[Federal Register Volume 65, Number 66 (Wednesday, April 5, 2000)]
[Notices]
[Pages 17908-17930]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-8211]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 11 through March 24, 2000. The last 
biweekly notice was published on March 22, 2000 (65 FR 15375).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
Involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) Create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) Involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the

[[Page 17909]]

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 
Review and Directives Branch, Division of Freedom of Information and 
Publications 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, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By May 5, 2000, the licensee may file a request for a hearing with 
respect to issuance of the amendment to the subject facility operating 
license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC, and electronically from 
the ADAMS Public Library component on the  NRC Web site,   http://www.nrc.gov (the Electronic Reading Room). If a request for a hearing 
or petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    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: Docketing and 
Services Branch; or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and electronically from the ADAMS Public 
Library component on the NRC Web site,   http://www.nrc.gov (the 
Electronic Reading Room).

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of amendment's request: February 15, 2000.
    Description of amendment's request: The proposed amendments would 
revise the technical specifications to permit use of the Westinghouse 
core monitoring and support system known as Best Estimate Analyzer for 
Core Operations Nuclear (BEACON).

[[Page 17910]]

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

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Power Distribution Monitoring System (PDMS) performs continuous 
core power distribution monitoring. It in no way provides any 
protection or control system functionality. Fission product barriers 
are not impacted by these proposed changes. The proposed changes 
occurring with PDMS will not result in any additional challenges to 
plant equipment that could increase the probability of any 
previously evaluated accident. The changes associated with the PDMS 
do not affect plant systems such that their function in the control 
of radiological consequences is adversely affected. These proposed 
changes will therefore not affect the mitigation of the radiological 
consequences of any accident described in the Updated Final Safety 
Analysis Report (UFSAR).
    Continuous on-line monitoring through the use of PDMS provides 
significantly more information about the power distributions present 
in the core than is currently available. This results in more time 
(i.e., earlier determination of an adverse condition developing) for 
operator action prior to having any adverse condition develop that 
could lead to an accident condition or to unfavorable initial 
conditions for an accident.
    Each accident analysis addressed in the Byron and Braidwood 
Stations' UFSAR will be examined with respect to changes in cycle-
dependent parameters, which are obtained from application of the NRC 
approved reloaddesign methodologies, to ensure that the transient 
evaluation of new reloads are bounded by previously accepted 
analyses. This examination, which will be performed in accordance 
with the requirements set forth in 10 CFR 50.59, ``Changes, tests 
and experiments,'' will ensure that future reloads will not involve 
a significant increase in the probability or consequences of any 
accident previously evaluated.
    The proposed change, therefore, does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    As stated previously, the implementation of the PDMS system has 
no influence or impact on plant operations or safety, nor does it 
contribute in any way to the probability or consequences of an 
accident. No safety-related equipment, safety function, or plant 
operation will be altered as a result of this proposed change. The 
possibility for a new or different type of accident from any 
accident previously evaluated is not created since the changes 
associated with PDMS does not result in a change to the design basis 
of any plant component or system. The evaluation of the effects of 
the PDMS changes shows that all design standards and applicable 
safety criteria limits are met. These changes, therefore, do not 
cause the initiation of any accident nor create any new failure 
mechanisms. All equipment important to safety will operate as 
designed. Component integrity is not challenged. The proposed 
changes do not result in any event previously deemed incredible 
being made credible. The PDMS changes will not result in more 
adverse conditions and will not result in any increase in the 
challenges to safety systems. The cycle specific variables required 
by the PDMS are calculated using NRC approved methods. The Technical 
Specifications (TS) will continue to require operation within the 
required core operating limits and appropriate actions will be taken 
when or if limits are exceeded.
    The proposed change, therefore, does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    The margin of safety is not affected by the implementation of 
PDMS. The margin of safety presently provided by current TS remains 
unchanged. Appropriate measures exist to control the values of these 
cycle-specific limits. The proposed changes continue to require 
operation within the core limits that are based on NRC approved 
reload design methodologies. The proposed changes continue to ensure 
that appropriate actions will be taken if limits are violated. These 
actions remain unchanged. The development of the reload specific 
limits, including Relaxed Axial Offset Control (RAOC) bands, for 
future reloads will continue to conform to those methods described 
in NRC approved documentation. In addition, each future reload 
involves a 10 CFR 50.59, ``Changes, tests and experiments,'' safety 
review to assure that operation of the units, within the cycle-
specific limits, will not involve a reduction in margin of safety.
    The proposed changes, therefore, do not impact the operation of 
the Byron and Braidwood Stations in any manner that involves a 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: February 18, 2000.
    Description of amendment request: The proposed amendments would 
remove the anticipatory reactor scram signal for turbine electro-
hydraulic control (EHC) low oil pressure trip from the reactor 
protection system (RPS) trip function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    The proposed changes remove the ``Turbine Electro-Hydraulic 
Control (EHC) Control Oil Pressure-Low'' scram function and the 
associated Limiting Safety System Setting (LSSS). The purpose of the 
Turbine EHC Control Oil Pressure scram is to anticipate the pressure 
transient which would be caused by imminent control valve fast 
closure on loss of control oil pressure. This function does not 
serve as an initiator for any accidents evaluated in Chapter 15 of 
the Updated Final Safety Analysis Report (UFSAR). In addition, this 
trip function is not credited in any design basis event and is 
functionally redundant to the Turbine Control Valve Fast Closure RPS 
trip function during a loss of EHC control oil. The Turbine Control 
Valve Fast Closure will initiate a scram on a loss of control oil 
event coincident with turbine control valve closure.
    Therefore, these 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 removal of this function does not represent a change in 
operating parameters or introduce a new mode of operation. The 
pressure switches associated with the Turbine Control Valve Fast 
Closure function provide equivalent protection from a loss of EHC 
oil. For this reason, the 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?
    Operation with the proposed changes in place will not change any 
plant operating parameters, nor any protective system actuation 
setpoints other than removal of the Turbine EHC Control Oil 
Pressure-Low scram function. The scram function associated with the 
Turbine Control Valve Fast Closure provides equivalent protection 
for events involving turbine control valve fast closure

