[Federal Register Volume 67, Number 83 (Tuesday, April 30, 2002)]
[Notices]
[Pages 21283-21301]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-10456]


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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 April 5, 2002 through April 18, 2002. The 
last biweekly notice was published on April 16, 2002 (67 FR 18641).

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.

[[Page 21284]]

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

[[Page 21285]]

    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].

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2 (HBRSEP2), Darlington County, South Carolina

    Date of amendment request: February 21, 2002.
    Description of amendment request: The proposed amendment to the 
Technical Specifications for HBRSEP2 will modify the containment vessel 
spray nozzle testing frequency specified in the Surveillance 
Requirement 3.6.6.8 from ``10 years'' to ``following activities which 
could result in nozzle blockage.''
    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:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The proposed change revises the Surveillance frequency from once 
per [``]10 years[``] to [``]following activities [which] could 
result in nozzle blockage.[``] The Containment Spray System is not 
considered as an initiator of any analyzed accident. The proposed 
change does not have a detrimental impact on the integrity of any 
plant structure, system, or component that initiates an analyzed 
accident. The proposed change will not alter the operation of, or 
otherwise increase the failure probability of any plant equipment 
that initiates an analyzed accident. As a result, the probability of 
any accident previously evaluated is not significantly increased.
    The proposed change revises the Surveillance frequency. Reduced 
testing is acceptable where operating experience has shown that 
components routinely pass the Surveillance when performed at the 
specified interval. The system design and construction materials 
provide assurance that the production of significant corrosion 
products is unlikely. Since activities that could introduce foreign 
material are the most likely cause for obstruction, testing or 
inspection following such an activity would verify that the nozzles 
are unobstructed and capable of performing their safety function. 
Such events would necessarily involve a substantive breakdown in 
foreign material controls during such activities. As a result, the 
consequences of any accident previously evaluated are not 
significantly affected.
    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 to the test frequency for the Containment 
Spray System nozzles does not involve the use or installation of new 
equipment. Currently installed equipment is not operated in a new or 
different manner. No new or different system interactions are 
created, and no new processes are introduced.
    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 margin between containment pressure response and containment 
design pressure will not be affected because the design and 
functioning of the Containment Spray System is unchanged. Since the 
system is not susceptible to corrosion induced obstruction, nor is 
the introduction of foreign material from the exterior likely, the 
proposed surveillance frequency is sufficient to provide high 
confidence that the Containment Spray System will be available to 
provide the flow necessary in the event that the safety function is 
required. Therefore, the capacity of the system will remain 
unchanged.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    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, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: March 13, 2002.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) for H. B. Robinson Steam 
Electric Plant (HBRSEP), Unit No. 2, to permit selective implementation 
of an alternative source term (AST). The proposed amendment would 
modify the TS requirements for movement of irradiated fuel and 
performing core alterations.
    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.
    Implementation of the Alternative Source Term does not affect 
the design or operation of HBRSEP, Unit No. 2; rather, once the 
occurrence of an accident has been postulated, the new source term 
is an input to evaluate the consequences of the postulated accident. 
A review of the HBRSEP, Unit No. 2, Updated Final Safety Analysis 
Report (UFSAR) shows that the components and systems affected by the 
proposed changes are not initiators of any previously analyzed 
accident. Therefore, there is no significant increase in the 
probability of any previously analyzed accident.
    The implementation of the Alternative Source Term has been 
evaluated in a revision to the HBRSEP, Unit No. 2, Fuel Handling 
Accident. Based on the results of this analysis, it has been 
demonstrated that, with the requested changes to the Technical 
Specifications, the dose consequences of a postulated Fuel Handling 
Accident are within the regulatory guidance provided by the NRC for 
use with the Alternative Source Term. This guidance is presented in 
10 CFR 50.67 and Regulatory Guide 1.183. Since automatic actuation 
of the control room emergency filtration system, automatic actuation 
of containment ventilation isolation, containment penetration 
operability, and the containment purge filter system are not 
credited in the revised analysis for the Fuel Handling Accident, 
eliminating these requirements during the movement of irradiated 
fuel assemblies will not result in a significant increase in the 
consequences of any previously evaluated accident. In addition, a 
review of the HBRSEP, Unit No. 2, UFSAR shows that the only accident 
resulting in dose consequences that is postulated to occur during 
core alterations is the Fuel Handling Accident. Therefore, the 
Applicability changes associated with core alterations will not 
result in a significant increase in the consequences of any 
previously evaluated accident.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The Proposed Change Does Not Create the Possibility of a New 
or Different Kind of Accident From Any Previously Evaluated.
    The proposed changes are supported by the revised design basis 
Fuel Handling Accident analysis. The proposed changes do not

[[Page 21286]]

introduce any new modes of plant operation and do not involve 
physical modifications to the plant.
    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 changes are associated with the implementation of a 
new licensing basis for HBRSEP, Unit No. 2. The new licensing basis 
implements an Alternative Source Term in accordance with 10 CFR 
50.67 and the associated Regulatory Guide 1.183. The results of the 
revised Fuel Handling Accident analysis, revised in support of this 
submittal, are subject to revised acceptance criteria. This analysis 
has been performed using conservative methodologies in accordance 
with the regulatory guidance. The dose consequences of the limiting 
Fuel Handling Accident are within the acceptance criteria also found 
in the regulatory guidance associated with Alternative Source Terms.
    The proposed changes continue to ensure that doses at the 
exclusion area and low population zone boundaries, as well as the 
control room, are within the corresponding regulatory limits. 
Specifically, the margin of safety for this accident is considered 
to be that provided by meeting the applicable regulatory limits, 
which are conservatively set below the 10 CFR 50.67 limits. With 
respect to control room personnel doses, the margin of safety (the 
difference between the 10 CFR 50.67 limits and the regulatory limits 
defined by 10 CFR 50, Appendix A, Criterion 19 (GDC-19)) continues 
to be satisfied.
    Therefore, because the proposed changes continue to result in 
dose consequences within the applicable regulatory limits, they 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: 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

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

    Date of amendment request: March 26, 2002
    Description of amendment request: The proposed amendments would 
modify the Technical Specifications definitions for Engineered Safety 
Feature Response Time and Reactor Trip System Response Time to provide 
for verification of response time for selected instruments provided 
that the instruments and methodology for verification have been 
previously reviewed and approved by the staff.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. The proposed amendment to the Technical Specifications does 
not result in the alteration of the design, material, or 
construction standards that were applicable prior to the change. The 
same reactor trip system (RTS) and engineered safety features 
actuation system (ESFAS) instrumentation is used, and the response 
time allocation/modeling assumptions in UFSAR [updated final safety 
analysis report] Chapter 15 analysis remain unchanged. Only the 
methodology of time response verification is changed. The proposed 
change will not result in the modification of any system interface 
that would increase the likelihood of an accident since these events 
are independent of the proposed change. The proposed amendment will 
not change, degrade, or prevent actions, or alter any assumptions 
previously made in evaluating the radiological consequences of an 
accident described in the UFSAR. Therefore, the proposed amendment 
does not result in the increase in the probability or consequences 
of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. This change does not alter the performance of the reactor 
protection system (RPS) or ESFAS systems. All RPS and ESFAS channels 
will still have response time verified by test before placing the 
channel in operational service and after any maintenance that could 
affect response time. Changing the method of periodically verifying 
instrument response for certain RPS and ESFAS channels (assuring 
equipment operability) from time response testing to calibration and 
channel checks will not create any new accident initiators or 
scenarios. Periodic surveillance of these instruments will detect 
significant degradation in the channel characteristic. 
Implementation of the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    No. This change does not affect the total system response time 
assumed in the safety analysis. The periodic system response time 
verification method is modified to allow for the use of actual test 
or engineering data. The method of verification still provides 
assurance that the total system response is within that defined in 
the safety analysis, since calibration tests will detect any 
degradation which might significantly affect channel response time. 
Based on the above, it is concluded that the proposed license 
amendment request does not result in a reduction in margin with 
respect to plant safety.

