paragraph, and TS 5.5.2.12d, delete 
reference to Regulatory Guide (RG) 1.52 and American Society of 
Mechanical Engineers (ASME) N510-1989; the testing frequency will 
continue to be in accordance with RG 1.52, Revision 2; (b) in TS 
5.5.2.12a and TS 5.5.2.12b, refer to American National Standards 
Institute (ANSI) Code N510-1975 instead of ASME N510-1989, and add a 
note to provide clarification regarding high energy particulate air 
(HEPA) filter qualification and testing methodology; and (c) in TS 
5.5.2.12c, include specific temperature and relative humidity for 
laboratory testing of charcoal adsorber samples.
    The October 24, 2001, submittal supercedes in its entirety the 
licensee's March 21, 2001, submittal which was previously noticed in 
the Federal Register on April 18, 2001.
    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 operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed change is to change the reference to ASME Code in 
subsection 5.5.2.12.a and 5.5.2.12.b from ASME N510-1989 to ANSI 
N510-1975. Technical Specification (TS) 3.7.11, ``Control Room 
Emergency Air Cleanup System'' (CREACUS), Surveillance Requirement 
(SR) 3.7.11.2 and TS 3.7.14, ``Fuel Handling Building Post-Accident 
Cleanup Filter System'' (PACU), SR 3.7.14.2 requires CREACUS and 
PACU filter testing in accordance with the Ventilation Filter 
Testing Program (VFTP).
    SONGS [San Onofre Nuclear Generating Station] TS 5.5.12.a, 
``Ventilation Filter Testing Program,'' states that the in-place 
HEPA filter testing is performed in accordance with RG 1.52, 
Revision 2 and ASME N510-1989. The discrepancy arises because the 
HEPA filter testing method used at SONGS does not entirely meet the 
methodology which are delineated in ASME N510-1989. In particular, 
the CREACUS in-place HEPA filter testing uses a method (Alternate 
Shroud Test) which is no longer specified in ASME N510-1989. But 
this method is specified in ANSI N510-1975 and was used when the 
plant was licensed. In addition, the PACU in-place HEPA filter 
testing methodology which is employed at SONGS, has a downstream 
point location which differs from the location suggested in ASME 
N510-1989.
    ANSI N510-1975, while providing a suggestion where downstream 
sample could be located, nevertheless does not provide a specific 
location. The test acceptance criteria are the same for methods 
cited in ANSI N510-1975 and ASME N510-1989. The method which is 
employed at SONGS provides more conservative results because the 
test is performed on individual HEPA filters, which ensures that 
each of the HEPA filters in the tested bank meets the acceptance 
criteria, as compared to the method suggested in ASME N510-1989.
    The locations of the PACU HEPA downstream sample points are 
different from the locations suggested in ASME N510-1989, though 
they meet the requirements delineated in ANSI N510-1975. ANSI N510-
1975 requires that a single representative downstream sample point 
be established, if possible, at the location where adequate mixing 
may be achieved, or at a point downstream of a fan, or multiple 
downstream sampling points may be used (such as in the Alternate 
Shroud Technique used in the CREACUS system) if a single downstream 
sample point is not feasible.
    Since the HEPA filters are tested to the same acceptance 
criteria, and the testing methodology is permitted by ANSI N510-
1975, to which the plant was licensed, it is concluded, that the 
proposed change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Section 5.5.2.12.c will be modified by specifying the 
temperature and relative humidity for laboratory testing of charcoal 
adsorber samples. This modification clarifies that samples shall be 
obtained in accordance with RG 1.52, Revision 2, and tested per 
methodology of ASTM D3803-1989, at 30'C and relative humidity of 
70%. This clarification eliminates possible misinterpretation of the 
current wording.
    The proposed change will also include Note 1 which clarifies the 
inplace testing of charcoal adsorbers and HEPA filters. Based on the 
provisions of section 10.4 of ASME N510-1989, this Note allows 
replacement of DOP [dioctyl phthalate] with a suitable alternate.
    The proposed change also clarifies the statement of subsection 
5.5.2.12.d. Pressure drop testing across combined HEPA filters, the 
prefilters, and the charcoal adsorbers is