[[Page 17911]]

including the loss of EHC control oil pressure. For this reason, 
eliminating the EHC Control Oil Pressure-Low scram function, which 
is redundant to other protective instrumentation, does not reduce 
the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: February 23, 2000.
    Description of amendment request: The proposed amendments would 
change the pressure-temperature (P-T) limits by revising the heatup, 
cooldown and inservice test limitations for the Reactor Pressure Vessel 
(RPV) to a maximum of 32 Effective Full Power Years (EFPY).
    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 of occurrence or consequences of an accident previously 
evaluated?
    The proposed changes do not modify the reactor coolant pressure 
boundary, do not make changes in operating pressure, materials or 
seismic loading. The proposed changes adjust the reference 
temperature for the limiting beltline material to account for 
radiation effects and provide the same level of protection as 
previously evaluated. The proposed changes do not adversely affect 
the integrity of the Reactor Coolant System (RCS) such that its 
function in the control of radiological consequences is affected. 
Therefore, 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 create the possibility of a new or 
different kind of accident previously evaluated for Dresden Nuclear 
Power Station. No new modes of operation are introduced by the 
proposed changes. The proposed changes will not create any failure 
mode not bounded by previously evaluated accidents. Use of the 
revised P-T curves will continue to provide the same level of 
protection as was previously reviewed and approved.
    Further, the proposed changes to the P-T curves do not affect 
any activities or equipment, and are not assumed in any safety 
analysis to initiate any accident sequence for Dresden Nuclear Power 
Station. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed changes reflect an update of the P-T curves to 
extend the RPV operating limit to 32 Effective Full Power Years 
(EFPYs). The revised curves are based on the latest American Society 
of Mechanical Engineers (ASME) guidance and actual operational data 
for the units. These proposed changes are acceptable because the 
ASME guidance maintains the relative margin of safety commensurate 
with that which existed at the time that the ASME Section XI 
Appendix G was approved in 1974. 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: February 29, 2000.
    Description of amendment request: The proposed amendments would 
revise the pressure-temperature (P-T) limits for heatup, cooldown, 
critical operation and inservice leak and hydrostatic test limitations 
for the reactor pressure vessel (RPV). The proposed changes replace the 
current RPV P-T limit curves with three recalculated curves that are 
applicable to 32 effective full power years.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    The proposed changes to the LaSalle County Station reactor 
pressure vessel (RPV) pressure-temperature (P-T) limits do not 
modify the boundary, operating pressure, materials or seismic 
loading of the rector coolant system. The proposed changes do adjust 
the P-T limits for radiation effects to ensure that the RPV fracture 
toughness is consistent with analysis assumptions and NRC 
regulations. Thus, the proposed changes do not involve a significant 
increase in the probability of occurrence of an accident previously 
evaluated.
    The proposed changes do not adversely affect the integrity of 
the reactor coolant system such that its function in the control of 
radiological consequences is affected. Therefore, the proposed 
changes do not involve a significant increase in the 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?
    The proposed changes to the reactor pressure vessel pressure-
temperature limits do not affect the assumed accident performance of 
any structure, system or component previously evaluated. The 
proposed changes do not introduce any new modes of system operation 
or failure mechanisms. Therefore, the proposed changes do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the change involve a significant reduction in the margin 
of safety?
    Appendices G, ``Fracture Toughness Requirements,'' and H, 
``Reactor Vessel Material Surveillance Program Requirements,'' of 10 
CFR 50 describe specific requirements for fracture toughness and 
reactor vessel material surveillance that must be considered in 
establishing P-T limits. Appendix G of 10 CFR 50 specifies fracture 
toughness and testing requirements for reactor vessel material in 
accordance with the American Society of Mechanical Engineers (ASME) 
Boiler and Pressure Vessel (B&PV) Code and that the beltline 
material in the surveillance capsules be tested in accordance with 
Appendix H of 10 CFR 50. Appendix G also requires the prediction of 
the effects of neutron irradiation on the vessel embrittlement. 
Generic Letter 88-11, ``NRC Position on Radiation Embrittlement of 
Reactor Vessel Materials And Its Impact on Plant Operations,'' 
requests that the methods in Regulatory Guide 1.99, Revision 2, 
``Effects of Residual Elements on Predicted Radiation Damage to 
Reactor Vessel Material,'' be used to predict the effect of neutron 
irradiation on the reactor vessel material.
    The current P-T limits for LaSalle County Station were approved 
by the NRC in Amendment No. 71 for Unit 1 and Amendment No. 55 for 
Unit 2. The NRC approval of the current pressure-temperature limits 
was based on their conformance to the requirements of Appendices G 
and H of 10 CFR 50. The NRC also noted that current P-T limits 
satisfied Generic Letter 88-11

[[Page 17912]]

because the method in Regulatory Guide 1.99, Revision 2 was used to 
calculate the Adjusted Reference Temperature (ART).
    The methodology used to generate the revised P-T limits in the 
proposed changes is similar to the methodology used to generate the 
currently approved P-T limits, in conformance with the requirements 
of Appendices G and H of 10 CFR 50, consistent with the methods of 
Regulatory Guide 1.99, Revision 2, and consistent with the 
calculations contained in our July 14, 1999, proposed TS change for 
power uprate operation. These proposed changes are acceptable 
because the ASME B&PV Code guidance maintains the relative margin of 
safety commensurate with that which existed at the time that the 
ASME B&PV Code Section XI, ``Rules for Inservice Inspection of 
Nuclear Power Plant Components,'' Appendix G was approved in 1974. 
Thus, the proposed change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, PO Box 767, 
Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

Consolidated Edison Company of New York, Inc., Docket No. 50-003, 
Indian Point Nuclear Generating Station, Unit 1, Buchanan, New York