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

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

    Date of amendment request: March 20, 2002
    Description of amendment request: The proposed amendment revises 
the reporting requirements specified in Section 2.E of the Facility 
Operating License and Technical Specification Section 5.6.4 by 
eliminating requirements that provide the U.S. Nuclear Regulatory 
Commission (NRC) with information that is not risk significant, and 
changes the reporting time period to be consistent with Section 50.73 
of Title 10 of the Code of Federal Regulations (10 CFR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    These changes involve administrative requirements only. The 
plant's design basis and the Updated Safety Analysis Report accident 
analysis are not affected. In addition, none of these reporting 
requirements support the plant's emergency plan. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change deletes non-risk significant reporting 
requirements and does not affect plant design or operation.

[[Page 21287]]

Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change only impacts administrative reporting requirements. 
It does not impact the design or operation of any plant system, 
structure, or component. In addition, no Technical Specification 
Safety Limit or instrument allowable value are affected. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.

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

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

    Date of amendment request: March 13, 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 March 13, 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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm

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

    Date of amendment request: March 13, 2002.
    Description of amendment request: The proposed amendment would 
correct several errors that were found subsequent to Nuclear Regulatory 
Commission (NRC) issuance of Amendment 215, which converted the 
Technical Specifications (TSs) for Arkansas Nuclear One, Unit 1 (ANO-
1), to Improved TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The non-administrative proposed changes describe required 
actions to be taken upon loss of required electrical equipment, the 
number of non-licensed operators required on-site in Modes 1, 2, 3, 
and 4, the limitations on radiological effluent releases, and a 
clarification of automatic isolation capabilities when the control 
room is operating in the emergency recirculation

[[Page 21288]]

mode. The proposed changes are not considered accident initiators 
nor do they result in a change to the physical characteristics of 
plant equipment. The proposed changes act to provide reasonable 
response to lost equipment, defense in depth, limitations on 
radiological release, and placing equipment in a fail-safe condition 
and do not adversely affect the accident analysis.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The non-administrative proposed changes describe required 
actions to be taken upon loss of required electrical equipment, the 
number of non-licensed operators required on-site in Modes 1, 2, 3, 
and 4, the limitations on radiological effluent releases, and a 
clarification of automatic isolation capabilities when the control 
room is operating in the emergency recirculation mode. None of the 
proposed changes can initiate any type of accident in and by 
themselves. The proposed changes act only to provide reasonable 
response to lost equipment, defense in depth, limitations on 
radiological release, and placing equipment in a fail-safe 
condition.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The aforementioned non-administrative proposed changes do not 
significantly impact the margin of safety. The proposed required 
action to be performed upon loss of a required inverter in Modes 5 
and 6 provides a possible alternative to suspending fuel movement by 
taking conservative action to identify and declare inoperable, all 
equipment affected by the loss of the required inverter. These 
actions may in turn, require suspension of refueling activities, but 
in any event act to ensure the unit is maintained in a safe shutdown 
condition.
    The increase in the number of required non-licensed operators 
on-site in Modes 1, 2, 3, and 4 is conservative and acts to improve 
the margin to safety by expanding the station's defense in depth.
    Complying with the radiological effluent release limitations as 
set forth in 10 CFR [part] 20 (prior to revision) provides 
acceptable assurance that the health and safety of the public will 
be maintained. ANO-1 current complies with this version of 10 CFR 
[part] 20.
    Finally, the proposed Bases change for the CREVS clarifies that 
automatic isolation signals and devices are not required when the 
control room is already isolated and operating in the emergency 
recirc[ulation] mode of operation. When in this configuration, the 
control room habitability system meets the safety function for which 
it was designed. Thus, this clarification does not affect the margin 
to safety.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

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

Entergy Operations, Inc., 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 amendment request: February 19, 2002.
    Description of amendment request: Entergy Operations, Inc. (EOI) 
requests a license amendment for the Grand Gulf Nuclear Station, Unit 
1. EOI proposes to amend Technical Specification (TS) 3.8.1, ``AC 
[Alternating Current] Sources--Operating'' to remove the reactor 
operational MODE restrictions for testing the High Pressure Core Spray 
(HPCS) Diesel Generator 13 (DG 13). The proposed change would remove 
the restriction associated with surveillance requirements (SRs) that 
prohibit performing the required DG 13 testing while the plant is in 
reactor operation MODES 1, 2, or 3. This TS change would allow the 
performance of all SRs for DG 13 during any MODE of plant operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The HPCS DG and its associated emergency loads are accident 
mitigating features, not accident initiating equipment. Therefore, 
there will be no impact on any accident probabilities by the 
approval of the requested amendment.
    The design of plant equipment is not being modified by these 
proposed changes. As such, the ability of the DG to respond to a 
design basis accident will not be adversely impacted by these 
proposed changes. The capability of the DG to supply power in a 
timely manner will not be compromised by permitting performance of 
DG testing during periods of power operation. Additionally, limiting 
testing to only one DG at a time ensures that design basis 
requirement for backup power is met, should a fault occur on the 
tested DG. Therefore, there would be no significant impact on any 
accident consequences.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new accident causal mechanisms would be created as a result 
of NRC [U.S. Nuclear Regulatory Commission] approval of this 
amendment request since no changes are being made to the plant that 
would introduce any new accident causal mechanisms. Equipment will 
be operated in the same configuration with the exception of the 
plant MODE in which the testing is currently conducted. This 
amendment request does not impact any plant systems that are 
accident initiators; neither does it adversely impact any accident 
mitigating systems.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. The proposed changes to the testing requirements for the 
HPCS DG do not affect the operability requirements for the DG, as 
verification of such operability will continue to be performed as 
required. Continued verification of operability supports the 
capability of the DG to perform its required function of providing 
emergency power to plant equipment that supports or constitutes the 
fission product barriers. Consequently, the performance of these 
fission product barriers will not be impacted by implementation of 
this proposed amendment.
    In addition, the proposed changes involve no changes to 
setpoints or limits established or assumed by the accident analysis. 
On this and the above basis, no safety margins will be impacted.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn,

[[Page 21289]]

