[Federal Register Volume 67, Number 102 (Tuesday, May 28, 2002)]
[Notices]
[Pages 36924-36939]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-13082]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from May 3, 2002 through May 16, 2002. The last 
biweekly notice was published on May 14, 2002 (67 FR 34481).

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

[[Page 36925]]

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 June 27, 2002, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714, which is 
available at the NRC's PDR, located at One White Flint North, 11555 
Rockville Pike (first floor), Rockville, Maryland. Publicly available 
records will be accessible from the Agencywide Documents Access and 
Management Systems (ADAMS) Public Electronic Reading Room on the 
internet at the NRC web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or an Atomic 
Safety and Licensing Board, designated by the Commission or by the 
Chairman of the Atomic Safety and Licensing Board Panel, will rule on 
the request and/or petition; and the Secretary or the designated Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland, by the above date. A copy of the petition should 
also be sent to the Office of the General Counsel, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and to the attorney 
for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. Publicly available records 
will be accessible from the Agencywide Documents Access and Management 
Systems (ADAMS) Public Electronic Reading Room on the internet at the 
NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not 
have access to ADAMS or if there are problems in accessing the 
documents located in ADAMS, contact the NRC PDR Reference staff at 1-
800-397-4209, 304-415-4737 or by e-mail to [email protected].

[[Page 36926]]

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

    Date of amendment request: April 26, 2002.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs), Section 2.3, ``Limiting 
Safety System Settings,'' Section 3.1, ``Protective Instrumentation,'' 
and Section 3.10, ``Core Limits,'' to reflect a methodology to assure 
coupled neutronic/thermal-hydraulic instabilities are adequately 
detected and suppressed. This methodology is identified as Option II by 
the Boiling Water Reactor (BWR) Owners Group. The proposed amendment 
includes technical (i.e., limiting safety system settings) and 
editorial changes, and is associated with the average power range 
monitoring (APRM) system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff reviewed the licensee's analysis and has 
performed its own, which is presented below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The APRM neutron monitoring system is not an initiator of, or a 
precursor to, a previously evaluated accident. The APRM system monitors 
the power level of the reactor core and provides automatic core 
protection signals in the event of a power transient. The revised 
requirements will result in a reactor scram, should one be needed, 
sooner than under the current requirements. These revised requirements 
do not lead to, and are not results of, physical design modifications. 
The APRM and other systems associated with the proposed TS requirements 
will thus continue to perform their functions as originally designed. 
Therefore, this amendment does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the amendment create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed amendment does not affect accident initiators or 
precursors because it does not alter any design feature, reactor fuel 
safety limit, equipment configuration, or manner in which the unit is 
operated. Further, it does not alter or prevent the ability of 
structures, systems, or components to perform their intended safety or 
accident mitigating functions. Accordingly, the proposed amendment does 
not create a new or different kind of accident from any accident 
previously evaluated.
    3. Does the amendment involve a significant reduction in a margin 
of safety?
    No. The proposed amendment does not change any design feature, 
analysis methodology, safety limits or acceptance criteria. The APRM 
system under the revised requirements will continue to perform its 
design functions. Therefore, the proposed amendment does not involve a 
significant reduction in a margin of safety.
    Based on the NRC staff's review, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the proposed amendment involves no 
significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Richard J. Laufer.

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

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

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

    Date of amendment request: April 10, 2002.
    Description of amendment request: The proposed changes to the 
Technical Specifications (TSs) will relocate the emergency diesel 
generator (EDG) 24-month maintenance inspection requirements to 
licensee-controlled documents, i.e., either the Updated Final Safety 
Analysis Report (UFSAR) or the Technical Requirements Manual as 
appropriate, either of which would be controlled in accordance with the 
requirements of 10 CFR 50.59.
    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 amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed TS changes are administrative changes to relocate 
the current EDG maintenance inspection requirements.
    The EDGs are designed to provide a reliable alternative source 
of AC [alternating current] electrical power in the event of an 
accident coincident with the loss of offsite power. The failure of 
an EDG itself is not considered as an accident evaluated in the 
UFSAR. The proposed administrative changes to relocate the 
maintenance inspection requirements do not affect the current 
accident initiators or precursors that could lead to a previously 
evaluated accident.
    The failure of a single EDG to respond when required to mitigate 
the consequences of an accident has already been considered as a 
subsequent single failure in the current plant safety analyses. The 
proposed administrative changes to relocate the maintenance 
inspection requirements do not alter the EDG design features, 
operation, or accident analysis assumptions which could affect the 
ability of the EDGs to mitigate the consequences of a previously 
evaluated accident. Current TS testing requirements for the EDGs, 
e.g., starting, timing, loading, and sequencing will continue to 
ensure reliable EDG operation and are not being changed in this 
request.
    Since only the relocation of EDG maintenance inspection 
requirements is involved, the proposed changes will not increase the 
likelihood of the malfunction of another system, structure or 
component which has been assumed as an accident initiator or 
credited in the mitigation of an accident.
    Based on the above discussion, the proposed TS changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    [2.] The proposed amendments do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The EDGs are designed to provide a reliable alternative source 
of AC electrical power in the event of an accident coincident with 
the loss of offsite power. The proposed TS changes are 
administrative changes to relocate the EDG maintenance inspection 
requirements.
    No change in the ability to perform the design function of the 
EDGs is involved. No change in the operation of the EDGs is 
required. Instrumentation setpoints, starting, sequencing, and 
loading functions associated with the EDGs are not affected by the 
proposed changes. No modifications to the EDGs are required to 
implement the proposed TS changes. Therefore, no new failure 
mechanism, malfunction, or accident initiator is considered 
credible.
    Additionally, the proposed TS changes do not affect other plant 
design, hardware, system operation, or procedures. Therefore, based 
on the above discussion, the proposed TS changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

[[Page 36927]]

    [3.] The proposed amendments do not involve a significant 
reduction in a margin of safety.
    The proposed TS changes are administrative changes to relocate 
the current EDG maintenance requirements.
    The consideration of safety margins for this amendment included 
a review of the acceptance criteria for emergency core cooling 
systems for light water nuclear power reactors in 10 CFR 50.46, and 
ECCS [emergency core cooling system] evaluation models in Appendix K 
to 10 CFR [Part] 50. The proposed amendments do not involve a 
relaxation of the criteria used to establish the safety limits, a 
relaxation of the bases for the limiting safety system setting, nor 
a relaxation of the bases for the limiting conditions for operation.
    Controlling values for the EDGs are included in current TS 
testing requirements, e.g., EDG starting, timing, loading, and 
sequencing. The proposed amendment will not modify these 
requirements or the accident analysis assumptions regarding the 
performance of the EDGs which could potentially challenge safety 
margins established to ensure fuel cladding integrity, as well as 
reactor coolant and containment system integrity.
    The safety analyses of the EDGs' ability to mitigate accidents 
do not require revision in order to implement the proposed 
amendment[s]. Modification of the existing margins is not required.
    Based on the above discussion, the proposed TS changes do not 
involve a significant reduction in a margin of safety.

