[Federal Register Volume 66, Number 247 (Wednesday, December 26, 2001)]
[Notices]
[Pages 66461-66477]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-31473]


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NUCLEAR REGULATORY COMMISSION


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

    (Note: The publication date for this notice will change from 
every other Wednesday to every other Tuesday, effective January 8, 
2002. The notice will contain the same information and will continue 
to be published biweekly.)

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 December 3, 2001 through December 14, 2001. 
The last biweekly notice was published on December 12, 2001 (66 FR 
64284).

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

[[Page 66462]]

of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, 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 January 25, 2002, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714, which is 
available at the NRC's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 
Publicly available records will be accessible electronically 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/NRC/ADAMS/index.html. 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 Branch, or may be delivered to the Commission's Public 
Document Room, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, by the above date. A copy of 
the petition should also be sent to the Office of the General Counsel, 
U.S. Nuclear Regulatory Commission,

[[Page 66463]]

Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Assess and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/NRC/ADAMS/index.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document room (PDR) Reference staff at 1-800-397-4209, 304-
415-4737 or by email to [email protected].

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

    Date of amendment request: April 17, 2001.
    Description of amendment request: The proposed amendment would make 
editorial and administrative corrections to Technical Specifications 
(TS) Section 3.3, ``Instrumentation'', and eliminate minor 
discrepancies between TS Section 3.3 and other plant licensing basis 
documents.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

Does the Change Involve a Significant Increase in the Probability or 
Consequences of an Accident Previously Evaluated?

    The proposed changes involve correction of editorial or 
administrative errors made during the conversion of the Clinton 
Power Station (CPS) Technical Specifications (TS) to the improved TS 
(ITS). These proposed changes are based upon current design and 
licensing basis requirements. The proposed changes involve 
correction or reformatting of the TS and do not involve any physical 
changes to plant systems, including those that mitigate the 
consequences of accidents or the manner in which these plant systems 
are operated. As such, these changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.

Does the Change Create the Possibility of a New or Different Kind of 
Accident From Any Accident Previously Evaluated?

    The proposed changes involve correcting errors or reformatting 
existing TS requirements that do not involve a physical alteration 
of the plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. These changes are consistent with the assumptions in the 
safety analyses and licensing basis. Thus, these changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.

Does the Change Involve a Significant Reduction in a Margin of Safety?

    The proposed changes involve correcting editorial or 
administrative errors introduced during the conversion of the CPS TS 
to the ITS. The change to the Allowable Value for the Control Room 
Ventilation System air intake radiation monitors setpoint in TS 
Table 3.3.7.1-1 is consistent with the supporting analyses for the 
trip setpoint value that was previously contained in the TS. The 
changes involve reformatting or correction of errors, and therefore 
will not reduce any margin of safety because there is no effect on 
any safety analysis assumptions. These proposed changes maintain 
requirements within the safety analyses and licensing basis. 
Therefore, these changes do not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Robert Helfrich, Mid-West Regional Operating 
Group, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, 
IL 60555.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: May 21, 2001.
    Description of amendment request: The proposed amendment would 
revise the actions required if the refueling equipment interlocks 
become inoperable. The proposed changes are consistent with the changes 
submitted to the Nuclear Regulatory Commission by the Technical 
Specifications Task Force, Issue number 225, Revision 1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

Does the Change Involve a Significant Increase in the Probability or 
Consequences of an Accident Previously Evaluated?

    The proposed addition of alternate actions in the event that the 
refueling equipment interlocks are determined to be in operable 
ensures that the safety function provided by the interlocks are 
enforced. This is accomplished through manually inserting a rod 
block to prevent the inadvertent withdrawal of a control rod when 
fuel is being moved over the core region.
    The refueling equipment interlocks are credited in the Control 
Rod Removal Error During Refueling--Fuel Insertion with Control Rod 
Withdrawn as described in Updated Safety Analysis (USAR Section 
15.4.1.1.2.2). The manual insertion of a control rod withdrawal 
block provides equivalent protection for the conditional rod block 
provided by the refueling equipment interlocks.
    The proposed change to the surveillance frequency does not 
change the means in which the refueling equipment operates. A review 
of surveillance history was performed for the past two refueling 
outages. In the last seven performances of the refueling equipment 
interlocks operability test, the interlocks have operated 
successfully with no corrective maintenance or corrective action 
necessary. Therefore, since the proposed changes do not result in 
any physical changes to the facility, or involve any modifications 
to plant systems or design parameters or conditions that contribute 
to the initiation of any accidents previously evaluated, the 
proposed changes do not increase the probability of any accident 
previously evaluated.
    Since the proposed changes maintain the same level of protection 
provided by the refueling equipment interlocks, the conclusion of 
the accident scenario remain valid. The probability of a criticality 
event during refueling remains such that no radioactive material 
would be released. Therefore, the proposed changes do not increase 
the consequences of an accident previously evaluated.
    In summary, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Does the Change Create the Possibility of a New or Different Kind of 
Accident From Any Accident Previously Evaluated?

    The proposed changes do not involve a change to the plant design 
or operation. Inserting a manual rod block is not considered an 
abnormal operation. The change to the SR [surveillance requirement] 
frequency does not increase the probability of a malfunction of the 
refueling equipment interlocks, since the interlocks are considered 
reliable and their function can be verified with each fuel move. As 
a result, the proposed changes do not affect any of the parameters 
or conditions that could contribute to the initiation of any 
accidents. No new accident modes or equipment failure modes are 
created by these changes.

[[Page 66464]]

Therefore, these proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.

Does the Change Involve a Significant Reduction in a Margin of Safety?

    The major challenge to the margin of safety would be a 
criticality event that would cause a potential failure of the fuel 
cladding. The proposed addition of alternative actions in the event 
that the refueling equipment interlocks are determined to be 
inoperable ensure that equivalent protection is in place during fuel 
loading movements. Given this equivalent protection, a criticality 
event is not credible. In addition, the increase in the SR frequency 
for performing the channel functional test of the refueling 
equipment interlocks does not impact the ability of the interlocks 
to perform their function, thereby maintaining the refueling 
interlocks function.
    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: Robert Helfrich, Mid-West Regional Operating 
Group, Exelon Generation Company, LLC, 4300 Windfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Anthony J. Mendiola.

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: December 7, 2001.
    Description of amendment request: The proposed amendments would 
allow implementation of 10 CFR Part 50, Appendix J, Option B, which 
governs performance-based containment leakage testing requirements, for 
Type B and C testing. In addition, the licensee also proposes to (a) 
modify Technical Specification (TS) 3.6.3 to delete the requirement for 
conducting soap bubble tests of welded penetrations during Type A tests 
which are not individually Type B or Type C testable, and (b) to modify 
TS 3.6.3 to delete a separate requirement for leak testing containment 
purge lower and upper compartment and instrument room valves with 
resilient seals. These valves will be covered by the overall 
Containment Leakage Rate Testing Program. Associated changes to the 
Bases are also proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The following discussion is a summary of the evaluation of the 
changes contained in this proposed amendment against the 10 CFR 
50.92(c) requirements to demonstrate that all three standards are 
satisfied. A no significant hazards consideration is indicated if 
operation of the facility in accordance with the proposed 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.

First Standard

    The proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. Implementation of these changes will provide continued 
assurance that specified parameters associated with containment 
integrity will remain within acceptance limits as delineated in 10 
CFR 50, Appendix J, Option B. The changes are consistent with 
current safety analyses. Although some of the proposed changes 
represent minor relaxation to existing TS requirements, they are 
consistent with the requirements specified by Option B of 10 CFR 50, 
Appendix J. The systems affecting containment integrity related to 
this proposed amendment request are not assumed in any safety 
analyses to initiate any accident sequence. Therefore, the 
probability of any accident previously evaluated is not increased by 
this proposed amendment. The proposed changes maintain an equivalent 
level of reliability and availability for all affected systems. In 
addition, maintaining leakage within analyzed limits assumed in 
accident analyses does not adversely affect either onsite or offsite 
dose consequences. Therefore, the proposed amendment does not 
increase the consequences of any accident previously evaluated.