    Date of application for amendment: February 14, 2000.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications Sections 2.10.2, 3.1.2, 3.2.1, 
4.1.8.1.b, and 4.1.8.1. Specifically, Sections 3.1.2, 3.2.1, and 
4.1.8.1.b, are organizational title changes that are administrative in 
nature and reflect a streamlining of the Consolidated Edison Company of 
New York, Inc.'s, management structure. Section 4.1.8.1 is changed to 
reference the current sections of Part 20 of Title 10 of the Code of 
Federal Regulations (10 CFR) and to remove any ambiguity that may exist 
by referring to obsolete sections of the regulations. A footnote was 
moved from Section 2.11 to Section 2.10.2.6 to improve the clarity of 
the Technical Specification since it pertains to text in subsection 
2.10.2.4.
    Basis for proposed no significant hazards 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.
    (a) Changes to Sections 3.1.2, 3.2.1, and 4.1.8.1.b To Reflect 
Organizational Title Changes
    (1) Does the proposed license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    No. The proposed change is administrative in nature. The changes 
involve updating Sections 3.2.1.h and 4.1.8.b to use the title 
``Shift Manager'' instead of ``Senior Watch Supervisor'' and 
updating Section 3.1.2 and 3.1.2.b to use the title ``Plant 
Manager'' instead of ``General Manager--Nuclear Power Generation'' 
and movement of the footnote, ``*Licensed Operator for IP2.'' These 
changes do not affect possible initiating events for accidents 
previously evaluated or alter the configuration or operation of the 
facility. The Limiting Safety System Settings and Safety Limits 
specified in the current Technical Specifications remain unchanged. 
Therefore, the proposed changes would not involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed changes are administrative in nature. The 
safety analysis of the facility remains complete and accurate. There 
are no physical changes to the facility and the plant conditions for 
which the design basis accidents have been evaluated are still 
valid. The operating procedures and emergency procedures are 
unaffected. Consequently no new failure modes are introduced as a 
result of the proposed change. Therefore, the proposed changes would 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed changes are administrative in nature. Since 
there are no changes to the operation of the facility or the 
physical design, the Updated Final Safety Analysis Report (UFSAR) 
design basis, accident assumptions, or Technical Specification Bases 
are not affected. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.
    (b) Change to Section 4.1.8.1 to Reference the Current Sections 
of 10 CFR 20
    (1) Does the proposed license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    No. The proposed change [to Section 4.1.8.1] is administrative 
in nature. The change involves updating Section 4.1.8.1 to reference 
10 CFR 20.1601(a) and 10 CFR 20.1601(b). This change does not affect 
possible initiating events for accidents previously evaluated or 
alter the configuration or operation of the facility. The Limiting 
Safety System Settings and Safety Limits specified in the current 
Technical Specifications remain unchanged. Therefore, the proposed 
change would not involve a significant increase in the probability 
or in the consequences of an accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed change is administrative in nature. The safety 
analysis of the facility remains complete and accurate. There are no 
physical changes to the facility and the plant conditions for which 
the design basis accidents have been evaluated are still valid. The 
operating procedures and emergency procedures are unaffected. 
Consequently no new failure modes are introduced as a result of the 
proposed change. Therefore, the proposed change would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed change is administrative in nature. Since there 
are no changes to the operation of the facility or the physical 
design, the Updated Final Safety Analysis Report (UFSAR) design 
basis, accident assumptions, or Technical Specification Bases are 
not affected. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

    The Nuclear Regulatory Commission (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: Brent L. Brandenburg, Esq., Consolidated 
Edison Co. of New York, Inc., 4 Irving Place-1830, New York, NY 10003.
    NRC Section Chief: Michael Masnik.

Consolidated Edison Company of New York, Inc., Docket Nos. 50-003, 50-
247 Indian Point Nuclear Generating Station, Units 1 and 2, Buchanan, 
New York

    Date of application for amendment: February 14, 2000.
    Description of amendment request: The proposed amendment to the 
Indian Point Nuclear Generating Station, Unit Nos. 1 and 2, 
Environmental Technical Specifications (ETS) would change Section 
5.4.1, eliminating the discussions of Section 4.2. Specifically, in ETS 
Section 5.4.1, Routine Reports, the proposed change seeks to delete the 
reference to and discussions about Section 4.2, which was deleted from 
the Unit 2 Operating License as part of Amendment #90. The change is 
administrative in nature and improves the clarity of the ETS by 
eliminating the reference to a section that no longer exists.
    Basis for proposed no significant hazards determination: As 
required by

[[Page 17913]]

10 CFR 50.91(a), the licensee has provided its analysis of the issue of 
no significant hazards consideration, which is presented below.

    (1) Does the proposed license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    No. The proposed change is administrative in nature. The change 
involves deleting, in Section 5.4.1, the reference to and the 
discussions about Section 4.2, which no longer exists. The 
monitoring requirements specified in the current Environmental 
Technical Specifications remain unchanged. Therefore, the proposed 
changes would not involve a significant increase in the probability 
or in the consequences of an accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed changes are administrative in nature. The 
safety analysis of the facility remains complete and accurate. There 
are no physical changes to the facility and the plant conditions for 
which the design basis accidents have been evaluated are still 
valid. The operating procedures and emergency procedures are 
unaffected. Consequently no new failure modes are introduced as a 
result of the proposed change. Therefore, the proposed changes would 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed changes are administrative in nature. Since 
there are no changes to the operation of the facility or the 
physical design, the Updated Final Safety Analysis Report (UFSAR) 
design basis, accident assumptions, or Technical Specification Bases 
are not affected. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.

    The Nuclear Regulatory Commission (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: Brent L. Brandenburg, Esq., Consolidated 
Edison Co. of New York, Inc., 4 Irving Place-1830, New York, NY 10003.
    NRC Section Chief: Michael Masnik.

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

    Date of amendment request: November 22, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Sections 3.7.B.1 and 3.7.B.2 to 
reference American Society for Testing and Materials (ASTM) D3803-1989 
for testing charcoal samples from the standby gas treatment system 
(SGTS) and the control room high efficiency air filtration systems 
(CRHEAFS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    (1) The operation of Pilgrim Station in accordance with the 
proposed amendment will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The accident analyses performed to ensure compliance with the dose 
limits of 10 CFR Part 100 and 10 CFR Part 50, Appendix A, GDC 19, use 
assumptions regarding SGTS and CRHEAFS performance. The analyses assume 
SGTS train efficiency for radioiodine removal of 99% and CRHEAFS train 
efficiency of 95%. They also assume individual charcoal bank 
efficiencies of 95%.
    Obtaining charcoal samples from both systems in accordance with 
Regulatory Position C.6.b of Regulatory Guide (RG) 1.52, Revision 2, 
March 1978, ensures the laboratory tests a representative sample of the 
activated charcoal in each system. Testing these samples in accordance 
with ASTM D3803-1989 at a temperature of 86  deg.F and 70% RH [relative 
humidity] ensures accurate and reproducible test results are obtained. 
Specifying the allowable removal efficiency as 97.5% ensures 
an appropriate safety factor is applied. This safety factor is 
consistent with GL 99-02. Inlet methyl iodide concentrations are 
specified by ASTM D3803-1989. Finally, increasing the acceptance 
criteria for halogenated hydrocarbon tests to 99.9% ensures system 
performance is consistent with accident analysis assumptions.
    No accident initiators are affected by the proposed change. 
Increasing charcoal adsorber efficiency and reducing allowable bypass 
leakage ensures SGTS and CRHEAFS performance are consistent with that 
assumed in Pilgrim's accident analyses. Therefore, the postulated 
consequences are unchanged from the previously evaluated analyses.
    There are no safety consequences and environmental impacts 
associated with the TS 5.0 pagination revision. The proposed pagination 
revision incorporates, in orderly fashion, pages approved by Amendments 
177 and 179.
    (2) The operation of Pilgrim 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.
    No new or different types of accidents or malfunctions than those 
previously analyzed in the Updated Final Safety Analysis Report are 
introduced by this proposed change because there are no new failure 
modes being introduced. Rather, the changes being proposed reduce the 
possibility that existing failure modes could occur. As discussed above 
in the first part of this No Significant Hazards Consideration, 
specifying sampling and testing of charcoal adsorber banks to NRC 
approved standards, increasing charcoal efficiency requirements and 
reducing allowable bypass leakage does not challenge plant safety and 
will not create the possibility of a new or different kind of accident 
from any accident previously analyzed.
    There are no safety consequences and environmental impacts 
associated with the TS 5.0 pagination revision. The proposed pagination 
revision incorporates, in orderly fashion, pages approved by Amendments 
177 and 179.
    (3) The operation of Pilgrim Station in accordance with the 
proposed amendment will not involve a significant reduction in the 
margin of safety.
    Collecting charcoal for testing in accordance with RG 1.52 ensures 
a representative sample is obtained. Testing the sample in accordance 
with ASTM D3803-1989 at 86  deg.F and 70% RH ensures accurate and 
reproducible results are obtained. Increasing the minimum allowable 
charcoal efficiency from 95% to 97.5% increases the margin of safety. 
Increasing the minimum allowable halogenated hydrocarbon removal 
requirement from 99% to 99.9% also increases the margin of safety.
    There are no safety consequences and environmental impacts 
associated with the TS 5.0 pagination revision. The proposed pagination 
revision incorporates, in orderly fashion, pages approved by Amendments 
177 and 179.
    Based on the staff's analysis, it appears that the three standards 
of 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: W. S. Stowe, Esquire, Entergy Nuclear 
Generation Company, 800 Boylston Street, 36th Floor, Boston, 
Massachusetts 02199.