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

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

    Date of amendments request: January 31, 2002.
    Description of amendments request: The proposed amendments would 
correct several administrative errors to Technical Specifications 
Sections 5.6.5.b and Appendix B. The changes would correct the title of 
a topical report, the date of issuance of a report, and the name of the 
state agency that issues pollution discharge elimination system 
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. Would Not Involve a Significant Increase in the Probability or 
Consequences of an Accident Previously Evaluated

    Technical Specification 5.6.5.b, Item 16 is being corrected to 
change the title of the publication for CENPD-382-P-A from ``C-E 
Methodology for Core Designs Containing Erbuim Burnable Absorbers'' 
to ``Methodology for Core Designs Containing Erbuim Burnable 
Absorbers.'' Correction of an administrative error does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    Technical Specification 5.6.5.b, Item 19 is being corrected to 
change the date of publication for CEN-161-(B)-P, Supplement 1-P, 
``Improvements to Fuel Evaluation Model'' from April 1989 to April 
1986. Correction of an administrative error does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Technical Specification Appendix B, Page 2-1 (both units) last 
paragraph is being corrected to change the State of Maryland 
Department of Health and Mental Hygiene to the Maryland Department 
of the Environment. The revision of a state organizational title to 
accurately reflect administrative changes made to that organization, 
does 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 [Kind] of 
Accident From Any Accident Previously Evaluated

    Technical Specification 5.6.5.b, Item 16 is being corrected to 
change the title of publication for CENPD-382-P-A from ``C-E 
Methodology for Core Designs Containing Erbuim Burnable Absorbers'' 
to ``Methodology for Core Designs Containing Erbuim Burnable 
Absorbers.'' Correction of an administrative error will not create 
the possibility of a new or different [kind] of accident from any 
accident previously evaluated.
    Technical Specification 5.6.5.b, Item 19 is being corrected to 
change the date of publication for CEN-161-(B)-P, Supplement 1-P, 
``Improvements to Fuel Evaluation Model'' from April 1989 to April 
1986. Correction of an administrative error will not create the 
possibility of a new or different [kind] of accident from any 
accident previously evaluated.
    Technical Specification Appendix B, Page 2-1 (both units) last 
paragraph is being corrected to change the State of Maryland 
Department of Health and Mental Hygiene to the Maryland Department 
of the Environment. The revision of a state organizational title to 
accurately reflect administrative changes made to that organization 
will not create the possibility of a new or different [kind] of 
accident from any accident previously evaluated.

3. Would Not Involve a Significant Reduction in [a] Margin of Safety

    Technical Specification 5.6.5.b, Item 16 is being corrected to 
change the title of publication for CENPD-382-P-A from ``C-E 
Methodology for Core Designs Containing Erbuim Burnable Absorbers'' 
to ``Methodology for Core Designs Containing Erbuim Burnable 
Absorbers.'' Correction of an administrative error will not involve 
a significant reduction in [a] margin of safety.
    Technical Specification 5.6.5.b, Item 19 is being corrected to 
change the date of publication for CEN-161-(B)-P, Supplement 1-P, 
``Improvements to Fuel Evaluation Model'' from April 1989 to April 
1986. Correction of an administrative error will not involve a 
significant reduction in [a] margin of safety.
    Technical Specification Appendix B, Page 2-1, (both units) last 
paragraph is being corrected to change the State of Maryland 
Department of Health and Mental Hygiene to the Maryland Department 
of the Environment. The revision of a state organizational title to 
accurately reflect administrative changes made to that organization 
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 
amendments 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: Richard J. Laufer.

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

    Date of amendments request: March 25, 2002.
    Description of amendments request: The proposed amendment would 
revise Surveillance Requirement (SR) 3.0.3 to extend the delay period, 
before entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period would be extended from the current limit 
of ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is less'' to `` * * * up to 24 hours or up to the 
limit of the specified Frequency, whichever is greater.'' In addition, 
the following requirement would be added to SR 3.0.3: ``A risk 
evaluation shall be performed for any Surveillance delayed greater than 
24 hours and the risk impact shall be managed.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on June 14, 2001 (66 FR 32400),

[[Page 36928]]

on possible amendments concerning missed surveillances, including a 
model safety evaluation and model no significant hazards consideration 
(NSHC) determination, using the consolidated line item improvement 
process. The NRC staff subsequently issued a notice of availability of 
the models for referencing in license amendment applications in the 
Federal Register on September 28, 2001 (66 FR 49714). The licensee 
affirmed the applicability of the following NSHC determination in its 
application dated March 25, 2002.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

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

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

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

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

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Thomas Koshy, Acting.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant (HBRSEP), Unit No. 2, Darlington County, South Carolina

    Date of amendment request: March 26, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 5.5.16, ``Containment Leakage Rate 
Testing Program,'' to require the performance of a Type A test within 
15 years from the last Type A test, which was performed on April 9, 
1992. The proposed change is supported by a plant-specific risk 
assessment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