Second Standard

    The proposed amendment will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. No changes are being proposed which will introduce any 
physical changes to the existing plant design. The proposed changes 
are consistent with the current safety analyses. Some of the changes 
may involve revision in the testing of components; however, these 
are in accordance with the McGuire's current safety analyses and 
provide for appropriate testing or surveillance that is consistent 
with 10 CFR 50, Appendix J, Option B. The proposed changes will not 
introduce new failure mechanisms beyond those already considered in 
the current safety analyses. No new modes of operation are 
introduced by the proposed changes. The proposed changes maintain, 
at minimum, the present level of operability of any system that 
affects containment integrity.

Third Standard

    The proposed amendment will not involve a significant reduction 
in a margin of safety. The provisions specified in Option B of 10 
CFR 50, Appendix J allow changes to Type B and Type C test intervals 
based upon the performance of past leak rate tests. 10 CFR 50, 
Appendix J, Option B allows longer intervals between leakage tests 
based on performance trends, but does not relax the leakage 
acceptance criteria. Changing test intervals from those currently 
provided in the TS to those provided in 10 CFR 50, Appendix J, 
Option B does not increase any risks above and beyond those that the 
NRC has deemed acceptable for the performance based option. In 
addition, there are risk reduction benefits associated with 
reduction in component cycling, stress, and wear associated with 
increased test intervals. The proposed changes provide continued 
assurance of leakage integrity of containment without adversely 
affecting the public health and safety and will not significantly 
reduce existing safety margins. Similar proposed changes have been 
previously reviewed and approved by the NRC, and they are applicable 
to McGuire.
    Based upon the preceding discussion, Duke Energy has concluded 
that the proposed amendment 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: Mr. Albert Carr, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina.
    NRC Section Chief: Richard J. Laufer, Acting.

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: November 15, 2001.
    Description of amendment request: Entergy Operations, Inc. is 
proposing that the Grand Gulf Nuclear Station (GGNS) Operating License 
be amended to revise the GGNS Technical Specification Surveillance 
Requirements (SRs) pertaining to testing of the standby emergency 
diesel generators (DGs) to allow DG testing during reactor operation. 
The proposed change would remove the restriction associated with these 
SRs that prohibits conducting the required testing of the DGs during 
reactor operating Modes 1, 2, or 3.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 66465]]

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 any accident previously 
evaluated.
    The DGs and their 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 DGs to respond to a 
design basis accident will not be adversely impacted by these 
proposed changes. The capability of the DGs 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 
requirements for backup power is met, should a fault occur on the 
tested DG. Therefore, there would be no significant impact on any 
accident consequences.
    Based on the above, the proposed change to permit certain DG 
surveillance tests to be performed during plant operation will have 
no effect on accident probabilities or consequences.
    (2) The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    No new accident causal mechanisms would be created as a result 
of NRC [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 conducted. This amendment request does 
not impact any plant systems that are accident initiators; neither 
does it adversely impact any accident mitigating systems.
    Based on the above, implementation of the proposed changes would 
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.
    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 
plant DGs do not affect the operability requirements for the DGs, as 
verification of such operability will continue to be performed as 
required (except during different allowed Modes).
    Continued verification of operability supports the capability of 
the DGs to perform their 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, implementation of the proposed changes would not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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: October 29, 2001.
    Description of amendment request: The proposed amendments would 
revise technical specification (TS) 3.9.3, ``Refueling Operations--
Decay Time,'' by reducing the amount of time that the reactor must be 
subcritical before the licensee is allowed to move irradiated fuel 
assemblies in the reactor pressure vessel from 150 hours to 100 hours. 
The amendment also makes various editorial, format and administrative 
changes.
    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 change does not alter the manner in which fuel 
assemblies are handled or core alterations are performed. The 
proposed change does not alter the manner in which heavy loads are 
controlled at BVPS. The proposed change does not result in changes 
being made to structures, systems, or components (SSCs), or to event 
initiators or precursors. Also, the proposed change does not impact 
the design of plant systems such that previously analyzed SSCs would 
now be more likely to fail. The initiating conditions and 
assumptions for accidents described in the Updated Final Safety 
Analysis Report (UFSAR) remain as previously analyzed. Thus, the 
proposed change does not involve a significant increase in the 
probability of an accident previously evaluated.
    The proposed revision of the decay time from 150 hours to 100 
hours is consistent with the assumptions used in the NRC approved 
fuel handling accident (FHA) analyses for Beaver Valley Power 
Station (BVPS) Unit Nos. 1 and 2. The BVPS radiological analyses 
demonstrates that should a FHA occur within the containment or the 
fuel building that involves irradiated fuel with at least 100 hours 
of decay, the projected offsite doses for this event will be well 
within the applicable regulatory limits.
    Limiting Condition for Operation (LCO) 3.9.3, ``Refueling 
Operations--Decay Time,'' will continue to ensure that irradiated 
fuel is not moved in the reactor pressure vessel until at least 100 
hours after shutdown which is consistent with the FHA radiological 
analysis. This LCO will continue to ensure that key assumptions used 
in the radiological safety analysis are met. The previously analyzed 
SSCs are unaffected by the proposed change and continue to provide 
assurance that they are capable of performing their intended design 
function in mitigating the effects of design basis accidents (DBAs). 
As such, the consequences of accidents previously evaluated in the 
UFSAR will not be increased and no additional radiological source 
terms are generated. Therefore, there will be no reduction in the 
capability of those SSCs in limiting the radiological consequences 
of previously evaluated accidents and reasonable assurance that 
there is no undue risk to the health and safety of the public will 
continue to be provided. Thus, the proposed change does not involve 
a significant increase in the consequences of an accident previously 
evaluated.
    The proposed administrative, editorial, and format changes do 
not affect the probability or consequences of any accident.
    Therefore, the proposed amendment does not significantly 
increase 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?
    No. The proposed amendment does not affect a previously 
evaluated accident; e.g., FHA. The proposed amendment takes credit 
for the normal decay of irradiated fuel and the existing 
radiological analyses for FHAs.
    The proposed change does not involve physical changes to 
analyzed SSCs or changes to the modes of plant operation defined in 
the technical specification. The proposed change does not involve 
the addition or modification of plant equipment (no new or different 
type of equipment will be installed) nor does it alter the design or 
operation of any plant systems. No new accident scenarios, accident 
or transient initiators or precursors, failure mechanisms, or 
limiting single failures are introduced as a result of the proposed 
change.
    The proposed change does not cause the malfunction of safety-
related equipment assumed to be operable in accident analyses. No 
new or different mode of failure has been created and no new or 
different equipment performance requirements are imposed for

[[Page 66466]]

accident mitigation. As such, the proposed change has no effect on 
previously evaluated accidents.
    The proposed administrative, editorial, and format changes do 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    No. The proposed revision of the decay time from 150 hours to 
100 hours is consistent with the assumptions used in the NRC 
approved FHA accident analyses for BVPS Unit Nos. 1 and 2 and thus 
does not involve a significant reduction in a margin of safety.
    The proposed amendment does not alter the manner in which fuel 
assemblies are handled or core alterations are performed. The 
proposed amendment does not alter the manner in which heavy loads 
are controlled at BVPS.
    The proposed changes to the TS requirements will continue to 
ensure that the necessary plant equipment is operable in the plant 
conditions where these systems are required to operate to mitigate a 
DBA. The proposed administrative, editorial, and format changes do 
not affect plant safety.
    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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: L. Raghavan, Acting.