[[Page 7422]]

industry-wide practice which is based on good engineering practice 
and operating experience. This change will not increase the 
probability or consequences of an accident previously evaluated.
    Therefore, the probability or consequences of an accident 
previously evaluated will not be increased by operating the facility 
in accordance with this proposed change.
    (2) Will operation of the facility in accordance with this 
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 change the design or configuration 
of the plant. The proposed change is to change the reference to ASME 
Code in subsection 5.5.2.12.a and 5.5.2.12.b from ASME N510-1989 to 
ANSI N510-1975 to reflect the standard used. Section 5.5.2.12.c will 
be modified by specifying the temperature and relative humidity for 
laboratory testing of charcoal adsorber samples. This is done for 
clarification purposes. Also, subsection 5.5.2.12.d will be changed 
by deleting the references to RG 1.52, Revision 2, and ASME N510-
1989 regarding pressure drop test across HEPA filters. RG 1.52, 
Revision 2 and ASME N510-1989 do not require pressure drop test[ing] 
across HEPA filters.
    Therefore, this proposed change will not create the possibility 
of a new or different kind of accident from any accident that has 
been previously evaluated.
    (3) Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed change is to change the reference to ASME Code in 
subsections 5.5.2.12.a, and 5.5.2.12.b from ASME N510-1989 to ASME 
N510-1975. The CREACUS units HEPA filters are currently tested to 
N510-1975. Although the test methodology is slightly different than 
that in N510-1989, the acceptance criteria are the same and the 
current methodology is conservative. Thus the current testing 
satisfies the acceptance criteria of N510-1989, even though the test 
method is different. Section 5.5.2.12.c will be clarified by 
specifying the temperature and relative humidity for laboratory 
testing of charcoal adsorber samples.
    The current methodology for HEPA filter testing will not change 
as a result of the proposed change. Also, deletion of references to 
RG 1.52, Revision 2 and ASME N510-1989 from subsection 5.5.2.12.d 
clarifies this section because these standards do not require HEPA 
filters pressure drop test. Consequently, there is no change to the 
design or operation of the plant as a result of this change.
    Therefore, the operation of the facility in accordance with this 
proposed change will not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendment: April 4, 2001, as supplemented 
on October 12, November 28, November 30, December 7, and December 20, 
2001.
    Brief description of amendment: The amendment deletes Technical 
Specifications (TSs) 5.3.1.B and 5.3.1.C. These TSs restricted the 
handling of heavy loads over irradiated fuel stored in the spent fuel 
pool. The basis for deleting these TSs is the upgrade of the reactor 
building crane and associated handling systems to a single-failure 
proof system.
    Date of Issuance: January 23, 2002.
    Effective date: January 23, 2002, and shall be implemented within 
60 days of issuance.
    Amendment No.: 223.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 31702).
    The supplemental letters dated October 12, November 28, November 
30, December 7, and December 20, 2001, provided clarifying information 
within the scope of the original application and did not change the NRC 
staff's initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated January 23, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 9, 2001.
    Brief description of amendment: The amendment incorporated changes 
to the Technical Specifications (TSs) to provide consistency with the 
changes to 10 CFR 50.59, ``Changes tests, and experiments,'' as 
published in the Federal Register (FR) (64 FR 53582), dated October 4, 
1999. Specifically, the changes replace the term ``safety evaluation'' 
with ``10 CFR 50.59 evaluation'' and ``unreviewed safety question'' 
with ``requires NRC approval pursuant to 10 CFR 50.59.''
    Date of Issuance: January 28, 2002.
    Effective date: January 28, 2002, and shall be implemented within 
60 days of issuance.

[[Page 7423]]