[[Page 17914]]

    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: March 8, 2000.
    Description of amendment request: The proposed amendment would 
change the technical specification definition of core alteration from 
``* * * the movement or manipulation of any component within the 
reactor pressure vessel with the vessel head removed and fuel in the 
vessel* * *'' to ``* * * the movement or manipulation of any fuel, 
sources, or reactivity control components [excluding coupling/
uncoupling of CEAs [control element assemblies]] within the reactor 
vessel with the vessel head removed and fuel in the vessel.* * *''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

    The intent of the definition is to ensure that activities which 
could result in reactivity changes or have the potential to cause 
fuel damage are considered a core alteration. The current definition 
could be [interpreted] to apply to other activities that would not 
result in reactivity changes or have the potential to cause fuel 
damage. Thus, the modification of the definition clarifies the 
wording such that movement of only those components that result in 
reactivity changes or have the potential to cause fuel damage are 
specified. The modified NUREG-1432 [Standard Technical 
Specifications, Combustion Engineering Plants] definition was 
derived to limit those actions that could cause reactivity changes 
and potentially affect the probability or consequences of fuel 
handling accidents. Therefore, changing the definition of a core 
alteration to movement of those components that directly affect 
reactivity will not result in an increase in the probability or 
consequences associated with a fuel handling accident.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.

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

    The proposed definition identifies specific components that if 
moved or manipulated would result in reactivity changes. The 
movement or manipulation of items such as lights, video cameras, and 
reactor vessel material specimen capsules within the reactor vessel 
will not result in changes in reactivity. Additionally, no 
reactivity change would result with the withdrawal and insertion of 
incore detectors or the movement of the reactor vessel upper 
internals within the reactor vessel with fuel in the vessel.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3  Does Not Involve a Significant Reduction in the Margin of 
Safety

    The core alteration definition is based on the need for control 
of reactivity changes and the consequences of fuel handling 
accidents. The proposed change provides clarity as to what component 
movement or manipulation results in reactivity changes. The proposed 
change is in accordance with the guidance provided in NUREG-1432 for 
a core alteration.

    Therefore, this change does not involve a significant reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: March 9, 2000.
    Description of amendment request: The proposed amendment would 
revise the license as follows:

    For Cycle 14 only, Entergy Operations[, Inc.] shall be permitted 
to operate the reactor based on a risk-informed demonstration that 
predicted steam generator tube integrity, with consideration of 
eggcrate axial flaws, is adequate to meet Regulatory Guide 1.174 
numerical acceptance criteria. In accordance with Principle 5 in 
Regulatory Guide 1.174 concerning monitoring operational experience 
to ensure that performance is consistent with risk predictions, if 
Entergy Operations plugs or repairs steam generator tubes during 
Cycle 14, then the steam generators shall be reinspected to the 
extent necessary to verify that they have been returned to a 
condition consistent with the risk assessment.

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

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

    A steam generator tube rupture is an accident previously 
evaluated in the ANO-2 [Arkansas Nuclear One, Unit 2] Safety 
Analysis Report. The probability of tube burst under design basis 
accident conditions is only slightly increased by the proposed 
change due to the minor reduction in margin of safety associated 
with tubing structural integrity, but is within the current industry 
guidance of NEI [Nuclear Energy Institute] 97-06, ``Steam Generator 
Program Guidelines.'' Detailed studies have been performed to 
evaluate the probable condition of the steam generator tubing for 
the remainder of cycle 14 operation. These studies show less than a 
0.1 percent increase in the probability of tube rupture under worst 
case design basis accident conditions as a result of the proposed 
change.
    This change does not modify any parameter that will increase 
radioactivity in the primary system or increase the amount of 
radioactive steam released from the secondary safety valves or 
atmospheric dump valves in the event of a tube rupture.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.

Criterion 2  Does Not Create the Possibility of a New or Different Kind 
of Accident from any Previously Evaluated

    The scope of this change does not establish a potential new 
accident precursor. The design basis accident analyses for ANO-2 
include the consequences of a double-ended break of one steam 
generator tube which bounds other postulated failure mechanisms. The 
proposed change does not modify any mode of operation or modify 
existing periodic inservice inspection requirements.
    Therefore, this change does not create the possibility or a new 
or different kind of accident from any previously evaluated.

Criterion 3  Does Not Involve a Significant Reduction in the Margin of 
Safety

    The proposed change justifies a minor reduction in the steam 
generator tubing structural integrity margin of safety of three 
times normal differential operating pressure (4050 psi). However, 
the margin of safety for a tube burst still remains well in excess 
of the 2500 psi maximum differential pressure used in the design 
basis accident analysis for a main steam line break. The proposed 
change is technically consistent with the criteria of NEI 97-06 and 
Regulatory Guide 1.174, ``An Approach for Using Probabilistic Risk 
Assessment in Risk-Informed Decisions on Plant-Specific Changes to 
the Licensing Basis.''
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