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: October 31, 2001.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TSs) by relocating the pressure 
temperature Limit Curves and Low Temperature Overpressure Protection 
(LTOP) and by creating a Pressure-Temperature Limits Report in 
accordance with Generic Letter 96-03 (GL-96-03), ``Relocation of the 
Pressure Temperature Limit Curves and Low Temperature Overpressure 
Protection System Limits.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed changes are a relocation of the Reactor Coolant 
System (RCS) pressure/temperature (P/T) limits, overpressure 
protection system (OPPS) setpoint, and the enable temperature from 
the Technical Specifications to the proposed Pressure and 
Temperature Limits Report (PTLR). The PTLR is created in accordance 
with the guidance provided by Generic Letter (GL) 96-03 and is 
consistent with the content of NUREG-1431. The RCS P/T limits, OPPS 
setpoint, and enable temperature will continue to meet the 
requirements of 10 CFR 50, Appendix G, and will be generated in 
accordance with the NRC approved methodology described in WCAP-
14040-NP-A, Rev. 2 with the exceptions noted in Technical 
Specification Section 6.9.6.
    Since the proposed changes are administrative in nature and do 
not involve any change to any values being relocated, 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. As stated above, the proposed changes to relocate the RCS P/
T limits, OPPS setpoint, and the enable temperature from the 
Technical Specifications to the PTLR are administrative changes. The 
proposed changes do not result in a physical change to the plant or 
add any new or different operating requirements on plant systems, 
structures, or components.
    Therefore, the proposed changes do not result in a significant 
increase in the possibility of a new or different accident from any 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    No. The margin of safety is not affected by the creation of the 
proposed PTLR. Operation of the plant in accordance with the limits 
specified in the PTLR will continue to meet the requirements of 10 
CFR 50, Appendix G, with the identified exceptions, and will assure 
that a margin of safety is not significantly decreased as the result 
of the proposed changes.
    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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Lakshminaras Raghavan (Acting).

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station (DBNPS), Unit 1, Ottawa County, Ohio

    Date of amendment request: October 9, 2001.
    Description of amendment request: The proposed amendment changes 
affected Technical Specifications (TS) 3/4.3.2.2, ``Instrumentation--
Steam and Feedwater Rupture Control System Instrumentation,'' including 
Table 3.3-11, ``Steam and Feedwater Rupture Control System 
Instrumentation,'' Table 3.3-12, ``Steam and Feedwater Rupture Control 
System Instrumentation Trip Setpoints,'' and Table 4.3-11 ``Steam and 
Feedwater Rupture Control System Instrumentation Surveillance 
Requirements.'' Related administrative changes are proposed to TS 3/
4.3.2.3, ``Instrumentation--Anticipatory Reactor Trip System 
Instrumentation,'' Table 3.3-17, ``Anticipatory Reactor Trip System 
Instrumentation,'' and TS 3/4.3.3.1, ``Instrumentation--Monitoring 
Instrumentation--Radiation Monitoring Instrumentation,'' Table 3.3-6, 
``Radiation Monitoring Instrumentation.'' Related changes to associated 
TS Bases 3/4.3.1 and 3/4.3.2, ``Reactor Protection System and Safety 
System Instrumentation,'' are also proposed.
    The main purpose for this license amendment request is to decrease 
the channel functional test frequency from monthly to quarterly for the 
Steam and Feedwater Rupture Control System (SFRCS) Instrumentation 
Channels.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below. These changes would:

    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because the proposed changes do not 
change any accident initiator, initiating condition, or assumption.
    The proposed changes would revise Technical Specification (TS) 
Table 3.3-11, ``Steam and Feedwater Rupture Control System 
Instrumentation,'' and Table 4.3-11 ``Steam and Feedwater Rupture 
Control System Instrumentation Surveillance Requirements,'' to 
identify the Steam and Feedwater Rupture Control System (SFRCS) 
output logic as a separate Functional Unit. In addition, the 
proposed changes would revise TS Table 3.3-12, ``Steam and Feedwater

[[Page 66467]]

Rupture Control System Instrumentation Trip Setpoints,'' to remove 
the ``Trip Setpoint'' values and also modify the ``Allowable 
Values'' entry for Functional Unit 3, ``Steam Generator Feedwater 
Differential Pressure--High,'' consistent with updated calculations 
and current setpoint methodology, and revise the applicability of TS 
Allowable Values for other SFRCS Functional Units in this table. The 
proposed changes would also revise TS Table 4.3-11 to change the 
Channel Functional Test surveillance requirements for the SFRCS 
instrument channels from monthly to quarterly, consistent with 
current methodology. The proposed changes would also make related 
administrative changes to TS Limiting Condition for Operation (LCO) 
3.3.2.2, TS Table 3.3-17, ``Anticipatory Reactor Trip System 
Instrumentation,'' TS Table 3.3-6, ``Radiation Monitoring 
Instrumentation,'' and the associated TS Bases.
    These proposed changes do not involve a significant change to 
plant design or operation.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
invalidate assumptions used in evaluating the radiological 
consequences of an accident, do not alter the source term or 
containment isolation, and do not provide a new radiation release 
path or alter radiological consequences.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because the proposed 
changes do not introduce a new or different accident initiator or 
introduce a new or different equipment failure mode or mechanism.
    3. Not involve a significant reduction in a margin of safety as 
defined in the basis for any Technical Specification. The SFRCS 
instrumentation setpoint analyses will continue to adequately 
preserve the margin of safety. In addition, there are no new or 
significant changes to the initial conditions contributing to 
accident severity or consequences. Therefore, there are no 
significant reductions in a margin of safety.
    Conclusion:
    On the basis of the above, the Davis-Besse Nuclear Power Station 
has determined that the License Amendment Request does not involve a 
significant hazards consideration. As this License Amendment Request 
concerns a proposed change to the Technical Specifications that must 
be reviewed by the Nuclear Regulatory Commission, this License 
Amendment Request does not constitute an unreviewed safety question.

    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 E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station (DBNPS), Unit 1, Ottawa County, Ohio

    Date of amendment request: October 12, 2001.
    Description of amendment request: The proposed amendment would 
change the Operating License (OL) paragraph 2.C(1), Maximum Power 
Level; OL paragraph 2.C(3)(d), Additional Conditions; Technical 
Specification (TS) 1.3, Definitions--Rated Thermal Power; TS 2.1.1, 
Safety Limits--Reactor Core, and associated Bases; TS 2.2.1, Limiting 
Safety System Settings--Reactor Protection System Setpoints, and 
associated Bases; TS 3/4.1.1.3, Reactivity Control Systems--Moderator 
Temperature Coefficient; TS 3/4.2.5, Power Distribution Limits--DNB 
Parameters; TS 3/4.4.9.1, Reactor Coolant System--Pressure/Temperature 
Limits, and associated Bases; and TS 6.9.1.7, Core Operating Limits 
Report. The purpose of this license amendment application would make 
the necessary revisions to the Davis-Besse Nuclear Power Station 
(DBNPS) TS to reflect an increase in the authorized rated thermal power 
from 2772 MWt to 2817 MWt (approximately 1.63 percent), based on the 
use of Caldon Inc. Leading Edge Flow Meter (LEFM) CheckPlus\TM\ System 
instrumentation to improve the accuracy of the feedwater mass flow 
input to the plant power calorimetric measurement.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated based on the comprehensive analytical 
efforts that were performed to demonstrate the acceptability of the 
proposed power uprate changes. The proposed changes include: 
revision of the maximum power level limit stated in Operating 
License (OL) paragraph 2.C(1) and Technical Specification (TS) 
Section 1.3, increasing the allowable power level from 2772 MWt to 
2817 MWt; revision of the reactor core safety limits specified in TS 
Section 2.1.1; revision of the Reactor Protection System (RPS) high 
flux and Reactor Coolant System (RCS) pressure-temperature setpoints 
provided in TS Section 2.2.1; revision of the RCS pressure-
temperature limits in TS Section 3/4.4.9.1, and a related change to 
OL paragraph 2.C(3)(d); and revision of administrative controls 
associated with the Core Operating Limits Report, as described in TS 
Section 6.9.1.7. In addition, related changes to the TS Bases 
associated with these TS Sections are proposed. An evaluation has 
been performed that identified the systems and components that could 
be affected by these proposed changes. The evaluation determined 
that these systems and components will function as designed and that 
performance requirements remain acceptable.
    The primary loop components (reactor vessel, reactor internals, 
control rod drive mechanisms (CRDMs), loop piping and supports, 
reactor coolant pumps, steam generators and pressurizer) will 
continue to comply with their applicable structural limits and will 
continue to perform their intended design functions. Thus, there is 
no increase in the probability of a structural failure of these 
components leading to an accident.
    The Leak-Before-Break analysis conclusions remain valid and the 
breaks previously exempted from structural consideration remain 
unchanged.
    All of the Nuclear Steam Supply System (NSSS) systems will 
continue to perform their intended design functions during normal 
and accident conditions. The pressurizer spray flow remains above 
its design value. Thus, the control system design analyses, which 
credit the flow, do not require any modification. The components 
continue to comply with applicable structural limits and will 
continue to perform their intended design functions. Thus, there is 
no increase in the probability of a structural failure of these 
components.
    All of the NSSS/Balance of Plant (BOP) interface systems will 
continue to perform their intended design functions. The main steam 
safety valves will provide adequate relief capacity to maintain the 
main steam system within design limits.
    The current loss of coolant accident (LOCA) hydraulic forcing 
functions remain bounding.
    The reduction in power measurement uncertainty through the use 
of the Caldon Leading Edge Flow Meter (LEFM) CheckPlusTM 
system, allows for certain safety analyses to continue to be used, 
without modification, at the 2827 MWt power level (102% of 2772 
MWt). Other safety analyses performed at a nominal power level of 
2772 MWt have been either re-performed or re-evaluated at the 2817 
MWt power level, and continue to meet their applicable acceptance 
criteria. Some existing safety analyses had been previously 
performed at a power level greater than 2827 MWt, and thus continue 
to bound the 2817 MWt power level.
    The proposed changes to the RCS pressure-temperature limit 
curves impose a conservative projection of the increase in neutron 
fluence associated with the power uprate. This projection will 
ensure that the requirements of 10 CFR 50 Appendix G, ``Fracture 
Toughness Requirements,'' will continue to be met following the 
proposed power uprate. The design basis events that were protected 
against by these limits have not changed, therefore, the probability 
of an accident previously evaluated is not increased.
    In addition to the changes related to the proposed power uprate, 
unrelated changes are proposed to revise the moderator temperature 
coefficient requirements listed