    Amendment No.: 224.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44162).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated January 28, 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: April 4, 2001.
    Brief description of amendments: The amendments revise Technical 
Specifications (TSs) 3.3.12 and 3.9.2 and associated bases pages to (1) 
clarify operability requirements for the boron dilution alarm system 
(BDAS) by adding Mode 6 applicability to TS 3.3.12, (2) ensure 
appropriate operator action when the BDAS is declared inoperable by 
adding a note to TS 3.9.2 and (3) delete Action 3.9.2.B.2.
    Date of issuance: January 29, 2002.
    Effective date: January 29, 2002, and shall be implemented within 
45 days of the date of issuance.
    Amendment Nos.: Unit 1--138, Unit 2--138, Unit 3--138.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 2, 2001 (66 FR 
22024). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 29, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 7, 2001, as supplemented 
June 29, 2001.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.4.3 and the associated Surveillance Requirement 
(SR) to eliminate the pressurizer water volume value in the 
specification and change ``volume'' to ``level'' in TS 3.4.3, SR 4.4.3, 
and the associated Bases.
    Date of issuance: February 5, 2002.
    Effective date: February 5, 2002.
    Amendment No. 109.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 25, 2001 (66 FR 
38760).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 5, 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: November 11, 2001.
    Brief description of amendment: The amendment revises Technical 
Specification Surveillance Requirement (SR) 3.0.3 to allow a longer 
period of time before entering a limiting condition for operation in 
the event of a missed surveillance. The time is extended from the 
current limit of ``* * * up to 24 hours or up to the limit of the 
specified Frequency, whichever is less'' to ``* * * up to 24 hours or 
up to the limit of the specified Frequency, whichever is greater.'' In 
addition, the following requirement 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: January 25, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 145.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64289).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 25, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 26, 2001.
    Brief description of amendment: The amendment approves a change to 
the Technical Specifications (TSs) and Bases related to reactor coolant 
pump flywheel inspection requirements and reactor coolant system 
structural integrity. The changes add Section 6.22, ``Reactor Coolant 
Pump Flywheel Inspection Program'' to the TSs and relocate the 
requirements of TS 3/4.4.10, ``Reactor Coolant System, Structural 
Integrity'' to the Technical Requirements Manual.
    Date of issuance: February 1, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 264.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 30, 2001 (66 FR 
29351).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 1, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 22, 2001, as supplemented 
by letter dated October 11, 2001.
    Brief description of amendments: The amendments revised the current 
MNS Technical Specifications (TS) surveillance requirement (SR) for the 
methodology and frequency for the chemical analyses of the ice 
condenser ice bed. Also, these amendments add a new TS SR to address 
sampling requirements for ice additions to the ice bed. In addition, 
the amendments revise the current MNS TS acceptance criteria and 
surveillance frequency for the inspection of ice condenser ice basket 
flow channel areas.
    Date of issuance: February 1, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 201 and 182.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 11, 2001 (66 FR 
36339).
    The supplement dated October 11, 2001, provided clarifying 
information that did not change the scope of the March 22, 2001, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 1, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 20, 2001.

[[Page 7424]]

    Brief description of amendment: The amendment revised Technical 
Specification Section 6.8.4.a to delete the requirements to have a 
program to obtain and analyze samples of reactor coolant and 
containment atmosphere under accident conditions (post accident 
sampling system).
    Date of issuance: January 30, 2002.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 222.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55013).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 30, 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: October 23, 2001.
    Brief description of amendment: The amendment deletes Technical 
Specification 5.5.3, ``Post Accident Sampling'' and thereby eliminates 
the requirements to have and maintain the Post Accident Sampling 
System.
    Date of issuance: February 6, 2002.
    Effective date: As of the date of issuance to be implemented within 
90 days.
    Amendment No.: 210.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64293).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 6, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 20, 2001, as 
supplemented by letters dated July 13, 2001, and December 28, 2001.
    Brief description of amendments: The amendments change the 
Technical Specifications (TS) Section 3.8.1, ``A.C. Sources-
Operating,'' to extend to 14 days the allowable completion time for the 
required actions associated with restoration of an inoperable Division 
1 or Division 2 Emergency Diesel Generator. In addition, the amendments 
change the TS completion time period associated with discovery of 
failure to meet TS limiting condition of operation 3.8.1 from 10 days 
to 17 days.
    Date of issuance: January 30, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 150 and 136.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 21, 2001 (66 FR 
15925).
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 30, 2002.
    No significant hazards consideration comments received: No.

Exelon Generation Company, Docket No. 50-352, Limerick Generating 
Station, Unit 1, Montgomery County, Pennsylvania

    Date of application for amendment: September 14, 2001.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Figure 3.4.6.1-1, ``Minimum Reactor Vessel Metal 
Temperature vs. Reactor Vessel Pressure,'' to extend the use of the 
reactor pressure vessel (RPV) pressure-temperature limit curves for one 
additional fuel cycle and approves a modification to the TS Table 
4.4.6.1.3-1, ``Reactor Vessel Surveillance Program--Withdrawal 
Schedule,'' RPV surveillance capsule withdrawal schedule.
    Date of issuance: January 30, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 155.
    Facility Operating License No. NPF-39: This amendment revised the 
TS.
    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 59506).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 30, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 28, 2001, as supplemented 
May 1, 2001, and June 13, 2001.
    Brief description of amendments: These amendments relocate certain 
Beaver Valley technical specifications (TSs) to the Licensing 
Requirements Manual and the Offsite Dosage Calculation Manual. The 
major change proposed in this request involves the application of the 
TS screening criteria of 10 CFR 50.36.
    Date of issuance: January 24, 2002.
    Effective date: As of date of issuance and shall be implemented 
within 120 days.
    Amendment Nos.: 246 and 124.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 20, 2001 (66 FR 
33111).
    The May 1 and June 13, 2001, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the scope of the initial 
Federal Register notice. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated January 24, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 29, 2001, as 
supplemented December 17, 2001.
    Brief description of amendments: These amendments revised Technical 
Specification Section 3.9.3 to reduce the minimum decay time required 
prior to fuel movement from 150 hours to 100 hours.
    Date of issuance: January 29, 2002.
    Effective date: Upon issuance and shall be implemented within 60 
days.
    Amendment Nos.: 247 and 126.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 26, 2001 (66 
FR 66465). The December 17, 2001, letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the application beyond 
the scope of the original Federal Register notice.