1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm

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 amendment request: February 25, 2002.
    Description of amendment request: Entergy Operations, Inc. (EOI) 
requests modification of the Grand Gulf Nuclear Station, Unit 1 
Technical Specifications (TS) to add a new Special Operations Limiting 
Condition for Operation (LCO) (Suppression Pool Makeup--MODE 3). The 
new TS provision would allow installation of the Upper Containment Pool 
(UCP) reactor cavity gates, and draining the reactor cavity pool 
portion of the UCP while still in reactor operation MODE 3, ``Hot 
Shutdown,'' with the reactor pressure less than 230 pounds per square 
inch gauge. EOI also requests modification to the applicability of the 
UCP gates surveillance requirement, TS 3.6.2.4, ``Suppression Pool 
Makeup (SPMU) System,'' to allow installation of the UCP gates in 
reactor operation MODE 1, ``Power Operation,'' MODE 2, ``Startup,'' or 
MODE 3, ``Hot Shutdown.'' The proposed changes would allow earlier 
installation of the UPC gates, and allow draining of the reactor cavity 
pool portion of the UCP while holding the plant in MODE 3 to facilitate 
an earlier start of certain refueling outage work evolutions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise the required water levels in the UCP 
and suppression pool. The probability of an accident previously 
evaluated is unrelated to the water levels in the pools since they 
are mitigative systems. The operation or failure of a mitigative 
system does not contribute to the occurrence of an accident. No 
active or passive failure mechanisms that could lead to an accident 
are affected by these proposed changes.
    The consequences of a previously evaluated accident are not 
significantly increased. The changes have no impact on the ability 
of any of the Emergency Core Cooling Systems (ECCS) to function 
adequately, since adequate net positive suction head (NPSH) is 
provided. The post-accident Containment temperature is not 
significantly affected by the proposed reduction in total heat sink 
volume. The increase in suppression pool water level to compensate 
for the reduction in UCP volume will provide reasonable assurance 
that the minimum post-accident vent coverage is adequate to assure 
the pressure suppression function of the suppression pool is 
accomplished. The suppression pool water level will be raised above 
the current high water limit for the proposed Special Operations LCO 
only after the reactor pressure has been reduced sufficiently to 
assure that the hydrodynamic loads from a loss of coolant accident 
will not exceed the design values. The reduced reactor pressure will 
also ensure that the loads due to main steam safety relief valve 
actuation with an elevated pool level are within the design loads. 
The reduced post-LOCA [loss of coolant accident] Containment 
pressure ensures that post-accident dose consequences with no 
fission product scrubbing by Containment Spray (CS) is bounded by 
the DBA [design basis accident] LOCA.
    Therefore, the proposed changes do not significantly increase 
the consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to the water level requirements for the UCP 
and the suppression pool do not involve the use or installation of 
new equipment. Installed equipment is not operated in a new or 
different manner. No new or different system interactions are 
created, and no new processes are introduced. The increased 
suppression pool water level does not increase the probability of 
flooding in the Drywell. No new failures have been created by the 
change in the water level requirements.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes to the UCP and suppression pool water 
levels do not introduce any new setpoints at which protective or 
mitigative actions are initiated. No current setpoints are altered 
by this change. The design and functioning of the Containment 
pressure suppression system is unchanged. The proposed total water 
volume is sufficient to provide high confidence that the pressure 
suppression and Containment systems will be capable of mitigating 
large and small break accidents. All analyzed transient results 
remain well within the design values for the structures and 
equipment. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm

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

    Date of amendment request: March 14, 2002.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TSs) by extending the allowed outage time 
(AOT), or completion time, associated with an inoperable Emergency Core 
Cooling System accumulator. The proposed changes are based on the 
methodology described in Topical Report WCAP-15049-A, ``Risk-Informed 
Evaluation of an Extension to Accumulator Completion Times,'' Revision 
1. In addition to the AOT extension, other changes would be 
incorporated to make the ``Emergency Core Cooling Systems'' TSs 
consistent with NUREG-1431, ``Standard Technical Specifications--
Westinghouse Plants.'' Format and editorial changes are included as 
necessary to facilitate the revision of the TS text to conform to the 
current TS page 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. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed changes consist of extending allowed outage 
times for required accumulator Technical Specification actions, 
elimination of alarm surveillance requirements associated with the 
accumulators, verifying boron concentration and editorial changes. 
These changes are independent of the probability or consequences of 
accidents previously evaluated in either of the Beaver Valley Power 
Station (BVPS) Updated Final Safety Analysis Reports (UFSARs). Since 
the accumulators are not accident initiators, they do not affect the 
probability of accidents. An NRC [Nuclear Regulatory Commission] 
approved generic analysis for Westinghouse plants, which is 
applicable to BVPS, concludes that extending the accumulator allowed 
outage time for reasons other than boron concentration out of limit 
is acceptable

[[Page 21290]]

because the impact of core damage frequency has been shown to be 
within acceptable limits. The extension to the allowed outage time 
for boron not being within limits is consistent with NUREG-1431 and 
acceptable because the boron is not assumed in the injection phase 
of a loss of coolant accident (LOCA).
    The accumulators, however, do perform an accident mitigation 
function. Their mitigation function is also not affected by the 
proposed changes since none of the associated accident mitigation 
parameters are changed. The accumulator volume available for 
injection remains the same as before the proposed changes, as does 
the boron concentration of the contained water. The accumulator 
valve position requirement to be open with its power removed, and 
the nitrogen cover pressure limit are also not changed by this 
request. As a result the same amount of water, at the same boron 
concentration, will be injected into the Reactor Coolant System 
(RCS) in the same amount of time after the proposed changes are made 
as it was before the proposed changes. Due to the fact that the 
accident mitigation function of the accumulators is not affected by 
the proposed changes, the consequences of an accident previously 
evaluated is also not changed.
    Since the duration of the allowed outage times is not an input 
into the safety analysis (i.e., the safety analysis assumes that all 
of the accumulators are operable), the extension of the allowed 
outage times has no impact on the safety analysis. Therefore, the 
proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. Extending allowed outage times for required Technical 
Specification actions and eliminating alarm surveillance 
requirements associated with the accumulators would not affect the 
operation or maintenance of the accumulators. The accumulators will 
not be operating in any different manner following the proposed 
changes than they were before the proposed changes are made. They 
will not be subjected to any new environmental conditions or 
operational modes, or placed into any new configurations that could 
lead to any new failure mechanisms. The role of the accumulators 
following a LOCA is not altered by adopting the proposed changes. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated for BVPS.
    3. Does the change involve a significant reduction in a margin 
of safety?
    No. The proposed changes do not involve any changes to 
accumulator parameters utilized in the accident analysis. There are 
no changes being made to the accumulator's water volume, boron 
concentration, nitrogen cover pressure or the position of the 
isolation valve. As a result, the assumptions made regarding the 
performance of the accumulators during an accident are unchanged. An 
NRC approved generic analysis for Westinghouse plants concludes that 
extending the accumulator allowed outage time for reasons other than 
boron concentration out of limit is acceptable because the impact on 
core damage frequency has been shown to be within acceptable limits. 
A plant specific risk assessment confirms that this generic analysis 
is applicable to BVPS. The extension to the allowed outage time for 
boron not being within limits is consistent with NUREG-1431 and 
acceptable because the boron is not assumed in the injection phase 
of a LOCA. Therefore, the proposed changes do not involve a 
significant reduction in a margin to 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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Richard J. Laufer.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station Unit No. 1, Oswego County, New York