[[Page 66468]]

in TS Section 3.1.1.3, and to revise requirements relating to the 
Departure from Nucleate Boiling (DNB) parameters listed in TS 
Section 3.2.5. These proposed changes are conservative changes and 
clarifications that do not involve any physical change to systems or 
components, nor do they alter the typical manner in which the 
systems or components are operated. Therefore, these changes will 
not result in a significant increase in the probability of an 
accident.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed power uprate 
changes do not alter any assumptions previously made in the 
radiological consequence evaluations, nor affect mitigation of the 
radiological consequences of an accident previously evaluated.
    The accident radiation dose evaluation was performed at 2827 MWt 
and is bounding when operating at the proposed 2817 MWt using the 
LEFM CheckPlusTM flow instrumentation.
    The proposed changes unrelated to the power uprate also do not 
alter any assumption previously made in the radiological consequence 
evaluations, nor do they affect mitigation of the radiological 
consequences of an accident previously evaluated. Therefore, these 
changes will not involve a significant increase in the consequences 
of an accident previously evaluated.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new 
accident scenarios, failure mechanisms or single failures are 
introduced as a result of the proposed power uprate changes as well 
as the proposed changes unrelated to the power uprate. All systems, 
structures, and components previously required for the mitigation of 
an event remain capable of fulfilling their intended design 
function. The proposed changes have no adverse effects on any 
safety-related system or component and do not challenge the 
performance or integrity of any safety-related system.
    3. Not involve a significant reduction in a margin of safety 
because extensive analyses of the primary fission product barriers, 
conducted in support of the proposed power uprate, have concluded 
that all relevant design criteria remain satisfied, both from the 
standpoint of the integrity of at the primary fission product 
barrier and from the standpoint of compliance with the regulatory 
acceptance criteria. As appropriate, all evaluations have been 
performed using methods that have either been reviewed and approved 
by the Nuclear Regulatory Commission (NRC) or that are in compliance 
with applicable regulatory review guidance and standards. The 
proposed changes unrelated to the power uprate do not involve a 
significant reduction in a margin of safety because they do not 
involve the potential for a significant increase in a failure rate 
of any system or component, and existing system and component 
redundancy is not affected. Also, these changes do not involve any 
new or significant changes to the initial conditions contributing to 
accident severity or consequences.
    Conclusion:
    On the basis of the above, the Davis-Besse Nuclear Power Station 
has determined that the License Amendment Request 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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.
    Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile 
Point Nuclear Station Unit No. 1, Oswego County, New York 
    Date of amendment request: November 26, 2001.
    Description of amendment request: The licensee proposed to amend 
the Technical Specifications (TSs) to delete Section 3/4.2.6, 
``Inservice Inspection and Testing,'' and its associated bases, revise 
Section 4.2.7, ``Reactor Coolant System Isolation Valves,'' and its 
associated bases, create a new Section 6.17, ``Inservice Testing 
Program,'' and delete several reporting requirements in Section 6.9.3, 
``Special Reports.'' These changes will improve the TSs, making it 
consistent with current NRC guidance and the improved Standard 
Technical Specifications for General Electric (GE) Boiling Water 
Reactor (BWR)/4 and BWR/6 plants (NUREG-1433 and NUREG-1434, 
respectively). Most of these changes would also render the TSs to be 
similar to the Nine Mile Point Nuclear Station, Unit No. 2 TSs, which 
is based on NUREG-1433 and NUREG-1434.
    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 operation of Nine Mile Point Unit 1 in accordance with 
the proposed amendment will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed amendment deletes duplicative and unnecessary 
inservice inspection (ISI) and inservice testing (IST) requirements 
from the Technical Specifications; clarifies remaining IST 
requirements; revises a requirement to perform quarterly testing of 
the reactor coolant isolation valves to conform to the periodic 
testing requirements of the ASME [American Society of Mechanical 
Engineers] Boiler and Pressure Vessel Code (ASME Code); and deletes 
unnecessary reporting requirements relating to routine ISI, primary 
containment leakage testing, and secondary containment leakage 
testing. These changes do not reduce the plant's existing ISI/IST 
commitments based on 10CFR50.55a, Section XI of the ASME Code, and 
Generic Letter 88-01. These changes also do not involve hardware 
changes, changes in plant setpoints, or changes in plant safety 
parameters.
    Based on the above, the operation of Nine Mile Point Unit 1 
(NMP1) in accordance with the proposed amendment, will not involve a 
significant increase in the probability or the consequences of an 
accident previously evaluated.
    2. The operation of Nine Mile Point Unit 1 in accordance with 
the proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes do not involve any physical modifications 
to the plant nor alter equipment configuration, setpoints, or safety 
parameters. The ISI/IST related changes are consistent with current 
NRC guidance and industry standards and will continue to ensure 
acceptable equipment operability and availability.
    Based on the above, the operation of NMP1 in accordance with the 
proposed amendment cannot create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The operation of Nine Mile Point Unit 1 in accordance with 
the proposed amendment will not involve a significant reduction in a 
margin of safety.
    The proposed changes do not affect any of the plant's fission 
product barriers or safety/operational limits. The ISI/IST related 
changes will continue to ensure acceptable equipment operability and 
availability.
    Based on the above, the operation of NMP1 in accordance with the 
proposed amendment 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 
requested amendment 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: L. Raghavan, Acting.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station Unit No. 2, Oswego County, New York

    Date of amendment request: November 20, 2001.
    Description of amendment request: The licensee proposed to amend 
the Technical Specifications (TSs) regarding the safety limit minimum 
critical power