    The proposed change to TS 5.5.16 provides a one-time extension 
to the testing interval for Type A (containment integrated leak 
rate) testing. The existing 10-year test interval is based on past 
test performance. The proposed TS change provides a one-time 
extension of the Type A test interval to 15 years for HBRSEP, Unit 
No. 2. The proposed TS change does not involve a physical change to 
the plant or a change in the manner in which the plant is operated 
or controlled. The containment vessel is designed to provide a leak 
tight barrier against the uncontrolled release of radioactivity to 
the environment in the unlikely event of postulated accidents. As 
such, the containment vessel is not considered as the initiator of 
an accident. Therefore, the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The proposed change involves only a one-time change to the 
interval between Type A containment leakage tests. Type B and C 
leakage testing will continue to be performed at the interval 
specified in 10 CFR Part 50, Appendix J, Option A, as currently 
required by the HBRSEP, Unit No. 2, TS. As documented in NUREG-1493, 
``Performance-Based Containment Leakage-Test Program,'' industry 
experience has shown that Type B and C containment leakage tests 
have identified a very large percentage of containment leakage paths 
and that the percentage of containment leakage paths that are 
detected only by Type A testing is very small. In fact, an analysis 
of 144 integrated leak rate tests results, including 23 failures, 
found that none of the failures involved containment liner breach. 
NUREG-1493 also concluded, in part, that reducing the frequency of 
Type A containment leakage rate testing to once per 20 years was 
found to lead to an imperceptible increase in risk. The HBRSEP, Unit 
No. 2, test history and risk-based evaluation of the proposed 
extension to the Type A test interval supports this conclusion. The 
design and construction requirements of the containment vessel, 
combined with the containment inspections performed in accordance 
with the American Society of Mechanical Engineers (ASME) Code, 
Section XI, and the Maintenance Rule (i.e., 10 CFR 50.65) provide a 
high degree of assurance that the containment vessel will not 
degrade in a manner that is detectable only by Type A testing. 
Therefore, the proposed Technical Specification change does not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    Therefore, this change does not involve a significant increase 
in the probability or

[[Page 36929]]

consequences of an accident previously evaluated

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

    The proposed change to TS 5.5.16 provides a one-time extension 
to the testing interval for Type A (containment integrated leak 
rate) testing. The existing 10-year test interval is based on past 
test performance. The proposed TS change will provide a one-time 
extension of the Type A test interval to 15 years for HBRSEP, Unit 
No. 2. The proposed change to the Type A test interval does not 
result in any physical changes to HBRSEP, Unit No. 2. In addition, 
the proposed test interval extension does not change the operation 
of HBRSEP, Unit No. 2, such that a failure mode involving the 
possibility of a new or different kind of accident from any accident 
previously evaluated is created.
    Thus, this change does 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 the 
Margin of Safety

    The proposed change to TS 5.5.16 provides a one-time extension 
to the testing interval for Type A (containment integrated leak 
rate) testing. The existing 10-year test interval is based on past 
test performance. The proposed TS change will provide a one-time 
extension of the Type A test interval to 15 years for HBRSEP, Unit 
No. 2. The NUREG-1493 generic study of the effects of extending 
containment leakage testing found that a 20 year extension for Type 
A leakage testing resulted in an imperceptible increase in risk to 
the public. NUREG-1493 found that, generically, the design 
containment leakage rate contributes a very small amount to the 
individual risk, and that the decrease in Type A testing frequency 
would have a minimal affect on this risk, because most potential 
leakage paths are detected by Type B and C testing.
    The proposed change involves only a one-time extension of the 
interval for Type A containment leakage testing; the overall 
containment leakage rate specified by the HBRSEP, Unit No. 2, 
Technical Specifications is being maintained. Type B and C 
containment leakage testing will continue to be performed at the 
frequency required by the HBRSEP, Unit No. 2, Technical 
Specifications. The regular containment inspections being performed 
in accordance with ASME, Section XI, and the Maintenance Rule (i.e., 
10 CFR 50.65) provide a high degree of assurance that the 
containment will not degrade in a manner that is only detectable by 
Type A testing. In addition, a plant-specific risk evaluation has 
demonstrated that the one-time extension of the Type A leakage test 
interval from 10 years to 15 years results in only a very small 
increase in risk for those accident sequences influenced by Type A 
testing.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Thomas Koshy, Acting.

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

    Date of amendment request: August 24, 2001.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) Limiting Condition for Operation 
(LCO) 3.7.3, ``Control Room Emergency Filtration (CREF) System,'' to 
address a degraded CREF pressure boundary. Specifically, the amendment 
would (1) add a note to the LCO that would allow the CREF pressure 
boundary to be opened under administrative control; (2) add a new 
Condition (B) for two CREF subsystems inoperable due to an inoperable 
control room pressure boundary--the associated Required Action would be 
to restore the control room pressure boundary to operable status and 
the Completion Time would be 24 hours; (3) add the phrase, ``for 
reasons other than Condition (B),'' to the Condition requiring entry 
into LCO 3.0.3 for two CREF subsystems or a nonredundant component or 
portion of the CREF system inoperable in Mode 1, 2, or 3; and (4) 
renumber the remaining existing Conditions and Required Actions of LCO 
3.7.3, as required.
    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

    No. The Control Room Emergency Filtration (CREF) System is not 
assumed to be an initiator of any analyzed accident. Therefore, the 
proposed change does not affect the probability of any accident 
previously evaluated. The proposed change to the CREF Technical 
Specifications would permit the control room pressure boundary to be 
opened intermittently under administrative control. Based on the 
proposed compensatory measures in the form of a dedicated individual 
who is in communication with the control room, and his ability to 
rapidly restore the pressure boundary, the capability to mitigate a 
design basis accident will be maintained. In addition, the proposed 
change adds a new Condition that would allow up to 24 hours to 
restore an inoperable control room pressure boundary to operable 
status and would modify existing Conditions to accommodate the new 
Condition (so as to maintain the requirements of the existing 
Conditions). The proposed change does not involve a significant 
increase in the consequences of an accident previously evaluated 
based on the availability of self-contained breathing apparatus 
equipment to minimize radiological dose due to iodine, and the 
ability to operate at least one CREF subsystem to maintain positive 
pressure or to at least minimize any inflow of air from outside of 
the control room.

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

    No. The proposed change would permit the control room pressure 
boundary to be opened intermittently under administrative control. 
In addition, the proposed change would add a new Condition that 
would permit a 24-hour period to take action to restore an 
inoperable control room pressure boundary to operable status. The 
proposed change does not alter the operation of the plant or any of 
its equipment, introduce any new equipment, or result in any new 
failure mechanisms or single failures. Therefore, this change does 
not create the possibility of a new accident, and does not change 
the way that an analyzed accident will progress.