    Date of amendment request: March 27, 2002.
    Description of amendment request: The licensee proposed to revise 
Section 3.6.3, ``Emergency Power Sources,'' of the Technical 
Specifications to extend the allowed outage time (AOT) for an 
inoperable diesel generator (DG) from the current 7 days to 14 days. In 
addition, Section 3.4.4, ``Emergency Ventilation System,'' and 3.4.5, 
``Control Room Air Treatment System,'' would be revised to delete the 
Limiting Condition for Operation (LCO) that the DGs associated with 
operation of these systems be operable at all times.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff reviewed the licensee's analysis and has 
performed its own, which is presented below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    No. The proposed amendment only affects the AOT for the DGs, and 
LCO for the emergency ventilation and control room air treatment 
systems. There will be no associated changes to the design, 
operational characteristics, function, or reliability of these 
systems.
    These systems were designed to mitigate the consequences of 
various previously evaluated accidents and, as such, are not 
postulated to cause such accidents. Thus, the proposed amendment 
does not affect the safety function of these systems nor the credits 
given to these systems for mitigating accident consequences. 
Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed amendment does not affect accident initiators 
or precursors because it does not alter any design parameter, 
condition, equipment configuration, or manner in which the affected 
systems are operated. Further, it does not alter or prevent the 
ability of structures, systems, or components to perform their 
intended safety or accident mitigating functions. Accordingly, the 
proposed amendment does not create a new or different kind of 
accident from any accident previously evaluated.
    3. Does the amendment involve a significant reduction in a 
margin of safety?
    No. The proposed amendment does not change any design parameter, 
analysis methodology, safety limits or acceptance criteria. The 
revised requirements will continue to ensure reliability and 
operability of the affected systems. Therefore, the proposed 
amendment does not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Joel Munday, Acting.

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

    Date of amendment request: March 28, 2002
    Description of amendment request: The proposed amendment would 
change Technical Specification 3.0.3 to allow a longer period of time 
to perform a missed surveillance. The time would be extended from the 
current limit of ``* * * up to 24 hours or up to the limit

[[Page 21291]]

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 the 
specification: ``A risk evaluation shall be performed for any 
Surveillance delayed greater than 24 hours and the risk impact shall be 
managed.''
    The Nuclear Regulatory Commission (NRC) staff issued a notice of 
opportunity for comment in the Federal Register on June 14, 2001 (66 FR 
32400), on possible amendments concerning missed surveillances, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on September 28, 2001 (66 FR 
49714). The licensee affirmed the applicability of the following NSHC 
determination in its application dated March 28, 2002.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident From Any Previously 
Evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. A 
missed surveillance will not, in and of itself, introduce new 
failure modes or effects and any increased chance that a standby 
system might fail to perform its safety function due to a missed 
surveillance would not, in the absence of other unrelated failures, 
lead to an accident beyond those previously evaluated. The addition 
of a requirement to assess and manage the risk introduced by the 
missed surveillance will further minimize possible concerns. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The extended time allowed to perform a missed surveillance does 
not result in a significant reduction in the margin of safety. As 
supported by the historical data, the likely outcome of any 
surveillance is verification that the LCO [Limiting Condition for 
Operation] is met. Failure to perform a surveillance within the 
prescribed frequency does not cause equipment to become inoperable. 
The only effect of the additional time allowed to perform a missed 
surveillance on the margin of safety is the extension of the time 
until inoperable equipment is discovered to be inoperable by the 
missed surveillance. However, given the rare occurrence of 
inoperable equipment, and the rare occurrence of a missed 
surveillance, a missed surveillance on inoperable equipment would be 
very unlikely. This must be balanced against the real risk of 
manipulating the plant equipment or condition to perform the missed 
surveillance. In addition, parallel trains and alternate equipment 
are typically available to perform the safety function of the 
equipment not tested. Thus, there is confidence that the equipment 
can perform its assumed safety function.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mr. Alvin Gutterman, Morgan Lewis, 1111 
Pennsylvania Avenue NW., Washington, DC 20004.
    NRC Section Chief: L. Raghavan

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

    Date of amendment request: March 29, 2002
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to change TS Section 5.5.12, 
``Primary Containment Leakage Rate Testing Program,'' to reflect a one-
time deferral of the Type A Containment Integrated Leak Rate Test 
(ILRT) to no later than September 2008. This would represent a one-time 
extension of the ILRT interval from 10 years to 15 years.
    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 amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed revision to the Technical Specifications involves a 
one-time extension to the current interval for Type A containment 
testing. The current test interval of ten (10) years would be 
extended on a one-time basis to no longer than fifteen (15) years 
from the last Type A test. The proposed Technical Specification 
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 
reactor containment is designed to provide an essentially leak tight 
barrier against the uncontrolled release of radioactivity to the 
environment for postulated accidents. As such the reactor 
containment itself and the testing requirements invoked to 
periodically demonstrate the integrity of the reactor containment 
exist to ensure the plant's ability to mitigate the consequences of 
an accident.
    Therefore, the proposed Technical Specification change does not 
involve a significant increase in the probability of an accident 
previously evaluated.
    The consequences of the evaluated accidents are the amount of 
radioactivity that is released to secondary containment and 
subsequently to the public.
    The proposed change involves only the extension of the interval 
between Type A containment leakage tests. Type B and C containment 
leakage tests will continue to be performed at the frequency 
currently required by plant Technical Specifications. Industry 
experience has shown, as documented in NUREG-1493, 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. The DAEC [Duane Arnold Energy Center] ILRT test history 
supports this conclusion. NUREG-1493, Performance-Based Containment 
Leak-Test Program, concluded, in part, that reducing the frequency 
of Type A containment leak tests to once per twenty (20) years leads 
to an imperceptible increase in risk. The integrity of the reactor 
containment is subject to two types of failure mechanisms which can 
be categorized as (1) activity based and (2) time based. Activity 
based failure mechanisms are defined as degradation due to system 
and/or component modifications or maintenance. Local leak rate test 
requirements and administrative controls such as design change 
control and procedural requirements for system restoration ensure 
that

[[Page 21292]]