[[Page 66469]]

ratio (SLMCPR) to reflect the results of cycle-specific calculations 
performed for the next fuel cycle (i.e., Cycle 9), using NRC-approved 
methodology for determining SLMCPR values. The proposed amendment would 
also editorially revise references to topical reports which document 
the approved methodology.
    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 operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The derivation of the revised Safety Limit Minimum Critical 
Power Ratio (SLMCPR) values for Nine Mile Point Unit 2 (NMP2) Cycle 
9 for incorporation into the Technical Specifications (TS) and their 
use to determine cycle-specific thermal limits has been performed 
using the NRC-approved methods and procedures in [Topical Report] 
NEDE-24011-P-A, ``General Electric Standard Application for Reactor 
Fuel'' (GESTAR II). The analysis methodology incorporates cycle-
specific parameters and reduced power distribution uncertainties in 
the determination of the SLMCPR values. These calculations do not 
change the method of operating the plant and have no effect on the 
probability of an accident initiating event or transient.
    The basis of the Minimum Critical Power Ratio Safety Limit is to 
ensure no mechanistic fuel damage is calculated to occur if the 
limit is not violated. The new SLMCPR values preserve the existing 
margin to transition boiling and the probability of fuel damage is 
not increased. The deletion of listed documents that are already 
incorporated by reference into GESTAR II is administrative only. 
Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The new SLMCPR values for the NMP2 Cycle 9 core reload have been 
calculated in accordance with the methods and procedures described 
in GESTAR II. These methods have been reviewed and approved by the 
NRC. The deletion of listed documents that are already incorporated 
by reference into GESTAR II is administrative only. The changes do 
not involve any new method for operating the facility and do not 
involve any facility modifications. No new initiating events or 
transients result from these changes. Therefore, the proposed TS 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not involve a significant reduction in 
a margin of safety.
    The margin of safety as defined in the TS bases will remain the 
same. The new, cycle-specific SLMCPR values are calculated using 
NRC-approved methods and procedures that are in accordance with the 
current fuel design and licensing criteria. The SLMCPR values remain 
high enough to ensure that greater than 99.9% of all fuel rods in 
the core are expected to avoid transition boiling if the limits are 
not violated, thereby preserving the fuel cladding integrity. The 
deletion of listed documents that are already incorporated by 
reference into GESTAR II is administrative only. Therefore, the 
proposed TS changes do not involve a significant reduction in [a] 
margin of safety.

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

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

    Date of amendment request: November 19, 2001.
    Description of amendment request: A change is proposed to Technical 
Specification (TS) 3.0.3 to allow a longer period of time to perform a 
missed surveillance. The time is extended from the current limit of ``* 
* * up to 24 hours or up to the limit of the specified Frequency, 
whichever is less'' to ``* * * up to 24 hours or up to the limit of the 
specified Frequency, whichever is greater.'' In addition, the following 
requirement would be added to 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 November 7, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

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

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

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

    The extended time allowed to perform a missed surveillance does 
not result in a significant reduction in the margin of safety. As 
supported by the historical data, the likely

[[Page 66470]]

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: John H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: William D. Reckley, Acting.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of amendment request: February 2, 2001, supplemented August 
31, 2001.
    Description of amendment request: The proposed amendments would 
revise the technical specifications (TSs) to clarify the plant 
conditions under which various specifications are applicable. The 
licensee stated in its amendment request that a literal reading of the 
current technical specifications wording may result in situations where 
a routine plant shutdown would seem to be prohibited by TSs and, 
thereby, require entry into TS 3.0.C. This amendment request also makes 
several administrative changes to the TSs, including revising 
references to the Chief Nuclear Corporate Officer, capitalizing defined 
terms, and updating references to previously relocated TS paragraphs 
and correcting the List of Figures. The licensee's supplement to the 
amendment request, dated August 31, 2001, proposed a correction of a 
typographical error in TS Table 3.5-2B, Action 33.
    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 operation of the facility with the proposed amendment 
involve a significant increase in the probability or consequences of 
an accident previously evaluated?
    The proposed changes are administrative in nature and clarify 
existing specifications without reducing or altering the 
requirements imposed by existing specifications. The proposed 
changes do not significantly affect any system that is a contributor 
to initiating events for previously evaluated accidents. Neither do 
the changes significantly affect any system that is used to mitigate 
any previously evaluated accidents. Therefore, the proposed changes 
do not involve any significant increase in the probability or 
consequence of an accident previously evaluated.
    2. Does operation of the facility with the proposed amendment 
create the possibility of a new or different kind of accident from 
any accident previously evaluated?
    The proposed changes are administrative in nature and clarify 
existing specifications without reducing or altering the 
requirements imposed by existing specifications. The proposed 
changes do not alter the design, function, or operation of any plant 
component and do not install any new or different equipment, 
therefore a possibility of a new or different kind of accident from 
those previously analyzed has not be[en] created.
    3. Does operation of the facility with the proposed amendment 
involve a significant reduction in a margin of safety?

    The proposed changes are administrative in nature and clarify 
existing specifications without reducing or altering the requirements 
imposed by existing specifications. Thus, the proposed change[s] do not 
involve a significant reduction in the margin of safety associated with 
the safety limits inherent in either the principle barriers to a 
radiation release (fuel cladding, RCS [reactor coolant system] 
boundary, and reactor containment), or the maintenance of critical 
safety functions (subcriticality, core cooling, ultimate heat sink, RCS 
inventory, RCS boundary integrity, and containment integrity).

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Acting Section Chief: William D. Reckley, Acting.

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

    Date of amendment request: November 21, 2001.
    Description of amendment request: The proposed amendment will 
revise Technical Specifications 2.15(5) and 2.15(6) to identify: (1) 
all indication and control functions required for the alternate 
(remote) shutdown panels, (2) panel locations of the functions, and (3) 
the number of operable channels required.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes to Technical Specifications Sections 
2.15(5) and 2.15(6) identify functions, instruments, and controls 
along with their location and the number of required channels. New 
Technical Specifications Section 2.15(5) addresses the regulatory 
requirements for equipment required for Alternative and Dedicated 
Shutdown Capability per 10 CFR part 50, Appendix R. It will ensure 
that proper Limiting Conditions for Operation are entered for 
equipment or functional inoperability. There are no physical 
alterations being made to the Alternate Shutdown Panels and the 
Auxiliary Feedwater Panel or related systems. Therefore, the 
proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes will not result in any physical alterations 
to the Alternate Shutdown Panels or the Auxiliary Feedwater Panel, 
or any plant configuration, systems, equipment, or operational 
characteristics. There will be no changes in operating modes, or 
safety limits, or instrument limits. With the proposed changes in 
place, Technical Specifications retain requirements for the 
Alternate Shutdown Panels and the Auxiliary Feedwater Panel. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes clarify the regulatory requirements for the 
Alternative and Dedicated Shutdown Capability as defined by 10 CFR 
Part 50, Appendix R. The proposed changes will not alter any 
physical

[[Page 66471]]

or operational characteristics of the Alternate Shutdown Panels and 
the Auxiliary Feedwater Panel and their associated systems and 
equipment. Therefore, the proposed changes do not involve a 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: November 21, 2001.
    Description of amendment request: The proposed amendment will add 
three topical report references to Technical Specification 5.9.5, 
``Core Operating Limit Reports.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment incorporates three additional Framatome 
ANP topical reports for conducting core reload analyses. Since the 
intent of the amendment request is to add references to NRC-approved 
reload analysis methods, the proposed 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 accident previously 
evaluated.
    No new or different modes of operation are proposed as a result 
of these changes. The proposed revision does not change any 
equipment required to mitigate the consequences of an accident. The 
proposed addition of NRC-approved topical reports to the Technical 
Specification does not modify the manner in which the topical 
reports may be implemented. The plant will continue to operate 
within the limits specified by the Core Operating Limits Report and 
corrective actions will be taken in accordance with the Technical 
Specifications should these limits be exceeded. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    As required by Technical Specification 5.9.5, the analytical 
methods used to determine the core operating limits shall be those 
previously reviewed and approved by the NRC. The proposed change 
incorporates methodologies applicable for use with fuel supplied by 
Framatome ANP that have been approved by the NRC as documented by 
Safety Evaluation Reports (References 10.1, 10.2, and 10.3 [of the 
November 21, 2001, amendment request]). Because Technical 
Specification 5.9.5 also requires that the core operating limits 
shall be determined and requires that all applicable limits of the 
safety analysis are met, 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: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power 
Plant, Unit 3, Humboldt County, California

    Date of amendment request: December 28, 2000, as supplemented by 
letters dated March 29 and October 31, 2001.
    Description of amendment request: The proposed amendment would 
convert the Humboldt Bay Power Plant Unit 3 Current Technical 
Specifications to a set of Permanently Defueled Technical 
Specifications with a more standardized format and content based on a 
revision to 10 CFR 50.36 (Technical Specifications) and technical 
specifications approved for other permanently shutdown nuclear power 
plants (Millstone Unit 1 and Trojan Nuclear Plant).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analyses of the issue of no significant hazards 
consideration, which are presented below.