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

[[Page 17915]]

amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: November 29, 1999, supplemented December 
20, 1999.
    Description of amendment request: The proposed amendment would add 
license condition 2.C(12) to allow a one-time extension of the steam 
generator inspection interval of Technical Specification 4.4.5.3.a. 
This would allow the steam generator inspection interval to coincide 
with the 8th refueling outage scheduled to begin in September 2000.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change is temporary and allows a one time extension 
of the steam generator (SG) surveillance requirement (SR) for Cycle 
8 to allow surveillance testing to coincide with the 8th refueling 
outage (2R8). The proposed surveillance interval extension will not 
cause a significant reduction in system reliability nor affect the 
ability of a system to perform its design function. Current 
monitoring of plant conditions and the surveillance monitoring 
required during normal plant operation will be performed as usual to 
assure conformance with technical specification (TS) operability 
requirements.
    The TS SG tube inspection is intended to prevent the ``Steam 
Generator Tube Failure'' analyzed in [Updated Final Safety Analysis 
Report] UFSAR Section 15.6.3 by maintenance of the integrity of the 
primary to secondary coolant boundary represented by SG tubes. The 
process by which this integrity is maintained is inspection of SG 
tubes at prescribed intervals, and the repair or removal of 
defective tubes from service. Inspection intervals are based on 
preventing corrosion growth from exceeding tube structural limits, 
thereby preventing tube failure. The 1998 SG inspection 
characterized existing tube degradation, and degraded tubes were 
removed from service at that time. Degradation growth rates were 
evaluated for the next operating interval and it was determined that 
the steam generator tube structural integrity is maintained. 
Degradation of SG tubes was prevented during the extended outage by 
a corrosion prevention program.
    The surveillance extension does not involve a change to plant 
equipment and does not affect the performance of plant equipment 
used to mitigate an accident. This change, therefore, does 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?
    Extending the surveillance interval for the performance of 
specific inspections will not create the possibility of any new or 
different kind of accidents. No change is required to any system 
configurations, plant equipment or analyses.
    SG tube inspections determine tube integrity and provide 
reasonable assurance that a tube rupture or primary to secondary 
leak will not occur. The only type of accident that can be 
postulated from extending the SG inspection interval would be a tube 
leak or rupture and these are analyzed in the UFSAR. No new failure 
modes are created by the surveillance extension. Therefore, this 
change will 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?
    Surveillance interval extensions will not impact any plant 
safety analyses since the assumptions used will remain unchanged. 
The safety limits assumed in the accident analyses and the design 
function of the equipment required to mitigate the consequences of 
any postulated accidents will not be changed since only the 
surveillance interval is being extended. Extending the surveillance 
interval for the performance of these specific inspections does not 
involve a significant reduction in the margin of safety derived from 
the required surveillances.
    The margin of safety depends upon maintenance of specific 
operating parameters within design limits. In the case of SGs, that 
margin is maintained through assurance of tube integrity as the 
primary to secondary boundary. Assurance of tube integrity is 
provided through periodic in-service inspection of tubes and repair 
or removal of defective tubes from service. Radiation monitors 
provide a detection capability of primary to secondary leakage to 
enable a prompt response. The water chemistry of the steam 
generators during shutdown was maintained as described previously in 
Section C [Section C of Attachment B to the licensee's November 29, 
1999, amendment request]. Maintenance of the SG water chemistry 
during power operation in accordance with Electric Power Research 
Institute (EPRI) guidelines provides additional margin of safety. 
Therefore, the plant will be maintained within the analyzed limits 
and the proposed extension will not significantly reduce 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: Mary E. O'Reilly, Attorney, FirstEnergy 
Nuclear Operating Company, FirstEnergy Corporation, 76 South Main 
Street, Akron, OH 44308.
    NRC Section Chief: Marsha Gamberoni.

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

    Date of amendment request: January 19, 2000.
    Description of amendment request: These proposed license amendments 
will revise the Technical Specifications to be consistent with the 
Standard Technical Specifications requirements that allow for an 
expanded as-found testing acceptance tolerance for the main steam 
safety valves (MSSV) and pressurizer code safety valves (PSV). Mode 5 
operability requirements for the PSVs will also be deleted.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The probability of occurrence of an accident previously 
evaluated has not been increased. The changes provided in this 
safety evaluation do not affect the assumptions or results of any 
accident evaluated in the UFSAR [updated final safety analysis 
report]. The actual setpoints and as-left setpoint tolerances of the 
MSSVs and PSVs are not changed as a result of this evaluation.
    Likewise, the consequences of any accident previously evaluated 
have not been increased. The ability of the MSSVs and PSVs to 
respond to accident conditions as assumed in any accident analysis 
has not been affected (i.e., adequate overpressure protection is 
provided). The proposed changes allow for the acceptance of safety 
valve lift test results based on tolerances that are consistent with 
accident analysis assumptions.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed activity does not create the possibility of an 
accident of a different type than any previously evaluated. No 
physical plant changes are being made and no new failure modes have 
been introduced by the proposed changes. This evaluation revises the 
acceptance criteria for MSSV and PSV lift

[[Page 17916]]

test results based on tolerances that are consistent with accident 
analysis assumptions. The actual setpoints and as-left setpoint 
tolerances of the MSSVs and PSVs are not changed as a result of this 
evaluation.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The margin of safety as defined in the basis for any Technical 
Specification or in any licensing document has not been reduced. 
MSSV and PSV setpoint values are not being changed. MSSV and PSV 
setpoints are still required to be set within a tolerance of plus or 
minus 1% (the as-left setpoint tolerance). This evaluation allows 
for the revision of acceptance criteria for MSSV and PSV lift test 
results such that testing criteria is consistent with accident 
analysis assumptions. This will allow for the accommodation of 
setpoint drift without invalidating the accident analyses. The 
proposed changes are consistent with the Standard Technical 
Specifications, which require MSSV and PSV setting within a plus or 
minus 1% tolerance, but allow surveillance testing to accept valves 
that lift within plus or minus 3%. A review of the plants' accident 
analyses has identified the plant-specific tolerances that may be 
used for this surveillance testing.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: February 16, 2000.
    Description of amendment request: These proposed license amendments 
will revise the Technical Specifications (TS) to delete references to 
certain motor operated valve thermal overload protection bypass devices 
for Unit 2 and to revise the TS for accident monitoring instrumentation 
for both Units 1 and 2. The proposed amendments also make an 
administrative change to the Unit 2 TS Index.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The addition of the new ACTION statements for the Unit 1 
accident monitoring instrumentation adds conservatism that does not 
exist in the current Technical Specifications. These changes are 
consistent with either FPL's originally proposed license amendment 
for this instrumentation or consistent with the Technical 
Specification allowed outage time for the component being monitored 
(i.e., the auxiliary feedwater pumps). Unit 2 valves MV-21-4A and 
MV-21-4B were modified to be manually operated valves and no longer 
perform an accident mitigation function. Unit 2 wide range 
Thot instrumentation is used to satisfy Regulatory Guide 
1.97 accident monitoring requirements.
    These Technical Specification changes either correct existing 
errors or add conservatism to the way the Unit is operated. Based on 
the above, the physical changes to plant equipment or plant 
operation would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    Accident monitoring instrumentation monitors the process of 
postulated events, and is not an accident initiator. Unit 2 valves 
MV-21-4A and MV-21-4B were modified to be manually operated valves 
and no longer have an active safety function, therefore, these 
valves are not accident initiators. These Technical Specification 
changes either correct existing errors or add conservatism to the 
way the Unit is operated. Based on the above, the physical changes 
to plant equipment or plant operation would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed amendments do not involve a significant reduction 
in a margin of safety. FPL determined that these proposed license 
amendments are necessary to correct existing errors or add 
conservatism to the way the Unit is operated. As such, the 
assumptions and conclusions of the accident analyses in the UFSAR 
[Updated Final Safety Analysis Report] remain valid and the 
associated safety limits will continue to be met.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: February 18, 2000.
    Description of amendment request: The technical specification (TS) 
changes are being proposed to provide flexibility of operation. These 
changes include: (1) The ability to have a standby Safety Injection 
(SI) pump available during Reactor Coolant System (RCS) reduced 
inventory conditions with the RCS pressure boundary intact; (2) The 
ability to respond more rapidly with additional makeup sources than 
currently established by TSs in the unlikely event of a loss of decay 
heat removal capability or unexpected reduction in RCS inventory; (3) 
Realigning a footnote to clarify the allowance of an inoperable SI pump 
to be energized for testing or filling accumulators; (4) Recognition 
that a substantial vent area exists for cold overpressure protection 
when the reactor vessel head is on and the studs are fully detensioned; 
(5) Limit maneuvering the plant beyond Hot Shutdown when one charging 
pump is operable; and (6) Establishment of a new value for the open 
permissive interlock associated with the Residual Heat Removal System 
suction isolation valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not affect plant systems such that their 
function in the control of radiological consequences is adversely 
affected. The proposed changes do not adversely affect accident 
initiators or precursors nor alter the design assumptions, conditions, 
or manner in which structures, systems, and components perform their 
intended safety function to mitigate the consequences of an initiating 
event within the acceptance limits assumed in the Updated Final Safety 
Analysis Report (UFSAR). The proposed changes do not affect the source 
term, containment isolation, or radiological