containment integrity is not degraded by plant modifications or 
maintenance activities. The design and construction requirements of 
the reactor containment itself combined with containment inspections 
performed in accordance with ASME Section XI, the Maintenance Rule 
and the DAEC's response to NRC Generic Letter 98-04 (``Potential for 
Degradation of the Emergency Core Cooling System (ECCS) and 
Containment Spray System (CSS) after a Loss-of-Coolant Accident 
(LOCA) because of Construction and Protective Coating Deficiencies 
and Foreign Material in Containment'') serve to provide a high 
degree of assurance that the containment will not degrade in a 
manner that is detectable only by Type A testing, thus maintaining 
containment leakage low. Additionally, the on-line containment 
monitoring capability that is inherent to inerted BWR containments 
allows for the detection of gross containment leakage that may 
develop during power operation.
    Therefore, the proposed Technical Specification change does not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    (2) The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed revision to the Technical Specifications involves a 
one-time extension to the current interval for Type A containment 
testing. Primary containment is designed to contain energy and 
fission products during and after an event. The reactor containment 
and the testing requirements invoked to periodically demonstrate the 
integrity of the reactor containment exist to ensure the plant's 
ability to mitigate the consequences of an accident and do not 
involve the prevention or identification of any precursors of an 
accident. Revision to the Type A test interval does not change the 
events that could lead to containment failure. There are no physical 
changes being made to the plant and there are no changes to the 
operation of the plant that could introduce a new failure mode 
creating an accident.
    Therefore, the proposed Technical Specification change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    (3) The proposed amendment will not involve a significant 
reduction in a margin of safety.
    The proposed Technical Specification 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 proposed change involves only 
the extension of the interval between Type A containment leakage 
tests. The current interval of 10 years, based on past performance, 
would be extended on a one-time basis to 15 years from the last Type 
A test. Type B and C containment leakage tests will continue to be 
performed at the frequency currently required by plant Technical 
Specifications.
    The NUREG-1493 generic study of the effects of extending 
containment leakage testing found that a 20-year interval in 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 about 0.1% to the individual risk and that 
increasing the Type A test interval would have minimal affect on 
this risk since about 95% of the potential leakage paths are 
detected by Type B and Type C testing. The DAEC and industry 
experience strongly supports the conclusion that Type B and C 
testing detects a large percentage of containment leakage paths and 
that the percentage of containment leakage paths that are detected 
only by Type A testing is small. The containment inspections 
performed in accordance with ASME Section XI, the Maintenance Rule 
and the DAEC's response to NRC Generic Letter 98-04 serve to provide 
a high degree of assurance that the containment will not degrade in 
a manner that is detectable only by Type A testing.
    The specific requirements and conditions of the Primary 
Containment Leakage Rate Testing Program, as defined in Technical 
Specifications, exist to ensure that the degree of reactor 
containment structural integrity and leak-tightness that is 
considered in the plant safety analysis is maintained. The overall 
containment leakage rate limit specified by Technical Specifications 
is maintained. Additionally, the on-line containment monitoring 
capability that is inherent to inerted BWR containments allows for 
the detection of gross containment leakage should it develop during 
power operation.
    Therefore, the proposed Technical Specification change will not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Alvin Gutterman, Morgan Lewis, 1111 
Pennsylvania Avenue NW., Washington, DC 20004.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: March 18, 2002.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TS) to remove the administrative 
requirement that a candidate for the plant operations manager position 
hold a Senior Reactor Operator License at the time of appointment. This 
proposed change to the TS endorses Regulatory Guide 1.8, Revision 3, 
``Qualification and Training of Personnel for Nuclear Power Plants,'' 
and would, therefore, allow a broader base of qualified candidates to 
hold this position.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No
    The proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed change provides enhancement to the current 
requirements and clarifies the qualification requirements for the 
operations manager position. This provides additional assurance that 
the personnel filling this position will be properly qualified. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No
    The proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed change provides enhancement to the current requirements 
and clarifies the qualification requirements for the operations 
manager position. There are no structures, systems, or components 
affected by this change. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No
    The proposed change does not involve a significant reduction in 
a margin of safety. The proposed change provides enhancement to the 
current requirements and clarifies the qualification requirements 
for the operations manager position. This provides additional 
assurance that the personnel filling this position will be properly 
qualified. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

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

[[Page 21293]]

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

    Date of amendment request: March 18, 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 LCO, 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 March 18, 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 [a] Margin of Safety.
    The extended time allowed to perform a missed surveillance does 
not result in a significant reduction in [a] margin of safety. As 
supported by the historical data, the likely outcome of any 
surveillance is verification that the LCO 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: Daniel F. Stenger, Ballard Spahr Andrews & 
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 
20005.
    NRC Section Chief: Richard J. Laufer.

Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364, 
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

    Date of amendment request: January 24, 2002.
    Description of amendment request: The proposed amendments would 
delete requirements from the Technical Specifications (TS) (and, as 
applicable, other elements of the licensing bases) to maintain a Post 
Accident Sampling System (PASS). Licensees were generally required to 
implement PASS upgrades as described in NUREG-0737, ``Clarification of 
TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory 
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Assess Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades was an outcome of the 
lessons learned from the accident that occurred at TMI, Unit 2. 
Requirements related to PASS were imposed by Order for many facilities 
and were added to or included in the TS for nuclear power reactors 
currently licensed to operate. Lessons learned and improvements 
implemented over the last 20 years have shown that the information 
obtained from PASS can be readily obtained through other means or is of 
little use in the assessment and mitigation of accident conditions. The 
NRC staff issued a notice of opportunity for comment in the Federal 
Register on August 11, 2000 (65 FR 49271) on possible amendments to 
eliminate PASS, including a model safety evaluation and model no 
significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on October 31, 
2000 (65 FR 65018). The licensee affirmed the applicability of the 
following NSHC determination in its application dated July 2, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase

[[Page 21294]]

in the Probability or Consequences of an Accident Previously Evaluated.
    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result of 
the TMI-2 accident. The specific intent of the PASS was to provide a 
system that has the capability to obtain and analyze samples of plant 
fluids containing potentially high levels of radioactivity, without 
exceeding plant personnel radiation exposure limits. Analytical results 
of these samples would be used largely for verification purposes in 
aiding the plant staff in assessing the extent of core damage and 
subsequent offsite radiological dose projections. The system was not 
intended to and does not serve a function for preventing accidents and 
its elimination would not affect the probability of accidents 
previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual benefit 
to post accident mitigation. Past experience has indicated that there 
exists in-plant instrumentation and methodologies available in lieu of 
a PASS for collecting and assimilating information needed to assess 
core damage following an accident. Furthermore, the implementation of 
Severe Accident Management Guidance (SAMG) emphasizes accident 
management strategies based on in-plant instruments. These strategies 
provide guidance to the plant staff for mitigation and recovery from a 
severe accident. Based on current severe accident management strategies 
and guidelines, it is determined that the PASS provides little benefit 
to the plant staff in coping with an accident.
    The regulatory requirements for the PASS can be eliminated without 
degrading the plant emergency response. The emergency response, in this 
sense, refers to the methodologies used in ascertaining the condition 
of the reactor core, mitigating the consequences of an accident, 
assessing and projecting offsite releases of radioactivity, and 
establishing protective action recommendations to be communicated to 
offsite authorities. The elimination of the PASS will not prevent an 
accident management strategy that meets the initial intent of the post-
TMI-2 accident guidance through the use of the SAMGs, the emergency 
plan (EP), the emergency operating procedures (EOP), and site survey 
monitoring that support modification of emergency plan protective 
action recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any accident 
previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident from any Previously Evaluated.
    The elimination of PASS related requirements will not result in any 
failure mode not previously analyzed. The PASS was intended to allow 
for verification of the extent of reactor core damage and also to 
provide an input to offsite dose projection calculations. The PASS is 
not considered an accident precursor, nor does its existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within the 
containment building.
    Therefore, this change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The elimination of the PASS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a neutral 
impact to the margin of safety. Methodologies that are not reliant on 
PASS are designed to provide rapid assessment of current reactor core 
conditions and the direction of degradation while effectively 
responding to the event in order to mitigate the consequences of the 
accident. The use of a PASS is redundant and does not provide quick 
recognition of core events or rapid response to events in progress. The 
intent of the requirements established as a result of the TMI-2 
accident can be adequately met without reliance on a PASS.
    Therefore, this change does not involve a significant reduction in 
the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: February 14, 2002
    Description of amendment request: The proposed amendment revises 
Technical Specifications (TS) 3.3.2 requirements for Loss of Power 
Instrumentation (Functional Unit 8) and the Technical Specifications 
3.8.1.1, 3.8.1.2, and 3.8.1.3, for AC Sources.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not change the plant design basis, 
system configuration or operation, and do not add or affect any 
accident initiator. Therefore, STPNOC concludes that there is no 
significant increase in the possibility of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No
    The proposed changes do not change the plant design basis, 
system configuration or operation, and do not add or affect any 
accident initiator. Therefore, STPNOC concludes the proposed change 
does not create the possibility of a new or different kind of 
accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No
    No actual plant equipment or accident analyses will be affected 
by the proposed change. Additionally, the proposed changes will not 
relax any criteria used to establish safety limits, will not relax 
any safety system settings, or will not relax the bases for any 
limiting conditions of operation. Therefore, STPNOC concludes the 
proposed changes do not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Morgan Lewis, 1111 Pennsylvania NW., 
Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