3. The Change Does Not Involve a Significant Reduction in the Margin of 
Safety

    No. The proposed change would permit the control room pressure 
boundary to be opened intermittently under administrative control. 
In addition, the proposed change would add a new Condition that 
would permit a 24-hour period to take action to restore an 
inoperable control room pressure boundary to operable status. The 
proposed change does not adversely affect the ability of the fission 
product barriers to perform their functions. The only safety-related 
equipment affected by the proposed change is the CREF system. 
Adequate compensatory measures are available to mitigate a breach in 
the CREF control room pressure boundary. The probability of a design 
basis accident that would place demands on the CREF System occurring 
during a period that the control room pressure boundary would be 
allowed to be inoperable have been shown to be negligible for this 
limited period of time. In addition, the proposed change would avoid 
the potential for placing the unit in TS Limiting Condition for 
Operation (LCO) 3.0.3, due solely to a breach in the control room 
pressure boundary. Therefore, this 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

[[Page 36930]]

amendment request involves no significant hazards consideration.
    Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: February 5, 2002, as supplemented on 
March 6, 2002.
    Description of amendment request: This request changes the term in 
the Technical Specifications ``once each REFUELING INTERVAL'' to ``once 
per 24 months'' in several surveillance requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:

1. Involve a Significant Increase in the Probability or Consequences of 
an Accident Previously Evaluated

    The proposed change is to revise the term ``once each REFUELING 
INTERVAL'' to ``once per 24 months'' in several technical 
specification surveillance requirements. These surveillances were 
previously approved for once per 24 months when the term REFUELING 
INTERVAL was defined as once per 24 months. The proposed change does 
not revise any surveillance requirements. The change does not alter 
any regulatory requirement or any acceptance criteria for any 
design-basis accidents described in the Millstone Unit No. 3 Final 
Safety Analysis Report (FSAR). The proposed change does not alter 
the method by which the surveillances are conducted, does not 
involve any physical changes to the plant, and does not modify the 
manner in which the plant is operated. Since the change does not 
change the frequency of surveillance, it cannot affect the 
likelihood or consequences of accidents. Therefore, the change will 
not increase the probability or consequences of an accident 
previously evaluated.

2. Create the Possibility of a New or Different Kind of Accident From 
Any Accident Previously Evaluated

    The proposed change does not involve a physical alteration of 
the plant or change the plant configuration (no new or different 
type of equipment will be installed). The proposed change does not 
require any new or unusual operator actions. The change does not 
alter the way any structure, system, or component functions and does 
not alter the manner in which the plant is operated. The change does 
not introduce any new failure modes and does not change the 
surveillance frequency. Therefore, the proposed change will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.

3. Involve a Significant Reduction in a Margin of Safety

    The proposed change does not change any analyses for the current 
design-basis accidents described in the Millstone Unit No. 3 FSAR. 
Therefore, the proposed change will not result in a reduction in a 
margin of safety.
    Based on this review, it appears that the three standards of 10 
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.

    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: April 15, 2002.
    Description of amendment request: The proposed amendments would 
modify the allowable value and surveillance requirements for reactor 
protection system instrumentation for the reactor vessel steam dome 
pressure--high function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

1. The Proposed TS Changes Do Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated

    The proposed Technical Specifications (TS) changes support a 
design change that upgrades the existing reactor vessel steam dome 
pressure--high instrumentation from pressure switches to analog trip 
units. Analog trip units use a proven technology that is more 
reliable than the existing equipment. The proposed design is 
consistent with a generic design that has been previously reviewed 
and approved by the NRC. Analog trip units are currently used in 
various applications at Dresden Nuclear Power Station (DNPS), 
including the reactor protection system (RPS) low water level scram 
function.
    The proposed TS changes add new channel check and trip unit 
calibration surveillance requirements (SRs), and modify other SRs in 
keeping with the use of pressure transmitters for the reactor vessel 
steam dome pressure--high function. The new SRs are not applicable 
to the existing instrumentation because the current pressure 
switches are non-indicating and do not employ trip units.
    TS requirements that govern operability or routine testing of 
plant instruments are not assumed to be initiators of any analyzed 
event because these instruments are intended to prevent, detect, or 
mitigate accidents. Therefore, these changes will not involve an 
increase in the probability of an accident previously evaluated. 
Additionally, these changes will not increase the consequences of an 
accident previously evaluated because the proposed change does not 
adversely impact structures, systems, or components. The planned 
instrument upgrade is a more reliable design than existing 
equipment. The proposed change establishes requirements that ensure 
components are operable when necessary for the prevention or 
mitigation of accidents or transients. Furthermore, there will be no 
change in the types or significant increase in the amounts of any 
effluents released offsite.
    In summary, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

2. The Proposed TS Changes Do Not Create the Possibility of a New or 
Different Kind of Accident From Any Accident Previously Evaluated

    The proposed changes support a planned instrumentation upgrade 
by incorporating SRs required to ensure operability. The change does 
not adversely impact the manner in which the instrument will operate 
under normal and abnormal operating conditions. Therefore, these 
changes provide an equivalent level of safety and will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated. The changes in allowable values and 
surveillance requirements do not affect the current safety analysis 
assumptions. Therefore, these changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

3. The Proposed TS Changes Do Not Involve a Significant Reduction in a 
Margin of Safety

    The proposed TS changes support a planned instrumentation 
upgrade. The proposed changes do not affect the probability of 
failure or availability of the affected instrumentation. The revised 
allowable values, addition of a channel check and trip unit 
calibration, and revision of other SRs do not affect the analytical 
limit assumed in the safety analyses for the actuation of the 
instrumentation. Therefore, the proposed changes do not result in a 
reduction in the margin of safety.

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

[[Page 36931]]