[[Page 17917]]

release assumptions used in evaluating the radiological consequences of 
an accident previously evaluated. Since there are no changes to 
previous accident analysis, the radiological consequences associated 
with these analyses remain unchanged; therefore, the proposed changes 
do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not result in a change to the design basis 
of any plant structure, system, or component. All equipment important 
to safety will operate as designed. The proposed TS changes in 
conjunction with administrative controls will provide adequate control 
measures to ensure component integrity is not challenged. The proposed 
changes do not cause the initiation of any accident nor create any new 
failure mechanisms. The changes do not result in any event previously 
deemed incredible being made credible. Therefore, the proposed changes 
do not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes do not adversely affect equipment design or 
operation and there are no changes being made to the TS-required safety 
limits or safety system settings that would adversely affect plant 
safety. The proposed TS changes in conjunction with administrative 
controls will provide adequate control measures to ensure component 
integrity is not challenged. Therefore, the proposed changes do not 
involve a significant reduction in a margin of safety.
    Based on this review, it appears that the three standards of 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: February 18, 2000.
    Description of amendment request: Changes to technical 
specification (TS) Sections 4.0.5 and 4.4.6.2.2.e are being proposed to 
clarify that the Inservice Testing (IST) program will be performed in 
accordance with the requirements of surveillance requirement (SR) 4.0.5 
and the American Society of Mechanical Engineers (ASME) Code for 
Operation and Maintenance of Nuclear Power Plants (ASME OM Code), 
instead of Section XI of the ASME Boiler and Pressure Vessel Code.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Seabrook Station has proposed to utilize the ASME OM Code-1995 
including the 1996 Addenda (OMa Code-1996) for the IST of pumps and 
valves as an alternative to the requirements of the 1989 Edition of 
Section XI pursuant to 10 CFR 50.55a(f)(4)(iv) subject to the 
limitations modifications listed in paragraph (b). The use of the ASME 
OM Code-1995 including the 1996 Addenda has been evaluated by the NRC 
(64 FR 51370) and has supplanted Section XI of the 1989 Edition of the 
ASME Boiler and Pressure Vessel Code as the Code referenced in 
paragraph (b) for the IST of pumps and valves effective November 22, 
1999. The proposed administrative changes only add ASME OM and 
applicable terms from that Code into the TSs. These proposed changes 
are administrative in nature and do not adversely affect accident 
initiators or precursors nor alter the design assumptions, conditions, 
or configuration of the facility. Therefore, the proposed changes do 
not involve a significant increase in the probability or consequences 
of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The changes to the TSs clarify that the IST program will be 
performed in accordance with the requirements of SR 4.0.5 and the ASME 
OM Code and to clarify the surveillance interval requirements for 
components tested on a Semi-quarterly and Biennial frequency. The 
proposed changes are administrative in nature and do not adversely 
affect accident initiators or precursors nor alter the design 
assumptions, conditions, or configuration of the facility. Therefore, 
the proposed change will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The changes to the TSs do not involve a reduction in the margin of 
safety. As previously identified the subject changes are administrative 
in nature and will clarify that the IST program will be performed in 
accordance with the requirements of SR 4.0.5 and the ASME OM Code. The 
use of the ASME OM Code-1995 including the 1996 Addenda in lieu of 
Section XI of the ASME Boiler and Pressure Vessel Code will result in a 
net improvement in the measures for performing the IST of pumps and 
valves and has been previously evaluated by the NRC. Therefore, the 
proposed changes to the TSs will not result in a significant reduction 
in a margin of safety. Based on this review, it appears that the three 
standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes 
to determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: February 29, 2000:
    Description of amendment request: The proposed amendment would 
approve continued use of two exceptions previously granted by the 
Nuclear Regulatory Commission (NRC) to the American Society of 
Mechanical Engineers N510-1989 testing requirements for the emergency 
filtration train (EFT) system, revise the Technical Specifications 
(TSs) to reflect modifications to the EFT system that eliminate the 
need for additional test exceptions, revise the TSs to be consistent 
with the guidance of NRC Generic Letter 99-02, and revise the TSs to 
include operability requirements for the EFT system during operations 
that could result in a fuel handling accident.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

[[Page 17918]]