[[Page 21295]]

Virginia Electric and Power Company, Docket No. 50-338, North Anna 
Power Station, Unit No. 1, Louisa County, Virginia

    Date of amendment request: December 7, 2001.
    Description of amendment request: This proposed amendment revises 
Technical Specifications Surveillance Requirement 4.6.1.2, 
``Containment Leakage.'' The proposed change will permit a one-time 5-
year extension to the 10-year performance-based Type A test interval.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The proposed extension to Type A testing cannot increase the 
probability of an accident previously evaluated since extension of 
the containment Type A testing is not a physical plant modification 
that could alter the probability of accident occurrence nor, is an 
activity or modification by itself that could lead to equipment 
failure or accident initiation.
    The proposed extension to Type A testing does not result in a 
significant increase in the consequences of an accident as 
documented in NUREG-1493. The NUREG notes that very few potential 
containment leakage paths are not identified by Types B and C tests. 
It concludes that reducing the Type A (ILRT [integrated leak rate 
test]) testing frequency to once per twenty years leads to an 
imperceptible increase in risk.
    North Anna provides a high degree of assurance through testing 
and inspection that the containment will not degrade in a manner 
detectable only by Type A testing. The last two Type A tests 
identified containment leakage within acceptable criteria, 
indicating a very leak-tight containment. Inspections required by 
the ASME [American Society of Mechanical Engineers] Code are also 
performed in order to identify indications of containment 
degradation that could affect leak-tightness. Separately, Types B 
and C testing, required by Technical Specifications, identifies any 
containment opening from design penetrations, such as valves, that 
would otherwise be detected by a Type A test. These factors 
establish that an extension to the North Anna Type A test interval 
will not represent a significant increase in the consequences of an 
accident.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    The proposed revision to North Anna Technical Specifications 
adds a one-time extension to the current interval for Type A 
testing. The current test interval of ten years, based on past 
performance, would be extended on a one-time basis to fifteen years 
from the last Type A test. The proposed extension to Type A testing 
does not create the possibility of a new or different type of 
accident since there are no physical changes being made to the plant 
and there are no changes to the operation of the plant that could 
introduce a new failure.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The proposed revision to North Anna Technical Specifications 
adds a one-time extension to the current interval for Type A 
testing. The current test interval of ten years, based on past 
performance, would be extended on a one-time basis to fifteen years 
from the last Type A test. The proposed extension to Type A testing 
will not significantly reduce the margin of safety. The NUREG-1493 
generic study of the effects of extending containment leakage 
testing found that a 20-year extension in 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 about 0.1 percent of the overall risk and that 
decreasing the Type A testing frequency would have a minimal affect 
on this risk since 95% of the Type A detectable leakage paths would 
already be detected by Type B and C testing. Furthermore, for North 
Anna, maintaining the containment subatmospheric during plant 
operations further reduces the risk of any containment leakage path 
going undetected.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Dominion Nuclear Connecticut, Inc., Millstone Power Station, 
Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, 
Connecticut 06385.
    NRC Section Chief: John A. Nakoski.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: November 29, 2001.
    Description of amendment request: This amendment proposes to revise 
the Technical Specifications containment air partial pressure versus 
service water temperature operating limits and surveillance 
requirements for the recirculation spray pump start delay times.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes to the containment air partial pressure 
versus service water temperature operating curve and recirculation 
spray timer delays will continue to ensure that the containment 
remains operable to mitigate Design Basis Accidents. The revised 
containment operating curve and timer delays do not affect the 
probability of occurrence of any accident previously analyzed. The 
revised containment licensing basis analyses use approved analytical 
methods and continue to demonstrate that the established accident 
analysis acceptance criteria are met. Therefore, there is no 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes to the containment air partial pressure 
versus service water temperature operating curve and recirculation 
spray timer delays will not create any new accident or event 
initiators. The containment will continue to be operated in a 
similar manner. No systems, structures, or components are being 
physically modified such that design function is being altered. The 
proposed change does not alter the nature of events postulated in 
the UFSAR [Updated Final Safety Analysis Report] nor does it 
introduce any unique precursor mechanisms. Therefore, the proposed 
changes do not create the possibility of any accident or malfunction 
of a different type than previously evaluated.
    3. Does the change involve a significant reduction in the margin 
of safety?
    The proposed changes to the containment air partial pressure 
versus service water temperature operating curve and recirculation 
spray time delays and supporting analyses maintain the existing 
safety margins. The revised containment analyses demonstrate that 
current acceptance criteria continue to be satisfied. Therefore, the 
proposed changes do not result in a significant reduction in margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Dominion Nuclear Connecticut, Inc., Millstone Power Station, 
Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, 
Connecticut 06385.
    NRC Section Chief: John A. Nakoski.

[[Page 21296]]

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: February 26, 2002.
    Description of amendment request: This proposed amendment revises 
the surveillance frequency of the quench spray and recirculation spray 
system nozzles from a time period of every 10 years to whenever 
maintenance is conducted that could contribute to nozzle blockage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The proposed change revises the surveillance frequency from 
every 10-years to ``following maintenance that could result in 
nozzle blockage.'' Analyzed events are initiated by the failure of 
plant structures, systems, or components. The containment spray 
system is not considered as an initiator of any analyzed event. The 
proposed change does not have a detrimental impact on the integrity 
of any plant structure, system or component that initiates an 
analyzed event. The proposed change will not alter the operation of, 
or otherwise increase the failure probability of any plant equipment 
that initiates an analyzed accident. As a result, the probability of 
any accident previously evaluated is not significantly increased.
    The proposed change revises the surveillance frequency. Reduced 
testing is justified where operating experience has shown that 
routinely passing a surveillance test performed at a specified 
interval has no apparent connection to overall component 
reliability. In this case, routine surveillance testing at the 
specified frequency is not connected to any activity which may 
initiate reduced component reliability and therefore, has been of 
limited value in ensuring component reliability. Thus, the proposed 
frequency change is not significant for a reliability standpoint. 
The proposed containment spray nozzle surveillance frequency has 
been established based on achieving acceptable levels of equipment 
reliability.
    This change does not affect the plant design. Due to the plant 
design, the spray ring headers are maintained dry. Formation of 
significant corrosion products is unlikely. Due to their location at 
the top of the containment, introduction of foreign material from 
exterior to the headers is unlikely. Since maintenance that could 
introduce foreign material is the most likely cause for obstruction, 
testing or inspection following such maintenance would verify the 
nozzle(s) remain unobstructed and the system's continued capability 
to perform its safety function. As a result, the consequences of any 
accident previously evaluated are not significantly affected by the 
proposed change.
    2. Does the proposed license amendment 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 (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    The margin of safety for this system is based on the capacity of 
the spray headers. The system is not susceptible to corrosion 
induced obstruction or obstruction from external sources to the 
system. Performance of maintenance on the spray ring header would 
now require evaluation of the potential for nozzle blockage and the 
need for a test or inspection. Consequently, the spray header 
nozzles should remain unblocked and available in the event that the 
safety function is required. Therefore, the capacity of the system 
would remain unaffected. Hence, 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Dominion Nuclear Connecticut, Inc., Millstone Power Station, 
Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, 
Connecticut 06385.
    NRC Section Chief: John A. Nakoski.