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

    Date of amendment request: May 1, 2002.
    Description of amendment request: The proposed amendments revise 
the required emergency diesel generator start time limit specified in 
Technical Specification (TS) Section 3.8.1, ``AC Sources--Operating,'' 
Surveillance Requirements from ``[le]10 seconds'' to ``[le]13 
seconds.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed TS change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The accidents previously evaluated in the Updated Final Safety 
Analysis Report (UFSAR) involving emergency diesel generator (EDGs) 
are failure of one EDG to start during a loss-of-offsite power 
(LOOP), with or without, simultaneous occurrence of the design basis 
loss-of-coolant accident (LOCA). New evaluations were necessary for 
the General Electric GE-14 (GE-14) fuel to be used at Quad Cities 
Nuclear Power Station (QCNPS), Units 1 and 2. The proposed change 
increases the time limit allowed for an EDG to start by 3 seconds. 
However, the increased EDG start time span allowance is still within 
the time delay assumed in these newly evaluated accidents. Thus, the 
probability for a successful EDG start is unchanged by this proposed 
change. A change in the start time of an EDG, but still within the 
bounds of the time delay assumed in analyzed accidents, does not 
affect previously evaluated accidents.
    In either accident specified above (i.e., failure of one EDG 
with either a LOOP or a LOOP plus LOCA), the UFSAR accident analysis 
assumes the limiting single failure, as required by 10 CFR 50 
Appendix K, ``ECCS Evaluation Models,'' which is the complete 
failure of the unit EDG to start. The limiting single failure is 
unchanged by the 3-second increase in EDG start time. For this 
limiting single failure, the redundant ``swing'' EDG starts within 
13 seconds and powers the essential loads delivering Emergency Core 
Cooling System (ECCS) flow to the core within the GE-14 LOCA 
analysis assumptions. This GE-14 analysis meets all of the same 10 
CFR 50.46, ``Acceptance criteria for emergency core cooling systems 
for light-water nuclear power reactors,'' requirements as the 
previously evaluated LOCA analysis assuming a 10-second EDG start 
time. Therefore, the consequences of an EDG start failure are not 
impacted by the proposed increase in the allowed EDG start time 
limit. Based on the above, the proposed TS change does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    In summary, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed TS change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change revises the required EDG start time limit 
utilized by the six Surveillance Requirements (SRs) that verify EDG 
start capability. No other changes in requirements are being 
proposed. The revised EDG start time limit is consistent with the 
EDG and ECCS start time delay assumed in the design basis accident 
analysis. Therefore, the proposed EDG start time utilized by the six 
SRs is still bounded by analyzed evaluations of a LOOP or a LOOP in 
conjunction with a LOCA. No new failure modes are introduced by this 
proposed change, In addition, the proposed change does not 
physically alter the plant and will not alter the operation of the 
structures, systems, or components of QCNPS. Therefore, the 
possibility of a new or different kind of accident from any accident 
previously evaluated will not be created.
    The proposed TS change does not involve a significant reduction 
in a margin of safety.
    The consequences of a LOOP or a LOOP in conjunction with a LOCA 
have been previously evaluated. New evaluations were performed for 
the GE-14 fuel to be used at QCNPS. These new evaluations assume a 
bounding and longer EDG start time delay, following detection of the 
LOOP condition, prior to powering permanent loads fed off its 
associated emergency bus. Since the longer EDG start time delay was 
assumed in these new evaluations, any EDG start within the longer 
start time is bounded. The currently specified TS EDG start time 
limit is based on the existing analyses for the fuel utilized by 
QCNPS. New analysis for GE-14 fuel has assumed more conservative and 
bounding time delays for the integrated ECCS delivery timing 
sequence. All of the acceptance criterion of 10 CFR 50.46 continue 
to be met with the new GE-14 conservative EDG and ECCS sequences 
analyzed. The proposed change does not alter the basis upon which 
the start time limit specified in the TS is derived. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

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

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

    Date of amendment request: April 15, 2002.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) to relocate boron 
concentration limits contained in certain TSs to the Core Operating 
Limits Report (COLR). The proposed amendment would change TS 2.1, 
``Safety Limits,'' to relocate Figure 2.1-1, ``Reactor Core Safety 
Limits-Four Loops in Operation,'' to the COLR; and would revise TSs 
2.1.1 and 2.1.2 limiting conditions and actions to be consistent with 
the improved Standard Technical Specifications (ITS). The proposed 
amendment also would relocate the Departure from Nucleate Boiling 
(DNB)-related parameters, specified in TS 3/4.2.5, to the COLR. TS 
6.8.1.6, ``Core Operating Limits Report,'' and the associated TS Bases, 
would be revised to reflect the above changes. Editorial and 
administrative changes, consistent with the ITS, would also be made to 
TS 6.8.1.6.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:

1. The Proposed Changes Do Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.

    The proposed changes to relocate cycle-specific parameters from 
the TSs to the COLR are administrative in nature and do not 
adversely affect accident initiators or precursors nor alter the 
design assumptions, conditions, and configuration of the facility or 
the manner in which it is operated. The proposed changes do not 
alter or prevent the ability of structures, systems, or components 
to perform their intended function to mitigate the consequences of 
an initiating event within the acceptance limits assumed in the 
Updated Final Safety Analysis Report (UFSAR).
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

2. The Proposed Changes Do Not Create the Possibility of a New or 
Different Kind of Accident From Any Previously Evaluated

    The proposed changes do not alter the design assumptions, 
conditions, or configuration of the facility or the manner in which 
it is operated. The proposed changes have no adverse impact on 
component or

[[Page 36932]]

system interactions. Since there are no changes to the design 
assumptions, parameters, conditions and configuration of the 
facility, or the manner in which the plant is operated and 
surveilled, the proposed changes do not create the possibility of a 
new or different accident from any previously analyzed.

3. The Proposed Changes Do Not Involve a Significant Reduction in the 
Margin of Safety.

    There is no adverse impact on equipment design or operation and 
there are no changes being made to the TSs themselves that would 
adversely affect any current margin of safety. The proposed changes 
are administrative in nature and impose alternative procedural and 
programmatic controls on these parameter limits.
    Therefore, relocation of the subject cycle-specific parameter 
limits and other proposed editorial changes, to be reflective of the 
relocated parameters, do not involve a significant reduction in the 
margin of safety.

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

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

    Date of amendment request: April 17, 2002.
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Nuclear Power Plant Technical Specification (TS) 
Section 6.3, ``Plant Staff Qualifications,'' to reflect the title 
change from Superintendent Plant Radiation Protection to Radiation 
Protection Manager. In addition, the licensee informed the United 
States Nuclear Regulatory Commission of its intention to reformat TS 
Section 6.3 to Microsoft WORD format.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

1. Involve a Significant Increase in the Probability or Consequences of 
an Accident Previously Evaluated.

    The proposed changes will not alter the intent of the TS. 
Reformatting TS Section 6.3 is administrative. Changing the title 
from Superintendent Radiation Protection to Radiation Protection 
Manager is also administrative in nature. There is no impact on 
accident initiators or plant equipment, and thus does not affect the 
probability or consequences of an accident.

2. Create the Possibility of a New or Different Kind of Accident From 
Any Accident Previously Evaluated.

    The proposed changes do not involve a change to the physical 
plant or operations. Since these are administrative changes they do 
not contribute to accident initiation. Therefore, they do not 
produce a new accident scenario or produce a new type of equipment 
malfunction.