[[Page 7425]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 29, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: January 18, 2001, as 
supplemented by letters dated February 20, June 9, June 26, June 29, 
October 31, and December 19, 2001.
    Brief description of amendments: The amendments changed the 
technical specifications associated with modifying the maximum power 
levels permissible with inoperable main steam safety valves.
    Date of issuance: January 29, 2002.
    Effective date: Effective as of the date of issuance and shall be 
implemented within 60 days.
    Amendment Nos.: 248 and 127.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 27, 2001 (66 FR 
39211).
    The February 20, June 9, June 26, June 29, October 31, and December 
19, 2001, letters provided clarifying information that did not change 
the initial proposed no significant hazards consideration determination 
or expand the amendment beyond the scope of the original notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 29, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 25, 2001.
    Brief description of amendment: The amendment approved increases to 
the BVPS-2 Technical Specification boron concentration limits for the 
refueling water storage tank, accumulators, and the reactor coolant 
system/refueling canal during Mode 6.
    Date of issuance: January 28, 2002.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment No: 125.
    Facility Operating License No. NPF-73. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 3, 2001 (66 FR 
50468).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 28, 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: February 21, 2001, as 
supplemented April 26, 2001.
    Brief description of amendment: The amendment revised the Improved 
Technical Specification 3.3.8, to clarify actions to be taken in the 
event that one or more channels of the loss of voltage or degraded 
voltage Emergency Diesel Generator start functions become inoperable.
    Date of issuance: January 29, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 202.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 21, 2001 (66 FR 
15925). The 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 January 29, 2002.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of application for amendments: October 17, 2001.
    Brief description of amendments: The amendments revised Technical 
Specification 6.8.4.h, to allow a one-time change in the containment 
integrated leakage rate test interval from the required 10 years to a 
test interval of 15 years.
    Date of issuance: January 29, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos: 218 and 212.
    Facility Operating License Nos.DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 59507).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 29, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 17, 2001.
    Brief description of amendments: The amendments revise Technical 
Specification 4.0.3 and its associated Bases to provide for a delay 
period in which to perform a surveillance which has been discovered not 
to have been performed within its specified frequency.
    Date of issuance: January 28, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 263 and 245.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44175).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 28, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 7, 2001.
    Brief description of amendments: The amendments would create 
Technical Specification (TS) 3.0.6 and associated bases to allow 
equipment that was removed from service or declared inoperable to be 
returned to service under administrative controls solely to perform the 
testing required to demonstrate its operability or the operability of 
other equipment.
    Date of issuance: February 1, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 264 and 246.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 59508).

[[Page 7426]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 1, 2002.
    No significant hazards consideration comments received: No.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of application for amendment: April 11, 2001.
    Brief description of amendment: The proposed change would revise 
Technical Specifications 4.2, Fuel Storage, and 5.6.5, Spent Fuel Pool 
Water Chemistry Program, by adding applicability statements that 
specify that these specifications apply only when irradiated fuel is 
stored in the spent fuel storage pool. These changes are being made to 
facilitate dismantlement of the spent fuel storage pool upon removal of 
the last irradiated fuel assembly from the spent fuel storage pool to 
the onsite independent spent fuel storage installation (ISFSI).
    Date of issuance: February 6, 2002.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 166.
    Facility Operating License No. DPR-36: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41623).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 6, 2002.
    No significant hazards consideration comments received: No.

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 application for amendments: November 19, 2001.
    Brief description of amendments: The amendment changes Surveillance 
Requirement (SR) 3.0.3 to allow a longer period of time before entering 
a Limiting Condition of Operations in the event of a missed 
surveillance. The time is extended from the current limit of ``* * * up 
to 24 hours or up to the limit of the specified Frequency, whichever is 
less'' to `` * * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is greater.'' In addition, the following 
requirement 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: January 28, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 202 and 207.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 59510).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 28, 2002.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket No. 50-387, Susquehanna Steam Electric 
Station, Unit 1, Luzerne County, Pennsylvania