Virginia Electric and Power Company, Docket No. 50-339, North Anna 
Power Station, Unit No. 2, Louisa County, Virginia

    Date of amendment request: February 11, 2002.
    Description of amendment request: This requested amendment would 
revise Facility Operating License Number NPF-7 to permit Virginia 
Electric and Power Company to irradiate a lead test assembly (LTA) at 
North Anna Power Station, Unit 2 to an end-of-life assembly average 
burnup of about 70 GWD/MTU, with the lead rod average burnup in this 
assembly approaching 73 GWD/MTU. The accompanying requested exemptions 
from 10 CFR 50.44, 10 CFR 50.46, and Appendix K of 10 CFR 50 will be 
processed separately.
    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 probability of occurrence or the consequences of an 
accident previously evaluated is not significantly increased. The 
Framatome lead test assembly is very similar in design to the 
Westinghouse fuel that comprises the remainder of the core. The 
reload core design for the North Anna cycle where this assembly will 
operate to high burnup will meet all applicable design criteria. The 
performance of the Emergency Core Cooling system will not be 
affected by the operation of the lead test assembly, and operation 
of the LTA to high burnup will not result in a change to the North 
Anna reload design and safety analysis limits. Operation of one 
Framatome LTA to high burnup will not result in a measurable impact 
on normal operating plant releases, and will not increase the 
predicted radiological consequences of accidents postulated in 
Chapter 15 of the North Anna UFSAR [Updated Final Safety Analysis 
Report]. Therefore, neither the probability of occurrence nor the 
consequences of any accident previously evaluated is significantly 
increased.
    2. The possibility for a new or different type of accident from 
any accident previously evaluated is not created. The Framatome lead 
test assembly is very similar in design (both mechanical and 
composition of materials) to the resident Westinghouse fuel. All 
design and performance criteria will continue to be met and no new 
single failure mechanisms will be created. The irradiation of this 
fuel assembly to high burnup does not involve any alteration to 
plant equipment or procedures which would introduce any new or 
unique operational modes or accident precursors. Therefore, the 
possibility for a new or different kind of accident from any 
accident previously evaluated is not created.
    3. The margin of safety is not significantly reduced. The 
operation of one Framatome lead test fuel assembly to high burnup 
does not change the performance requirements of any system or 
component such that any design criteria will be exceeded. The normal 
limits on core operation defined in the North Anna Technical 
Specifications will remain applicable for the irradiation of this 
assembly to high burnup. Evaluations will be performed to confirm 
that safety analyses based on the resident Westinghouse fuel remain 
applicable for the core in which the high burnup assembly is 
irradiated. Therefore, the margin of safety as defined in the Bases 
to the North Anna Technical Specifications is not significantly 
reduced.

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

[[Page 21297]]

    Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Dominion Nuclear Connecticut, Inc., Millstone Power Station, 
Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, 
Connecticut 06385.
    NRC Section Chief: John A. Nakoski.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendment: June 21, 2001.
    Brief description of amendment: The amendment replaces individual 
main steamline leakage limits with an aggregate leakage limit, revising 
technical specification surveillance requirement 3.6.1.3.9, which 
provides leakage rate limits applicable to the main steamline isolation 
valves.
    Date of issuance: March 26, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 145.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 3, 2001 (66 FR 
50464). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 26, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 17, 2001.
    Brief description of amendment: The amendment revises the test 
frequency for the containment spray nozzles from ``once per 10 years'' 
to ``following activities that could result in nozzle blockage.''
    Date of issuance: March 28, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 146.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 17, 2001 (66 FR 
52796). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 28, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 5, 2001, as supplemented 
December 28, 2001, and March 1, 2002.
    Brief description of amendment: The amendment relaxes operability 
requirements for primary containment, secondary containment systems, 
and the standby gas treatment system during the movement of irradiated 
fuel and during core alterations.
    Date of issuance: April 3, 2002
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 147
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64286). The supplemental letters contained clarifying information 
and did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice. The Commission's related evaluation of the amendment 
is contained in a Safety Evaluation dated April 3, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 21, 2001
    Brief description of amendment: The amendment revises the actions 
required if the refueling equipment interlocks become inoperable. The 
additional request in the application to revise the frequency of the 
refueling equipment interlock inputs channel functional test from 7 to 
31 days is not included in the issued amendment and will be addressed 
by separate correspondence.
    Date of issuance: April 4, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 148.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 26, 2001 (66 
FR 66463). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 4, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 18, 2001, as supplemented 
September 7 and 28, October 17, 23, 26, and 31, November 8 (2 letters), 
20, 21, 29, and 30, and December 5, 6, 7, 13 (2 letters), 20, 21, and 
26, 2001, and January 8, 15, 16, and 24, and March 15, 22, and 29, 
2002.

[[Page 21298]]

    Brief description of amendment: The amendment would allow an 
increase in the licensed power from 2894 megawatts thermal (MWt) to 
3473 MWt.
    Date of issuance: April 5, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 149.
    Facility Operating License No. NPF-62: The amendment revised the 
Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2002 (67 FR 
5001). The supplemental letters contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 5, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 30, 2001.
    Brief description of amendment: The amendment revises surveillance 
requirement (SR) 3.0.3 to extend the delay period, before entering a 
limiting condition for operation, following a missed surveillance. The 
delay period is extended from the current limit of ``. . . up to 24 
hours or up to the limit of the specified Frequency, whichever is 
less'' to ``. . . up to 24 hours or up to the limit of the specified 
Frequency, whichever is greater.'' In addition, the following 
requirement is added to SR 3.0.3: ``A risk evaluation shall be 
performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    Date of issuance: April 9, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 150.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 19, 2002 (67 
FR 7411). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 9, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 27, 2001, as supplemented 
on January 16 and February 26, 2002.
    Brief description of amendments: The amendments add additional 
references to Technical Specification 5.6.5.b to allow the use of 
ZIRLOTM clad fuel rods in the Calvert Cliffs reactor cores.
    Date of issuance: April 8, 2002.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: 251, 228.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 5, 2001 (66 
FR 46476). The January 16 and February 26, 2002, letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated April 8, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 31, 2001, as supplemented 
by letter dated February 5, 2002.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) to allow an extension of the three-year inspection 
interval of the reactor coolant pump flywheel volumetric examination to 
10 years. In addition, the inspection interval requirement would be 
moved to the administrative controls section of the TSs.
    Date of issuance: April 11, 2002.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 241.
    Facility Operating License No. NPF-6: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44167). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 11, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 30, 2001, as 
supplemented by letters dated February 25 and March 13, 2002.
    Brief description of amendment: The license amendment request 
proposes changes to Arkansas Nuclear One, Unit 2 Technical 
Specification (TS) 3/4.4.9, ``Pressure/Temperature Limits,'' and TS 
3.4.12, ``Low Temperature Overpressure Protection (LTOP) System.'' The 
primary changes are to update the existing pressure/temperature limits 
from 21 to 32 effective full power years and to include additional 
restrictions in the LTOP TSs.
    Date of issuance: April 15, 2002.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 242.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64294). The supplemental letters provided clarifying information 
that did not change the staff's proposed no significant hazards 
consideration determination or expand the application beyond the scope 
of the Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 15, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 2, 2001.
    Brief description of amendment: The amendment revises Technical 
Specification 3.3.2.1 Table 3.3-4, ``Engineered Safety Feature 
Actuation System Instrumentation Trip Values,'' Functional Unit 7.b, 
``Loss of Power, 460 volt Emergency Bus Undervoltage,'' by changing the 
referenced bus from the 460 volt (V) to the 480 V bus, by removing the 
trip setpoint, and by slightly increasing the range of allowable values 
for the degraded voltage setting and its associated time delay.
    Date of issuance: April 16, 2002.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 243.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.