3. Involve a Significant Reduction in the Margin of Safety

    Since these are administrative changes, they do not involve a 
significant reduction in the margin of safety. The proposed changes 
do not affect plant equipment or operation. Safety limits and 
limiting safety system settings are not affected by this change.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: L. Raghavan, Section Chief

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

    Date of amendment request: April 2, 2002.
    Description of amendment request: The proposed amendment would 
revise Surveillance Requirement (SR) 3.0.3 to extend the delay period, 
before entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period would be extended from the current limit 
of ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is less'' to ``* * * up to 24 hours or up to the 
limit of the specified Frequency, whichever is greater.'' In addition, 
the following requirement would be added to SR 3.0.3: ``A risk 
evaluation shall be performed for any Surveillance delayed greater than 
24 hours and the risk impact shall be managed.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments 
concerning missed surveillances, including a model safety evaluation 
and model no significant hazards consideration (NSHC) determination, 
using the consolidated line-item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on September 28, 2001 (66 FR 49714). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated April 2, 2002.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

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

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

[[Page 36933]]

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

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

    The NRC staff has reviewed the licensee's incorporation of the 
above analysis by reference and, based on this review, it appears that 
the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the 
NRC staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Richard J. Laufer.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: April 9, 2002.
    Description of amendment request: The proposed amendment would 
revise the Ginna Station Improved Technical Specification (ITS) 
associated with Safety Limits, Instrumentation Setpoints, and the Core 
Operating Limits Report. The purpose of this license amendment is to 
provide a clear and consistent identification of instrumentation 
setpoints and their operability basis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Operation of Ginna Station in accordance with the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated. The Reactor Trip 
System (RTS) Instrumentation, Engineered Safety Feature Actuation 
System (ESFAS), Loss of Power (LOP) Diesel Generator (DG) Start 
Instrumentation, and Containment Ventilation Isolation trip 
functions are part of the accident mitigation response and are not 
themselves an initiator for any transient. Therefore, the 
probability of an accident previously evaluated is not significantly 
affected, including the probability of a spurious actuation. This 
proposed amendment includes changes to Allowable Values that have 
been determined with the use of an accepted methodology. The new 
values ensure that all automatic protective actions will be 
initiated at or before the condition assumed in the safety analysis. 
This change will allow the nominal trip setpoints to be adjusted 
within the calibration tolerance band allowed by the setpoint 
methodology. Plant operation with these revised values will not 
cause any design or analysis acceptance criteria to be exceeded. The 
structural and functional integrity of plant systems is unaffected. 
There will be no adverse effect on the ability of the channels to 
perform their safety functions as assumed in the safety analyses. 
Since there will be no adverse effect on the trip setpoints or the 
instrumentation associated with the trip setpoints, there will be no 
significant increase in the consequences of any accident previously 
evaluated.
    Other changes in trip system function, content and format are 
proposed based on the current configuration of the trip system 
hardware. Similarly, since the ability of the instrumentation to 
perform its safety function is not adversely affected, there will be 
no significant increase in the consequences of any accident 
previously evaluated. The proposed editorial, administrative and 
format changes do not affect plant safety and are in accordance with 
NUREG-1431.
    The proposed change to relocate core safety limits and trip 
setpoint parameter values to the Core Operating Limits Report (COLR) 
is a programmatic and administrative change that does not physically 
alter safety-related systems, nor does it affect the way in which 
safety-related systems perform their functions. Because the design 
of the facility and system operating parameters are not being 
changed, the proposed amendment does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated. The cycle-specific values relocated into the 
COLR will continue to be controlled by the Ginna Station programs 
and procedures. Accident analyses addressed in the UFSAR [Updated 
Final Safety Analysis Report] will be examined with respect to 
changes in the cycle-dependent parameters, which are obtained from 
the use of Nuclear Regulatory Commission (NRC) approved reload 
design methodologies, to ensure that the transient evaluation of new 
reloads are bounded by previously accepted analyses. This 
examination, which will be conducted per the requirements of 10 CFR 
50.59, will ensure that future reloads will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. Therefore, the probability or 
consequences of an accident previously evaluated is not 
significantly increased.
    The proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    Operation of Ginna Station in accordance with the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated. The proposed 
amendment includes changes to the format and magnitudes of nominal 
trip setpoints and allowable values that preserve all safety 
analysis assumptions related to accident mitigation. The protection 
system will continue to initiate the protective actions as assumed 
in the safety analysis.
    The proposed change will continue to ensure that the trip 
setpoints are maintained consistent with the setpoint methodology 
and the plant safety analysis. Plant operation will not be changed.
    Other proposed changes are made so that the technical 
specifications more accurately reflect the installed plant specific 
trip system hardware. Furthermore, the proposed changes do not alter 
the functioning of the protection systems. No new mode of failure 
has been created and no new equipment performance requirements are 
imposed. The proposed amendment has no affect on any previously 
evaluated accident.
    The proposed change to relocate core safety limits and trip 
setpoint parameter values to the COLR is a programmatic and 
administrative change and does not result in any change in the 
manner in which the plant is operated or the way in which the 
Reactor Trip System provides plant protection. All of the accident 
transients analyzed in the UFSAR will continue to be protected by 
the same trip functions with the required trip setpoints. Removal of 
the cycle specific variables has no influence or impact on, nor does 
it contribute in any way to the probability or consequences of an 
accident. No safety-related equipment, safety function, or plant 
operation will be altered as a result of this proposed change. The 
cycle specific variables are calculated using the NRC approved 
methods, and submitted to the NRC to allow the staff to continue to 
review the values of these limits. The technical specifications will 
continue to require operation within the core operating limits, and 
appropriate actions will be required if these limits are exceeded. 
Therefore, the possibility for a new or different kind of accident 
from any accident previously evaluated is not created.
    The proposed change does not involve a significant reduction in 
[a] margin of safety.
    Operation of Ginna Station in accordance with the proposed 
changes does not involve

[[Page 36934]]