    The conversion of the Humboldt Bay Power Plant (HBPP) Current 
Technical Specifications (CTS) to Permanently Defueled Technical 
Specifications (PDTS) involves the following four types of 
dispositions:

A  Administrative reformatting and rewording
D  Item deleted from the Technical Specifications (TS)
LG  Relocating items from CTS to the Defueled Safety Analysis Report 
(DSAR), PDTS, or other Licensee-Controlled Document
N  Addition of new requirements of new sections to the PDTS

Administrative Reformatting and Rewording

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change involves reformatting and editorially 
rewording of the CTS. As such, this change is administrative in 
nature and does not impact initiators of analyzed events or assumed 
mitigation of accidents or transient events. 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 accident previously 
evaluated.
    The proposed change does not necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or changes in parameters governing normal plant 
operation. The proposed change will not impose any different 
operational requirements and any administrative additions are non-
operational in nature and have not been identified and justified. 
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 a margin of safety.
    The proposed change will not reduce a margin of safety because 
it has no impact on any safety analysis assumptions. This change is 
administrative in nature. As such, no question of safety is 
involved.

Items Deleted from the Technical Specifications that are Duplicative in 
Nature to Other Regulatory Requirements

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change involves deleting information from the CTS. 
The information being deleted is still required to be performed and 
is being performed by the licensee because the information is 
contained in regulatory requirements contained in the Code of 
Federal Regulations. 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 accident previously 
evaluated.
    The proposed change does not necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or changes in parameters governing normal plant 
operation. The proposed change will not impose any different 
operational requirements. 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 a margin of safety.
    The proposed change will not reduce a margin of safety because 
it has no impact on

[[Page 66472]]

any safety analysis assumptions. This change is administrative in 
nature. The requirements being deleted from the CTS are still 
required to be met and are being met by the licensee because these 
requirements exist in the Code of Federal Regulations. As such, no 
question of safety is involved.

Items Deleted from the Technical Specifications That Have No 
Application in the Proposed HBPP PDTS

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change involves deleting information from the CTS. 
The deletion process involves no technical changes to the CTS. As 
such, this change is administrative in nature and does not impact 
initiators of analyzed events or assumed mitigation of accidents or 
transient events. 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 accident previously 
evaluated.
    The proposed change does not necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or changes in parameters governing normal plant 
operation. The proposed change will not impose any different 
operational requirements. 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 a margin of safety.
    The proposed change will not reduce a margin of safety because 
it has no impact on any safety analysis assumptions. This change is 
administrative in nature. As such, no question of safety is 
involved.

Relocating Information from CTS to the DSAR, PDTS Bases or Other 
Licensee-Controlled Documents

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change relocates requirements and descriptive 
information from the CTS to the PDTS Bases, DSAR, or other licensee-
controlled documents. The PDTS Bases, DSAR, or other licensee-
controlled documents containing the relocated requirements and 
information will be maintained using provisions of 10 CFR 50.59 or 
other appropriate regulatory controls. Since any future changes to 
the PDTS Bases, DSAR, or other licensee-controlled documents will be 
evaluated per the requirements of 10 CFR 50.59 or other appropriate 
regulatory controls, proper controls are in place to adequately 
limit the probability or consequences of an accident previously 
evaluated. 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 accident previously 
evaluated.
    The proposed change does not necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or changes in parameters governing normal plant 
operation. The proposed change will not impose any different 
requirements and adequate control of the information will be 
maintained. 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 a margin of safety.
    The proposed change will not reduce a margin of safety because 
it has no impact on any safety analysis assumptions. In addition, 
the requirements and information to be relocated from the CTS to the 
PDTS Bases, DSAR, or other licensee-controlled documents are not 
being revised; they are being relocated verbatim. Since any future 
changes to these requirements in the PDTS Bases, DSAR, or other 
licensee-controlled documents will be evaluated per the requirements 
of 10 CFR 50.59 or other appropriate regulatory controls, proper 
controls are in place to maintain an appropriate margin of safety. 
Therefore this change does not involve a significant reduction in a 
margin of safety.

Addition of New Requirements or New Sections to the PDTS

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change involves the addition of requirements or 
sections to the proposed PDTS. Each addition either provides 
equivalent or potentially more restrictive controls than previously 
provided. The additional requirements or controls do not impact 
initiators of analyzed events or assumed mitigation of accidents or 
transient events. 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 accident previously 
evaluated.
    The proposed change does not necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or changes in parameters governing normal plant 
operation. The proposed change will not impose any different 
operational requirements and any addition is non-operational in 
nature and has been identified and justified. 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 a margin of safety.
    The proposed change will not reduce a margin of safety because 
it has no impact on any safety analysis assumptions. This change 
provides the equivalent or more restrictive requirements on the 
surveillance and control of TS parameters. As such, no question of 
safety is involved.

    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: Christopher J. Warner, Esquire, Pacific Gas 
and Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

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.
    Description of amendment request: Approve reactor core power 
uprate, and revise the Technical Specifications to reflect the power 
uprate.
    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:

    STPNOC [South Texas Project Nuclear Operating Company] has 
evaluated whether or not a significant hazards consideration is 
involved with the proposed amendment by focusing on the three 
standards set forth in 10 CFR 50.92, ``Issuance of amendment,'' as 
discussed below.
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The comprehensive analytical efforts performed to support the 
proposed uprate conditions include a review and evaluation of all 
components and systems (including interface systems and control 
systems) that could be affected by this change. The revised power 
uprate value was input to applicable safety analyses. The proposed 
change is not an initiator of any design-basis accident. All of the 
Nuclear Steam Supply System or Balance of Plant interface systems 
will continue to perform their intended design functions and meet 
all performance requirements. The primary loop components (reactor 
vessel, reactor internals, control rod drive mechanisms, loop piping 
and supports, reactor coolant pump, steam generator, and 
pressurizer) continue to comply with their applicable structural 
limits and will continue to perform their intended design functions. 
Therefore, there is no increase in the probability of a structural 
failure of these components.
    The auxiliary systems and components continue to comply with 
applicable structural limits and will continue to perform their 
intended design functions. Therefore, there is no increase in the 
probability of a structural failure of these components. The steam 
generator safety valves will provide adequate relief capacity to 
maintain the steam generators within design limits. The steam dump 
system will still relieve 40