[[Page 21299]]

    Date of initial notice in Federal Register: October 31, 2001 (66 FR 

55015). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 16, 2002.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy 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 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 8, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 192 and 186.
    Facility Operating License Nos. DPR-19 and DPR-25: 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 8, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: 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 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 8, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 153 and 139.
    Facility Operating License Nos. NPF-11 and NPF-18: 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 8, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 22, 2002.
    Brief description of amendments: The amendments would relocate 
technical specifications (TSs) 3/4.9.6, ``Refueling Operations--
Manipulator Crane Operability,'' and TSs 3/4.9.7, ``Refueling 
Operations--Crane Travel--Spent Fuel Storage Pool Building,'' with 
associated Bases to the D. C. Cook updated final safety analysis 
report.
    Date of issuance: April 18, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 267 and 248.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 19, 2002, (67 FR 
12603). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 18, 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: August 2, 2001, as supplemented November 
2, December 4, and December 19, 2001, and January 7, 2002.
    Description of amendment request: The amendment modifies the 
Technical Specifications to allow a one-time extension of the Appendix 
J Type A test (containment integrated leakage rate test) interval from 
10 years to 15 years.
    Date of issuance: April 11, 2002.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 82.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64298). The December 4 and December 19, 2001, and the January 7, 
2002, supplements were clarifying in nature, did not change the scope 
of the original Federal Register notice, and did not affect the staff's 
original proposed finding of no significant hazards considerations. The 
November 2, 2001, supplement was considered in the staff's proposed 
finding of no significant hazards considerations.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 11, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 2, 2001.
    Brief description of amendment: The amendment changes Technical 
Specification Table 3.3.1-1, ``Reactor Protective System 
Instrumentation,'' Item 1, ``Variable High Power Trip [VHPT],'' by 
increasing the maximum allowable value for the VHPT from less than or 
equal to 106.5 percent rated thermal power (RTP) to less than or equal 
to 111 percent RTP.
    Date of issuance: April 10, 2002.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 208.
    Facility Operating License No. DPR-20: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 59510). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 10, 2002.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station, 
Unit No. 2, Salem County, New Jersey

    Date of application for amendment: January 17, 2002, as 
supplemented on March 8 and 22, 2002.
    Brief description of amendment: The amendment revises Salem, Unit 
No. 2, Technical Specifications Section 6.8.4.f, and provides for an 
alternate method for complying with the requirements of Title 10 of the 
Code of Federal Regulations (10 CFR) Section 50.54(o), and 10 CFR Part 
50, Appendix J, Option B. Specifically, the amendment allows a

[[Page 21300]]

one-time interval increase for the Salem, Unit No. 2, Type A, 
Integrated Leakage Rate Test from a maximum of a 10-year interval to a 
maximum 15-year interval.
    Date of issuance: April 11, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 232.
    Facility Operating License No. DPR-75: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 7, 2002 (67 FR 
10450). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 11, 2002.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of application for amendment: April 18, 2001.
    Brief description of amendment: This amendment revises volumetric 
air flow units for Technical Specifications (TS) 4.7.6.c.1, c.3, e.1, 
e.3, f, and g to identify standard air flow units expressed as standard 
cubic feet per minute. Volumetric air flow units for TS 4.6.3.b.1, b.2, 
c.1, and d, and TS 4.9.11.b.1, b.3, d.1, e, and f are being revised to 
identify actual air flow units and are expressed as actual cubic feet 
per minute.
    Date of issuance: April 11, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 159.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 30, 2001 (66 FR 
29361). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 11, 2002.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket No. 50-321, Edwin I. Hatch Nuclear 
Plant, Unit 1, Appling County, Georgia

    Date of application for amendment: January 4, 2002, as supplemented 
by letter dated March 15, 2002.
    Brief description of amendment: The amendment revised the Safety 
Limit Minimum Critical Power Ratio for single loop operation in 
theTechnical Specifications to reflect the results of a cycle-specific 
calculation for Unit 1 Cycle 21.
    Date of issuance: April 5, 2002
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 229.
    Facility Operating License No. DPR-57: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 5, 2002 (67 FR 
5333). The supplement dated March 15, 2002, provided clarifying 
information that did not change the scope of the January 4, 2002, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 5, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 14, 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 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 8, 2002.
    Effective date: As of the date of issuance and shall be implemented 
by August 1, 2002.
    Amendment Nos.: 125 and 103.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications and associated Bases.
    Date of initial notice in Federal Register: March 5, 2002 (67 FR 
10015). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 8, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 22, 2001, as supplemented by 
letters dated January 21; February 5, 14, and 27; and March 4, 2002. 
The supplementary letters provided clarifications of the application 
dated August 22, 2001, and did not alter the NRC staff's conclusions 
regarding no significant hazards consideration determination.
    Brief description of amendments: The proposed amendments revise the 
Technical Specifications to reflect a 1.4 percent increase in the 
reactor core thermal power level from 3,800 megawatts thermal (MWt) to 
3,853 MWt.
    Date of issuance: April 12, 2002.
    Effective date: The amendments are effective as of the date of 
issuance, to be implemented within 60 days from the date of issuance 
for Unit 1 and 60 days from date of installation of 94 steam 
generators for Unit 2.
    Amendment Nos.: Unit 1-138; Unit 2-127.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 26, 2001 (66 
FR 66472). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 12, 2002.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia

    Date of application for amendment: December 11, 2000, as 
supplemented by letters dated May 30, June 18, July 16, July 20, August 
13, August 27, September 27, October 10, October 17, November 8, 
November 19, November 29, December 3, December 7, December 12, and 
December 13, 2001, and January 2, January 25, January 31, February 11, 
February 18, February 22, February 27, March 7, March 18, March 22, and 
March 26, 2002.
    Brief description of amendment: These amendments replace, in their 
entirety, the current technical specifications with a set of improved 
technical specifications based on NUREG-1431, Revision 1, ``Standard 
Technical Specifications, Westinghouse Plants,'' dated April 1995.
    Date of issuance: April 5, 2002.
    Effective date: As of the date of issuance and shall be implemented 
no later than September 2, 2002.

[[Page 21301]]

    Amendment Nos.: 231 and 212.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments change 
the Technical Specifications and the Facility Operating Licenses.
    Date of initial notice in Federal Register: February 26, 2002 (67 
FR 8827). The February 27, March 7, March 18, and March 22, 2002 
supplements contained clarifying information only, and did not change 
or expand the scope of the February 26, 2002, Federal Register notice. 
The March 26, 2002 supplement withdrew a beyond scope issue and reduced 
the scope of the Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 5, 2002.

    Dated at Rockville, Maryland, this 23rd day of April 2002.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 02-10456 Filed 4-29-02; 8:45 am]
BILLING CODE 7590-01-P