a significant reduction in a margin of safety. The proposed trip 
setpoint Allowable Values are calculated with an accepted 
methodology. The proposed changes will continue to ensure that the 
trip setpoints are maintained consistent with the setpoint 
methodology and the plant safety analysis. The response of the 
protection system to accident transients reported in the Updated 
Final Safety Analysis Report (UFSAR) is unaffected by this change. 
Therefore, accident analysis acceptance criteria are not affected.
    Other proposed changes are made so that the protection system 
technical specifications more accurately reflect the plant-specific 
trip system hardware. The proposed change does not involve revisions 
to any safety limits or safety system setting that would adversely 
impact plant safety. The proposed change does not alter the 
functional capabilities assumed in a safety analysis for any system, 
structure, or component important to the mitigation and control of 
design bases accident conditions within the facility. Nor does this 
change revise any parameters or operating restrictions that are 
assumptions of a design basis accident. In addition, the proposed 
change does not affect the ability of safety systems to ensure that 
the facility can be placed and maintained in a shutdown condition 
for extended periods of time.
    The proposed change to relocate core safety limits and trip 
setpoint parameter values to the COLR represents an administrative 
change and no hardware changes are involved; therefore, no accident 
analysis acceptance criteria are affected. The margin of safety is 
not affected by the removal of cycle specific core operating limits 
from the technical specifications. The margin of safety presently 
provided by current technical specifications remains unchanged. 
Appropriate measures exist to control the values of these cycle 
specific limits. The proposed amendment continues to require 
operation within the core limits as obtained from NRC approved 
methodologies, and the actions to be taken if a limit is exceeded. 
The development of the limits for future reloads will continue to 
conform to those methods described in NRC approved documentation. In 
addition, each future reload will involve a 10 CFR 50.59 review. The 
proposed amendment is a programmatic and administrative change that 
provides assurance that plant operations continue to be conducted in 
a safe manner. The proposed amendment does not result in any change 
in the manner in which the plant is operated or the way in which the 
Reactor Trip System (RTS) provides plant protection. The proposed 
relocation does not alter the manner in which safety limits, 
limiting safety system setpoints or limiting conditions for 
operation are determined. Therefore, the response of the RTS to 
accident transients described in the UFSAR is unaffected by this 
change. As stated previously, this portion of the proposed amendment 
does not physically alter safety-related systems, nor does it affect 
the way in which safety-related systems perform their functions. The 
accident transients are unaffected and the safety analysis 
acceptance limits are unaffected. The design of the facility and 
system operating parameters are not being changed. Therefore, this 
change does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Daniel F. Stenger, Ballard Spahr Andrews & 
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 
20005.
    NRC Section Chief: Richard J. Laufer.

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].

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: March 1, 2002.
    Brief description of amendments: The amendments revise Surveillance 
Requirement (SR) 3.0.3 to extend the delay period before entering a 
limiting condition for operation following a missed 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: May 7, 2002.
    Effective date: May 7, 2002, and shall be implemented within 60 
days of the date of issuance.
    Amendment Nos.: Unit 1--141, Unit 2--141, Unit 3--141.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 2, 2002 (67 FR 
15621). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 7, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 4, 2001.
    Brief description of amendment: These amendments modify the 
Millstone Nuclear Power Station, Unit No. 2 (MP2) and Unit No. 3 (MP3) 
Technical Specifications (TSs) to relocate selected MP2 and MP3 
technical specifications related to the reactor coolant system to the 
respective Technical Requirements Manual (TRM),

[[Page 36935]]

with the exception of MP3 Technical Specification Section 4.4.10, which 
will be relocated to Section6 of MP3's TS. The Bases of the affected 
TSs will be modified to address the proposed changes.
    Date of issuance: May 8, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 266, 204.
    Facility Operating License Nos. DPR-65 and NPF-49: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55011). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 8, 2002.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423, 
Millstone Nuclear Power Station, Unit No. 3, New London County, 
Connecticut Date of application for amendment: August 27, 2001.

    Brief description of amendment: The amendment changes the Millstone 
Nuclear Power Station, Unit No. 3 Technical Specifications (TSs) action 
and surveillance requirements associated with the containment airlock. 
The Bases of the affected TSs will be modified to address the proposed 
changes.
    Date of issuance: May 15, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 205.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 14, 2001 (66 
FR 57119). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 15, 2002.
    No significant hazards consideration comments received: No.

Entergy Nuclear Generation Company, Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of application for amendment: August 24, 2001, as supplemented 
December 20, 2001, and February 15, 2002.
    Brief description of amendment: The amendment makes changes to the 
license and technical specifications to reflect the transfer of 
operating authority to Entergy Nuclear Operations, Inc.
    Date of issuance: May 5, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 193.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: October 4, 2001 (66 FR 
50694). The supplemental information received after the initial notice 
did not expand the application beyond the scope of the notice or affect 
the applicability of the Commission's generic no significant hazards 
consideration determination pursuant to 10 CFR 2.1315. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated May 5, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 28, 2001.
    Brief description of amendment: The amendment revises the 
Anticipated Transient Without Scram Recirculation Pump Trip Reactor 
Pressure High setpoint by replacing the current conditional setpoints, 
which are based upon the number of Safety Relief Valves out of service, 
with a single setpoint.
    Date of issuance: May 8, 2002.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 273.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64294). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 8, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: January 31, 2002.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 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 in the Entergy System.''
    Date of issuance: May 15, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of implementation of Amendment No. 215.
    Amendment No.: 216.
    Renewed Facility Operating License No. DPR-51: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: February 19, 2002 (67 
FR 7416). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 15, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 25, 2001, as 
supplemented by letter dated February 20, 2002.
    Brief description of amendment: This amendment revises Technical 
Specification 3.1.4, Control Rod Scram Times, to increase the control 
rod scram time testing interval from 120 days to 200 days of full power 
operation.
    Date of issuance: May 14, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 150.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 21, 2001 (66 FR 
15923). The supplemental letter provided clarifying information that 
did not change the original application nor expand the scope of the 
Federal Register notice as published. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
May 14, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 30, 2001.
    Brief description of amendments: The amendments revise Surveillance 
Requirement (SR) 3.0.3 to extend the delay period, before entering a 
Limiting Condition for Operation, following a missed surveillance. The 
delay period is

[[Page 36936]]