[[Page 66473]]

percent of the maximum full-load steam flow.
    Therefore, the proposed change does not involve a significant 
increase in the probability of an accident previously evaluated.
    The applicable analyses have been evaluated with respect to the 
increase in core power associated with this change. All applicable 
radiological acceptance criteria continue to be met. Therefore, the 
proposed change does not involve a significant increase in the 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change neither causes the initiation of any 
accident nor creates any new limiting single failures. All of the 
affected systems and components continue to perform their intended 
design functions. The proposed change has no adverse effects on any 
safety-related system or component and does not challenge the 
performance or integrity of any safety-related system.
    The proposed 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 a margin of safety.
    The WRB-2M DNB methodology is used to demonstrate that core 
thermal-hydraulic limits are maintained without any significant 
reduction in margin of safety for the uprated power level of 3853 
MWt (1.4-percent uprate) assuming core designs composed of Robust 
Fuel Assemblies. The WRB-1 DNB correlation demonstrates that there 
is no significant reduction in the margin of safety for core designs 
composed of standard or Vantage 5 Hybrid (V5H) fuel types. Extensive 
analyses of the primary fission product barriers have concluded that 
all relevant design criteria remain satisfied, both from the 
standpoint of the integrity of the primary fission product barrier 
and from the standpoint of compliance with the regulatory acceptance 
criteria. As appropriate, all evaluations have been performed using 
methods that either have been reviewed and approved by the Nuclear 
Regulatory Commission or are in compliance with all applicable 
regulatory review guidance and standards.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, STPNOC concludes that the proposed amendment 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and accordingly, a finding of ``no 
significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: October 23, 2001.
    Brief description of amendments: The proposed amendment would 
revise Technical Specification (TS) 5.5.9, ``Steam Generator Tube 
Surveillance Program,'' to permit tube sleeving repair techniques, 
developed by Westinghouse Electric Company (Westinghouse) and referred 
to as ``Westinghouse Leak Tight Sleeves.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The Westinghouse Leak Tight Sleeves are designed using the 
applicable American Society of Mechanical Engineers (ASME) Boiler 
and Pressure Vessel Code and, therefore, meet the design objectives 
of the original steam generator tubing. The applicable design 
criteria for the sleeves conforms to the stress limits and margins 
of safety of Section III of the ASME code. Mechanical testing has 
shown that the structural strength of repair sleeves under normal, 
upset, and faulted conditions provides margin to the acceptance 
limits. These acceptance limits bound the most limiting (three times 
normal operating pressure differential) burst margin recommended by 
Draft Regulatory Guide 1.121. Burst testing of sleeved tubes has 
demonstrated that no unacceptable levels of primary-to-secondary 
leakage are expected during any plant condition.
    Evaluation of the repaired steam generator tubes indicates no 
detrimental effects on the sleeve or sleeve-tube assembly from 
reactor coolant system flow, primary or secondary coolant 
chemistries, thermal conditions or transients, or pressure 
conditions as may be experienced at CPSES [Comanche Peak Steam 
Electric Station]. Corrosion testing of sleeve-tube assemblies 
indicates no evidence of sleeve or tube corrosion considered 
detrimental under anticipated service conditions.
    The installation of the proposed sleeves is controlled via the 
sleeving vendor's proprietary processes and equipment. The 
Westinghouse process has been in use since 1984 and has been 
implemented more than 24 times for the installation of over 4,200 
sleeves. The CPSES steam generator design was reviewed and found to 
be compatible with the installation processes and equipment.
    The implementation of the proposed amendment has no significant 
effect on either the configuration of the plant or the manner in 
which it is operated. The consequences of a hypothetical failure of 
the sleeved tube is bounded by the current steam generator tube 
rupture (SGTR) analysis described in the CPSES FSAR [Final Safety 
Analysis Report]. Due to the slight reduction in diameter caused by 
the sleeve wall thickness, primary coolant release rates would be 
slightly less than assumed for the steam generator tube rupture 
analysis, depending on the break location, and therefore, would 
result in lower total primary fluid mass release to the secondary 
system. A main steam line break or feed line break will not cause a 
SGTR since the sleeves are analyzed for a maximum accident 
differential pressure greater than that predicted in the CPSES 
safety analysis. The proposed reduction of the steam generator 
primary to secondary operational leakage limit provides added 
assurance that leaking flaws will not propagate to burst prior to 
commencement of plant shutdown.
    In conclusion, based on the discussion above, these changes will 
not significantly increase 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?
    Response: No.
    The Westinghouse Leak Tight Sleeves are designed using the 
applicable ASME Code as guidance; therefore, they meet the 
objectives of the original steam generator tubing. As a result, the 
functions of the steam generators will not be significantly affected 
by the installation of the proposed sleeves. The proposed repair 
sleeves do not interact with any other plant systems. Any accident 
as a result of potential tube or sleeve degradation in the repaired 
portion of the tube is bounded by the existing tube rupture accident 
analysis. The continued integrity of the installed sleeve is 
periodically verified by the Technical Specification requirements.
    The implementation of the proposed amendment has no significant 
effect on either the configuration of the plant or the manner in 
which it is operated. As discussed above, the reduced primary to 
secondary leakage limit is considered a conservative change in the 
plant limiting conditions for operation. Therefore, TXU Electric 
concludes that this proposed change does 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?
    Response: No.
    The repair of degraded steam generator tubes with Westinghouse 
Leak Tight Sleeves restores the structural integrity of the degraded 
tube under normal operating and postulated accident conditions. The 
design safety factors utilized for the repair sleeves are consistent 
with the safety factors in the ASME Code used in the original steam 
generator design. The portions of the installed sleeve assembly that 
represents the reactor coolant pressure boundary can be monitored 
for the initiation and progression of sleeve/tube wall degradation. 
Use of the

[[Page 66474]]

previously identified design criteria and design verification 
testing assures that the margin of safety is not significantly 
different from the original steam generator tubes. The proposed 
sleeve inspection requirements are more stringent than existing 
requirements for inspection of the steam generator tubes, and the 
reduction in the operational limit for primary to secondary leakage 
through the steam generator tubes is more conservative than current 
requirements. Therefore, TXU Electric concludes that the proposed 
change does not involve a significant reduction in a margin of 
safety.
    EPRI [Electric Power Research Institute] qualified eddy current 
techniques will be used for the detection of tube degradation in 3/4 
inch welded sleeved tubes. Alternate inspection techniques, may be 
used as they become available, as long as it can be demonstrated 
that the technique used provides the same degree or greater degree 
of inspection rigor.
    The effect of sleeving on the design transients and accident 
analyses were reviewed and found to remain valid up to the level of 
steam generator tube plugging consistent with the minimum reactor 
flow rate as specified in Technical Specification 3.4.1. Continued 
compliance with the RCS [Reactor Coolant System] flow limits of 
Technical Specification 3.4.1 is assured through precision flow 
measurements.
    Because all relevant safety analyses were reviewed and found to 
remain valid, and because the appropriate design margins are 
maintained through compliance with the relevant ASME Code 
requirements, it is concluded that 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: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: November 20, 2001.
    Description of amendment request: The proposed amendment would: (1) 
Move Table 4.7.2, ``Primary Containment Isolation Valves'' and 
references to the Table from the Vermont Yankee Nuclear Power Station 
(VY) Technical Specifications (TSs) to the Technical Requirements 
Manual; (2) change Surveillance Requirement 4.7.B.1.b to reflect that 
the Standby Gas Treatment system (SBGT) duct heater needs to meet 
relative humidity design basis; (3) add section 3.7.E, ``Reactor 
Building Automatic Ventilation System Isolation Valves,'' to the Table 
of Contents; (4) remove wording in 3.5.A.4.a and b referencing a one-
time 30-day Limiting Condition for Operation; and (5) make 
administrative changes to Sections 5.3 and 6.4.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. The operation of the Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    The proposed changes consist of removal of the primary containment 
isolation valve component list from the VY TS, revision of the SBGT 
inlet heater surveillance minimum power rating and other administrative 
changes. The probability of occurrence of a previously evaluated 
accident is not increased because neither containment isolation nor the 
SBGT heater are accident initiators, and the proposed changes do not 
impact any accident initiating conditions. The consequences of an 
accident previously evaluated are not increased because the proposed 
changes do not impact the ability of containment to restrict, or SBGT 
to filter, the release of any fission product radioactivity to the 
environment. The proposed changes to remove the primary containment 
isolation valve component list from TS, relocate the information to a 
licensee controlled document, and to change the SBGT inlet heater power 
input surveillance requirement, will have no significant impact on any 
safety related structures, systems or components. The TS requirements 
for the primary containment isolation valves and SBGT operability and 
surveillance will not be changed. Additionally, the administrative 
changes do not affect any system operation or function.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve any physical alteration of 
plant equipment and do not change the method by which any safety-
related system performs its function. No new or different types of 
equipment will be installed. The proposed changes do not create any new 
accident initiators or involve an activity that could be an initiator 
of an accident of a different type.
    Therefore, the proposed changes will not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a significant 
reduction in a margin of safety.
    The proposed administrative changes, the removal of the primary 
containment isolation valve component list from TS and the change to 
the SBGT inlet heater power input surveillance requirement, do not 
alter the TS requirements for containment integrity, containment 
isolation, SBGT operability, or adversely affect their capability. The 
changes will not alter the basic operation of process variables, 
systems, or components as described in the safety analysis. No new 
equipment is introduced.
    The proposed changes do no impact design margins of the primary 
containment isolation system, SBGT or any other system to perform their 
safety functions. The essential safety functions of providing primary 
containment integrity and providing filtration of airborne radioactive 
releases, are maintained. There is no physical or operational change 
being made which would alter the sequence of events, plant response, or 
margins in existing safety analyses. The proposed changes result in no 
impact on analyzed accident event precursors or effects.
    These proposed changes do not alter the physical design of the 
plant. There is no change in methods of operation. The proposed changes 
do not alter the means by which primary containment isolation 
capability is maintained and SBGT is operated.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

[[Page 66475]]

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/NRC/ADAMS/index.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by email to [email protected].