    Date of application for amendment: May 31, 2001, as supplemented 
December 5, 2001.
    Brief description of amendment: The amendment revised the minimum 
critical power ratio safety limits.
    Date of issuance: January 31, 2002.
    Effective date: As of date of issuance and shall be implemented 
upon startup following the Unit 1 twelfth refueling and inspection 
outage.
    Amendment No.: 199.
    Facility Operating License No. NPF-14: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 5, 2001 (66 
FR 46480). The supplemental letter provided additional information but 
did not change the initial no significant hazards consideration 
determination or expand the amendment beyond the scope of the initial 
notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 31, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 28, 2000.
    Brief description of amendments: The amendments expanded the 
allowable suppression chamber-to-drywell vacuum breaker setpoint range.
    Date of issuance: January 29, 2002.
    Effective date: As of date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 198 and 173.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 10, 2001 (66 FR 
2023).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 29, 2002.
    No significant hazards consideration comments received: No.

Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco 
Nuclear Generating Station, Sacramento County, California

    Date of application for amendment: May 21, 2001.
    Brief description of amendment: The amendment deletes the 
Definitions, Limiting Conditions for Operation and Surveillance 
Requirements, and several sections from the Administrative Controls 
portion of the technical specifications once the spent nuclear fuel has 
been transferred from the 10 CFR part 50 licensed site to the 10 CFR 
part 72 licensed Independent Spent Fuel Storage Installation.
    Date of issuance: February 5, 2002.
    Effective date: February 5, 2002, to be implemented within 30 days 
after the transfer of the last cask of spent nuclear fuel from the 
spent fuel pool to the Independent Spent Fuel Storage Installation is 
complete.
    Amendment No.: 129.
    Facility Operating License No. DPR-54: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 11, 2001 (66 FR 
36343).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 5, 2002.
    No significant hazards consideration comments received: No.

Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco 
Nuclear Generating Station, Sacramento County, California

    Date of application for amendment: June 7, 2001.
    Brief description of amendment: The amendment deletes 
administrative requirements which are no longer applicable once the 
spent nuclear fuel has been transferred from the 10 CFR part 50 
licensed site to the 10 CFR part 72 licensed Independent Spent Fuel 
Storage Installation. Other requirements are transferred to the Rancho 
Seco Quality Manual.
    Date of issuance: February 5, 2002.
    Effective date: February 5, 2002, to be implemented within 30 days 
after the transfer of the last cask of spent nuclear fuel from the 
spent fuel pool to the Independent Spent Fuel Storage Installation is 
complete.

[[Page 7427]]

    Amendment No.: 130.
    Facility Operating License No. DPR-54: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 11, 2001 (66 FR 
36344).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 5, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 7, 2001.
    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 upon a missed SR 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: February 5, 2002.
    Effective date: February 5, 2002, and shall be implemented within 
60 days of the date of issuance.
    Amendment No.: 147.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64307).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 5, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 7, 2001 (ULNRC-04557).
    Brief description of amendment: The amendment revised Surveillance 
Requirements (SRs) 3.3.1.2 and 3.3.1.3 in the Technical Specifications 
(TSs) on reactor trip system (RTS) instrumentation. The change to SR 
3.3.1.2 replaces the reference to the nuclear instrumentation system 
(NIS) channel output by a reference to the power range channel output 
and deletes Note 1 to the SR. The change to SR 3.3.1.3 is editorial.
    Date of issuance: February 5, 2002.
    Effective date: February 5, 2002, and shall be implemented, 
including adding the changes to the Bases of the Technical 
Specifications as described in the licensee's application of November 
7, 2001, before the startup from refueling outage 12, which is 
scheduled for the Fall of 2002.
    Amendment No.: 148.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64308).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 5, 2002.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket No. 50-338, North Anna 
Power Station, Unit 1, Louisa County, Virginia Date of application for 
amendment: January 9, 2001.

    Brief description of amendment: This amendment revises the Facility 
Operating License (FOL) and Technical Specifications (TS) to remove 
obsolete license conditions, make editorial changes in the FOL, 
relocate license conditions, and remove redundant license conditions 
covered elsewhere in the license.
    Date of issuance: January 31, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 230.
    Facility Operating License No. NPF-4: Amendment changes the FOL and 
TS.
    Date of initial notice in Federal Register: February 21, 2001 (66 
FR 11064).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 31, 2002.
    No significant hazards consideration comments received: No.

    For the Nuclear Regulatory Commission.
    Dated at Rockville, Maryland, this 11th day of February 2002.
John A. Zwolinski,
 Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.

[FR Doc. 02-3750 Filed 2-18-02; 8:45 am]
BILLING CODE 7590-01-P