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: April 30, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 128 and 123.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 19, 2002 (67 
FR 7417). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 30, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 1, 2001, as supplemented 
on February 8, 2002.
    Brief description of amendments: These amendments revise Limiting 
Condition for Operation 3.6.1.7 concerning drywell average air 
temperature.
    Date of issuance: May 9, 2002.
    Effective date: As of date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 159, 121.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41619). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 9, 2002.
    No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: May 9, 2001, as supplemented by 
electronic mail dated February 28, 2002.
    Brief description of amendment: The amendment consists of changes 
to the Technical Specifications (TSs) Table 3.3.1.1-1, Function 2.b due 
to an error in Amendment 184. The supplementary information in the 
electronic mail dated February 28, 2002, provided clarifications, and 
did not alter the Commission's conclusions regarding significant 
hazards consideration.
    Date of issuance: May 9, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 191.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 59509). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 9, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: February 21, 2002.
    Description of amendment request: The amendment relocates specific 
pressure, differential pressure, and flow values, as well as specific 
test methods, associated with certain Engineered Safeguards Features 
(ESF) pumps from the Technical Specifications to the Seabrook Station 
Technical Requirements Manual.
    Date of issuance: May 2, 2002.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 83.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 19, 2002 (67 FR 
12604). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 2, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 21, 2001, as supplemented March 
25 and April 8, 2002.
    Description of amendment request: The amendment revises Technical 
Specification Surveillance Requirements 4.3.1.2 and 4.3.2.2 to allow 
verification in place of demonstration of response time associated with 
certain pressure sensors, differential pressure sensors, process 
protection racks, nuclear instrumentation, and logic systems.
    Date of issuance: May 2, 2002.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 84.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2925). The supplements dated March 25 and April 8, 2002, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally published, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register on 
January 22, 2002, (67 FR 2925). The Commission's related evaluation of 
the amendment is contained in a Safety Evaluation dated May 2, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 13, 2001, as 
supplemented by letters dated February 26, 2002, March 11, 2002, and 
April 18, 2002.
    Brief description of amendments: The amendments revise the 
technical specifications to incorporate a new alternate repair criteria 
(ARC) for steam generator (SG) tubes with axial primary water stress 
corrosion cracking (PWSCC) at dented tube support plate (TSP) 
intersections. These amendments will apply to future operating cycles 
of Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2.
    Date of issuance: May 1, 2002.
    Effective date: May 1, 2002, to be implemented within 30 days from 
the date of issuance.
    Amendment Nos.: Unit 1--152; Unit 2--152.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 8, 2002 (67 FR 
929). The February 26, 2002, March 11, 2002, and April 18, 2002, 
supplemental letters provided additional clarifying information, did 
not expand the scope of the application as originally noticed, and did 
not change the original proposed no significant hazards consideration 
determination. The Commission's related evaluation of the

[[Page 36937]]

amendments is contained in a Safety Evaluation dated May 1, 2002.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant, 
Unit 2, Hamilton County, Tennessee

    Date of application for amendment: October 9, 2001, as supplemented 
March 13 and April 11, 2002.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 6.8.4.h, to extend the requirement to perform a 
containment integrated leak test from once every 10 years to 11.5 years 
for the current interval only.
    Date of issuance: May 7, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days of issuance.
    Amendment No.: 265.
    Facility Operating License No. DPR-79: Amendment revises the 
technical specifications.
    Date of initial notice in Federal Register: November 14, 2001 (66 
FR 57127). The supplemental letters provided clarifying information 
that was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated May 7, 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: December 6, 2001.
    Brief description of amendment: The amendment revises several of 
the Required Actions in the Callaway Plant Technical Specifications 
(TSs) that require suspension of operations involving positive 
reactivity additions or suspension of operations involving reactor 
coolant system (RCS) boron concentration reductions. In addition, the 
proposed amendment revises several Limiting Condition for Operation 
(LCO) Notes that preclude reductions in RCS boron concentration. This 
amendment revises these Required Actions and LCO Notes to allow small, 
controlled, safe insertions of positive reactivity, but limits the 
introduction of positive reactivity such that compliance with the 
required shutdown margin or refueling boron concentration limits will 
still be satisfied. This amendment is based on an NRC-approved 
traveler, Technical Specification Task Force (TSTF)-286, Revision 2.
    Date of issuance: May 1, 2002.
    Effective date: May 1, 2002, and shall be implemented within 60 
days from the date of issuance.
    Amendment No.: 149.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 5, 2002 (67 FR 
5340). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 1, 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: December 13, 2001, as 
supplemented by letter dated April 8, 2002.
    Brief description of amendment: The amendment revises the Limiting 
Condition for Operation (LCO) 3.5.5, Required Action A.1 for the LCO, 
and Surveillance Requirement 3.5.5.1 in Technical Specification (TS) 
3.5.5, ``Seal Injection Flow.'' The revision replaces the flow and 
differential pressure limits that were stated for the reactor coolant 
pump seal injection flow by limits provided in Figure 3.5.5-1, which 
has been added to the TSs.
    Date of issuance: May 2, 2002.
    Effective date: May 2, 2002, and shall be implemented prior to 
entering Mode 3 ascending during the restart from Refueling Outage 12, 
which is scheduled for the Fall of 2002, subject to the note above 
Surveillance Requirement 3.5.5.1.
    Amendment No.: 150.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 5, 2002 (67 FR 
5341). The supplemental letter of April 8, 2002, does not expand the 
scope of the application as noticed, clarifies the proposed changes 
given in the application, and does not change the staff's proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 2, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 20, 2001.
    Brief description of amendment: The amendment changes Technical 
Specifications Table 3.2.6 by revising the Allowed Outage Times (AOTs) 
and associated action requirements for certain post-accident monitoring 
instrumentation.
    Date of Issuance: May 10, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 207.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 8, 2002 (67 FR 
934). The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated May 10, 2002.
    No significant hazards consideration comments received: No.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of

[[Page 36938]]

telephone comments, the comments have been recorded or transcribed as 
appropriate and the licensee has been informed of the public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland. Publicly available records will be 
accessible from the Agencywide Documents Assess and Management Systems 
(ADAMS) Public Electronic Reading Room on the internet at the NRC Web 
site, http://www.nrc.gov/NRC/ADAMS/index.html. If you do not have 
access to ADAMS or if there are problems in accessing the documents 
located in ADAMS, contact the NRC Public Document room (PDR) Reference 
staff at 1-800-397-4209, 304-415-4737 or by e-mail to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By June 27, 2002, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, located at One White Flint North, 
11555 Rockville Pike (first floor), Rockville, Maryland 20852, and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room). If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's

[[Page 36939]]

Public Document Room, located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland 20852, by the above date. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-001, 
and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant, 
Unit 2, Hamilton County, Tennessee

    Date of amendment request: May 6, 2002, as supplemented on May 8, 
2002 (TSC 02-05).
    Description of amendment request: The amendment changed the 
Technical Specifications (TSs) for Sequoyah Nuclear Plant, Unit 2. The 
proposed change would modify TS Surveillance Requirement 4.4.5.4.a.8 to 
clarify the scope of the steam generator (SG) tube inspections required 
in the SG tubesheet region.
    Date of issuance: May 10, 2002.
    Effective date: May 10, 2002.
    Amendment No.: 266.
    Facility Operating License No. DPR-79: The amendment revises the 
TSs.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): No. The Commission's related evaluation of the 
amendment, finding of emergency circumstances, state consultation, and 
final NSHC determination are contained in a safety evaluation dated May 
10, 2002.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Thomas Koshy, Acting.

    Dated at Rockville, Maryland, this 20th day of May, 2002.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 02-13082 Filed 5-24-02; 8:45 am]
BILLING CODE 7590-01-P