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck 
Plant, Middlesex County, Connecticut

    Date of amendment request: May 30, 2001, as supplemented on 
November 6, 2001.
    Brief description of amendment: Connecticut Yankee Atomic Power 
Company (the licensee) requested changes to the Technical 
Specifications (TSs) for the Haddam Neck Plant. The changes to Sections 
5 and 6 of the TSs correct terminology, clarify the specifications for 
consistency with established programs and Standard TSs, and reflect 
current plant conditions. The changes also reflect the licensee's 
current organization titles. For information only, the licensee also 
included proposed changes to the TS Bases for spent fuel pool water 
level and cooling. The NRC staff did not review the proposed changes to 
the TS Bases.
    Date of issuance: December 4, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 196.
    Facility Operating License No. DPR-61: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44164).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 4, 2001.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut Inc., et al., Docket Nos. 50-336 and 50-
423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London 
County, Connecticut Date of application for amendment: August 9, 2001

    Brief Description of amendment: The amendments modify the Millstone 
Nuclear Power Station, Unit Nos. 2 and 3 Technical Specifications to 
clarify the licensed operator qualification standards.
    Date of issuance: December 5, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 258 and 199.
    Facility Operating License Nos. NPD-69 and NPF-49: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 17, 2001 (66 FR 
52798).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 5, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 23, 2001.
    Brief description of amendment: The amendment approves a change to 
Technical Specification (TS) 3.8.1.1, ``Electrical Power System--A.C. 
Sources.'' The change removes Surveillance Requirement 4.8.1.1.2.c.1 
regarding Emergency Diesel Generator inspection at least once per 18 
months during shutdown condition.
    Date of issuance: December 7, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 259.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31705).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 7, 2001.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina Date of 
application for amendments: June 13, 2000, as supplemented August 30 
and September 10, 2001.
    Brief description of amendments: The amendments revised the 
Facility Operating License of each unit to (1) delete license 
conditions that have been fulfilled; and (2) make other corrections and 
editorial changes.
    Date of issuance: December 5, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 200 and 181.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Facility Operating License.
    Date of initial notice in Federal Register: November 1, 2000 (65 FR 
65341).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 5, 2001.
    No significant hazards consideration comments received: No.

[[Page 66476]]

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: June 15, 2001.
    Brief description of amendments: Eliminate the Technical 
Specifications (TS) requirement that the Automatic Depressurization 
System (ADS) designated Safety/Relief Valves (S/RVs) open during the 
manual actuation of the ADS and rewords the Surveillance Requirement 
(SR) 3.5.1.8 frequency to require the testing of all required ADS 
manual actuation solenoids during the performance of SR 3.5.1.8 in 
place of testing on a staggered basis.
    Date of issuance: December 13, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 151 and 137.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41618).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 13, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 13, 2001.
    Brief description of amendments: These amendments delete Technical 
Specifications (TS) Section 6.8.4, which required a Post-Accident 
monitoring program, for Beaver Valley Power Station, Unit Nos. 1 and 2, 
and thereby eliminate the requirements to have and maintain the post-
accident sampling system (PASS) for those units.
    Date of Issuance: December 6, 2001.
    Effective date: Upon issuance and shall be implemented within 180 
days.
    Amendment Nos.: 245, 123.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 19, 2001 (66 
FR 48286).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 6, 2001.
    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: May 17, 2001, as supplemented 
by letter dated September 5, 2001.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3/4.9.3, ``Decay Time,'' to allow the start of core 
offload at 100 hours after reactor subcriticality between September 15 
and June 15, when the lake temperature is assumed to be not higher than 
77.8 deg.F, and 148 hours after reactor subcriticality between June 16 
and September 14, when the lake temperature is assumed to be not higher 
than 85 deg.F. TS 3/4.9.3 currently prohibits fuel movement in the 
reactor pressure vessel until the reactor has been subcritical for at 
least 168 hours. The 168-hour decay time was placed in the CNP TS with 
Amendment Nos. 169 and 152 to DPR-58 and DPR-74, respectively, on 
January 14, 1993.
    Date of issuance: November 30, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 260 and 243.
     Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44174)
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 30, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 30, 2001, as supplemented 
October 10 and November 16, 2001.
    Brief description of amendment: The amendment revises the Technical 
Specification safety limit minimum critical power ratio for two 
recirculation pump operation for Cycle 21.
    Date of issuance: December 6, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 125.
    Facility Operating License No. DPR-22: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 3, 2001 (66 FR 
50470)
    The October 10 and November 16, 2001, supplements provided 
clarifying information that was within the scope of the original 
Federal Register notice and did not change the staff's initial proposed 
no significant hazards considerations determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated December 6, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: February 7, 2001, as supplemented by 
letters dated October 17 and November 2, 2001.
    Brief description of amendment: The requested changes replaced the 
current accident source term used in the design basis radiological 
analyses for control room habitability with an alternative source term 
(AST) pursuant to 10 CFR 50.67, ``Accident Source Term.'' OPPD 
requested a full implementation of the AST. Changes were also made to 
the Ft. Calhoun Technical Specifications to make them consistent with 
the revised associated accident analysis.
    Date of issuance: December 5, 2001.
    Effective date: December 5, 2001, to be implemented within 60 days 
from the date of issuance.
    Amendment No.: 201.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 2, 2001 (66 FR 
22031).
    The October 17 and November 2, 2001, supplemental letters provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 5, 2001.
    No significant hazards consideration comments received: No.

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 amendments request: May 3, 2001.
    Brief Description of amendments: The amendments relocate cycle-
specific

[[Page 66477]]

reactor coolant system parameter limits from the Technical 
Specifications (TS) and associated Bases, to the Core Operating Limits 
Report. The amendments also, add a reference to the Refueling Boron 
Concentration to TS 5.6.5 to correct an omission.
    Date of issuance: December 4, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 151 and 143.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55024).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 4, 2001.
    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: February 12, 2001.
    Brief description of amendments: The amendments consist of deleting 
Surveillance Requirement 4.4.6.2.2.e of South Texas Project Technical 
Specifications Section 3/4.4.6.2.
    Date of issuance: December 11, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--134; Unit 2--123.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31715).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 11, 2001.
    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 14, 2000.
    Brief description of amendment: These amendments revise Technical 
Specifications Sections 4.7.7.1.d.1 and 4.7.7.2.a. These changes 
increase the specified minimum number of compressed bottles of air from 
84 to 102, and revise the differential pressure limit across the 
Control Room Emergency Ventilation System HEPA Filter, demister filter, 
and charcoal adsorber.
    Date of issuance: December 12, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 228 and 209.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments change 
the Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7687).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 12, 2001.
    No significant hazards consideration comments received: No.

    (Note: The publication date for this notice will change from 
every other Wednesday to every other Tuesday, effective January 8, 
2002. The notice will contain the same information and will continue 
to be published biweekly.


    Dated at Rockville, Maryland, this 17th day of December, 2001.
    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 01-31473 Filed 12-21-01; 8:45 am]
BILLING CODE 7590-01-P