    During an accident, the Control Room Emergency Filtration [EFT] 
System provides filtered air to pressurize the Control Room to 
minimize the activity, and therefore the radiological dose, inside 
the Control Room. The SBGT [standby gas treatment] System maintains 
a small negative pressure in the Reactor Building to minimize ground 
level escape of airborne radioactivity. Technical Specification 
operability and surveillance requirements are established in order 
to ensure that the SBGT and EFT Systems will perform their safety 
functions during an accident. The proposed amendment documents the 
test method for laboratory testing of charcoal adsorbers in both 
systems, implements adequate test acceptance criteria, and improves 
the methodology of in-place testing of charcoal filters in the EFT 
System. The additional operability requirements for the EFT System 
ensure that the systems will be available when required. The 
surveillances adequately show that the system is operable and 
capable of performing its safety function. Dose to the public and 
the Control Room operators are not affected by the proposed change.
    Since neither system is an accident initiator, the probability 
of an accident is not increased.
    The proposed Technical Specification change does not introduce 
new equipment operating modes, nor does the proposed change alter 
existing system relationships. The proposed amendment does not 
introduce new failure modes.
    Therefore, the proposed amendment will not significantly 
increase the probability or the consequences of an accident 
previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed Technical Specification change does not introduce 
new equipment operating modes, nor does the proposed change alter 
existing system relationships. The proposed amendment does not 
introduce new failure modes. The proposed surveillance requirements 
are consistent with industry and regulatory guidance and show that 
the system is capable of performing its safety function. The added 
operability requirements for the EFT System ensure that the system 
will be available when required.
    Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The proposed amendment is consistent with current industry and 
regulatory standards for testing filters. The proposed amendment 
maintains margins of safety. Off-site and Control Room dose 
assessments are not affected by the proposed amendment, since the 
ability of the SBGT and EFT Systems to perform their safety function 
is shown by the proposed surveillance requirements. The proposed 
change to the surveillances provides assurance that the system will 
perform at the filter efficiency used in the evaluation of the 
radiological consequences of the postulated events. Therefore, the 
proposed amendment will 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: Jay E. Silberg, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

PP&L, Inc., Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric 
Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: January 13, 2000.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TSs) for both units to clarify Figure 3.4.10-
1, ``Reactor Vessel Pressure vs. Minimum Vessel Temperature.'' The 
amendment would also revise the Unit 2 TS to correct a reference in TS 
5.6.5.b, ``Core Operating Limits Report (COLR).''
    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.
    This proposal does not involve an increase in the probability or 
consequences of an accident previously evaluated. The proposed 
revision to Technical Specification Figure 3.4.10.1 and the proposed 
revision to the references in the Unit 2 Technical Specification 
section 5.6.5.b are administrative and/or editorial in nature, and 
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.
    This proposal does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed revision to Technical Specification Figure 3.4.10.1 and 
the proposed revision to the references in the Unit 2 Technical 
Specification section 5.6.5.b are administrative and/or editorial in 
nature. The proposed revisions do not change any plant systems, 
structures, or components, nor do they change any existing accident 
analysis, or create any 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.
    This proposal does not involve a significant reduction in the 
margin of safety. The proposed revision to Technical Specification 
Figure 3.4.10.1 and the proposed revision to the references in the 
Unit 2 Technical Specification section 5.6.5.b are administrative 
and/or editorial in nature, and do not result in [a] significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PP&L, Inc., 2 North Ninth St., GENTW3, Allentown, PA 18101-
1179.
    NRC Section Chief: Marsha Gamberoni, Acting.

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

    Date of amendment request: March 8, 2000.
    Description of amendment request: The proposed amendment would 
revise the R. E. Ginna Nuclear Power Plant Improved Technical 
Specifications associated with the Spent Fuel Pool Storage (SFP) 
(limiting condition for operation (LCO) 3.7.13), and Design Features 
Fuel Storage (4.3).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of Ginna [Nuclear Power Plant] in accordance with 
the proposed changes does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. The 
administrative change only involves how the maximum initial fuel 
assembly enrichment is described and has no impact on the 
probability or consequences of an accident. The remaining change is 
evaluated below.
    The regions of the SFP and specific storage cell types differ 
from each other in regards to the specific absorber material within 
the cells. Administrative controls are used to maintain the 
specified storage patterns and to assure storage of a fuel assembly 
in a proper location based on initial U-235 enrichment, burnup, and 
decay time. Procedures which perform this surveillance will include 
independent verification provisions.
    There is no significant increase in the probability of an 
accident concerning the potential insertion of a fuel assembly in an

[[Page 17919]]

incorrect location in the storage racks. Ginna currently uses 
administrative controls to move fuel assemblies from location to 
location within the SFP. Fuel assembly placement will continue to be 
controlled pursuant to approved fuel handling procedures and will be 
in accordance with the Improved Technical Specification spent fuel 
rack storage configuration limitations. Fuel movement procedures are 
planned to include independent verification of fuel handling steps.
    There is no increase in the consequences of the accidental 
misloading of spent fuel assemblies into the spent fuel pool racks. 
The criticality safety analysis demonstrate that the pool 
Keff will remain 0.95 following an accidental 
misloading due to the boron concentration of the pool. The existing 
Improved Technical Specification limitation on soluble boron within 
the SFP will ensure that an adequate boron concentration is 
maintained.
    Based on the above, it is concluded that the proposed changes do 
not significantly increase the probability or consequences of any 
accident previously analyzed.
    2. Operation of Ginna [Nuclear Power Plant] in accordance with 
the proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The administrative change to the Improved Technical Specifications 
has no impact on plant hardware or operations and therefore cannot 
create a new or different kind of an accident.
    Criticality accidents in the SFP are not new or different types 
of accidents, they have been analyzed in the Updated Final Safety 
Analysis Report and in criticality safety analysis reports 
associated with specific licensing amendments for fuel enrichments 
up to the nominal 5.0 weight percent U-235 that is assumed for the 
proposed change.
    The current Improved Technical Specifications contain 
limitations on the minimum SFP boron concentration. The proposed 
changes to the Improved Technical Specifications to allow credit for 
soluble boron for a Keff  0.95 in the SFP is consistent 
with the results of the new criticality safety analysis. Since 
soluble boron has always been maintained in the SFP water, and is 
currently required by Improved Technical Specifications, the 
implementation of this new requirement will have no effect on normal 
SFP operations and maintenance. A dilution of the spent fuel pool 
soluble boron has always been a possibility, however, it has been 
shown in the SFP boron dilution analysis that there are no credible 
dilution events for which the spent fuel pool Keff could 
increase to >0.95. Therefore, the implementation of crediting 
soluble boron in the SFP will not result in the possibility of a new 
kind of accident.
    The proposed changes to Improved Technical Specifications LCO 
3.7.13 continue to specify the requirements for the spent fuel rack 
storage configurations. Since the proposed SFP storage configuration 
limitations will be similar to the current ones, the new limitations 
will not have any significant effect on normal spent fuel pool 
operations and maintenance and will not create any possibility of a 
new or different kind of accident. Verifications will be performed 
to ensure that the spent fuel pool loading configuration meets 
specified requirements.
    The misloading of a fuel assembly in the required storage 
configuration has been evaluated. In all cases, the rack 
Keff remains 0.95.
    Under the proposed amendment, no changes are being made to the 
racks themselves, any other systems, or to the physical structures 
of the Auxiliary Building itself. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Operation of Ginna [Nuclear Power Plant] in accordance with 
the proposed changes does not involve a significant reduction in a 
margin of safety. The proposed administrative change to the Improved 
Technical Specifications has no impact on any acceptance criteria, 
plant operations or the actual failure of any systems, components or 
structure; therefore the change has no impact on the margin of 
safety.
    The spent fuel storage operation limits will provide adequate 
safety margin to ensure that the stored fuel assembly array will 
always remain subcritical. Those limits are based on a plant 
specific criticality safety analysis performed in a manner analogous 
to that of the NRC approved Westinghouse spent fuel rack criticality 
safety analysis methodology.
    While the criticality safety analysis utilized credit for 
soluble boron, storage configurations have been defined using 95/95 
Keff calculations to ensure that the spent fuel rack 
Keff will be 1.0 with no soluble boron. Soluble boron 
credit is used to offset uncertainties, tolerances, and off-normal 
conditions (such as a misplaced assembly) and to provide subcritical 
margin such that the spent fuel pool Keff is maintained 
at 0.95.
    The loss of substantial amounts of soluble boron from the spent 
fuel pool which could lead to Keff exceeding 0.95 has 
been evaluated and shown to be not credible. An evaluation has been 
performed which shows that dilution of the SFP boron concentration 
from 2300 ppm to 975 ppm is not credible. Also, the spent fuel rack 
Keff will remain 1.0 (with a 95/95 confidence level) with 
the SFP flooded with unborated water. These analyses demonstrate a 
level of safety comparable to the conservative criticality safety 
analysis methodology required by Westinghouse WCAP-14416. 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: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005.
    NRC Section Chief: Marsha Gamberoni, Acting.

Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry 
Nuclear Plant, Units 2 and 3, Limestone County, Alabama

    Date of amendment request: March 15, 2000.
    Description of amendment request: The proposed amendment would 
change to the technical specifications, to provide a completion time of 
7 days of continued reactor operation with two CAD subsystems 
inoperable.
    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:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The safety-related function of the Containment Atmosphere 
Dilution (CAD) system is to mitigate the effects of a loss-of-
coolant-accident (LOCA) by limiting the volumetric concentration of 
oxygen in the primary containment atmosphere. The CAD System is not 
an event initiator, therefore, the probability of the occurrence of 
an accident is not affected by this proposed Technical Specification 
(TS) change. Emergency procedures preferentially use the normal 
containment inerting system to provide post-accident vent and purge 
capability, with the CAD system only serving in a backup role to 
this system. Hence, in the event of the inoperability of both CAD 
subsystems, the proposed TS require the normal containment inerting 
system to be verified available as an alternate oxygen control 
means.
    Therefore, the proposed TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    This TS change does not result in any changes to the CAD 
equipment design or capabilities or to the operation of the plant. 
Since the change impacts only the required action completion time 
for periods of CAD subsystem inoperability and does not result in 
any change in the response of the equipment to an accident, the 
change does not create the possibility of a new or different kind of 
accident from any previously analyzed.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    As stated in GL 84-09, a Mark I type boiling water reactor (BWR) 
plant is not considered to rely upon purge/repressurization systems 
such as CAD as its primary means of hydrogen control when the 
unit(s) is operated in accordance with certain technical criteria. 
The BFN units are operated in accordance with these criteria. The 
BFN Unit 2 and Unit 3 containments are inerted

[[Page 17920]]

with nitrogen during normal operation, recycled containment 
atmosphere is used for pneumatically operated components inside 
containment, and there are no potential sources of oxygen generation 
inside containment other than the radiolytic decomposition of water. 
The system preferred by the EOIs for oxygen control post-accident is 
the normal primary containment inerting system. Because the 
probability of an accident involving hydrogen and oxygen production 
is small, CAD is not the primary system used to mitigate the 
creation of combustible containment atmosphere mixtures, and because 
the requested LCO where both CAD subsystems is inoperable is not 
long, no significant reduction in the margin of safety is associated 
with this proposed amendment.

    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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard Correia.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: February 18, 2000.
    Description of amendment request: The proposed amendment would 
revise the technical specifications to identify (1) M5 alloy as a 
material used in the construction of fuel assemblies, and (2) The 
associated topical report that describes the fuel.
    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:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed TS revision will allow the use of a new advance 
alloy material for the fuel rod cladding. The new M5 alloy 
properties are not significantly different than the characteristics 
of the currently used zircaloy-4 as demonstrated in the NRC approved 
Topical Report BAW-10227P-A for the use of the M5 alloy for fuel rod 
cladding. In this topical, the M5 alloy was shown to perform very 
similar to the zircaloy-4 with improved performance in several areas 
including fuel cladding corrosion, hydrogen pickup, fuel rod and 
fuel assembly growth, and fuel rod cladding creep. The proposed 
revision will not alter the operating characteristics of the plant 
or plant components. The fuel rod cladding function will not be 
changed even though some of the rod cladding properties could be 
enhanced.
    The M5 alloy will maintain fuel rod cladding integrity such that 
the potential for rod cladding failures is not increased. The fuel 
rod cladding is not assumed to arbitrarily fail as an accident 
initiator even though it does function to ensure that initial core 
conditions are within the analysis assumptions and to provide a 
barrier to the release of radiation. Therefore, the proposed 
revision will not increase the possibility of an accident based on 
the new M5 alloy having similar properties as the zircaloy-4 
material.
    The ability of the new M5 fuel rod cladding material to provide 
a barrier against the release of radioactive fuel material has not 
been reduced with respect to the zircaloy-4 material and the 
generation of hydrogen has been reduced. The approved topical report 
evaluated postulated accidents that involved adverse core conditions 
and the release of radionuclides and found the M5 alloy to perform 
similar to the current fuel rod cladding material. Rod cladding 
failures are assumed to occur in the fuel handling accident; 
however, the consequences of this event is independent of the 
properties of the fuel rod cladding. This is based on the fuel 
handling event assuming the rupture of fuel rods regardless of the 
rod cladding material. Therefore, based on the topical report 
results, the proposed revision to allow the use of M5 fuel rod 
cladding material will not significantly increase the consequences 
of an accident and the potential for the release of radioactive 
material to the environment.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed M5 rod cladding material has been demonstrated to 
have properties that are not significantly different than the 
current zircaloy-4 in maintaining the integrity of the fuel rods. 
The new material will not alter the functions of the rod cladding 
which is to provide a barrier against the release of radioactive 
material. Initial plant conditions, which is considered in the 
accident analysis, will also be maintained such that no new plant 
conditions will exist that could affect the analysis results. Since 
plant functions and conditions are not impacted by the proposed 
revision and the new M5 rod cladding is not postulated to become an 
accident initiator based on the similarity with zircaloy-4, the 
possibility of a new or different kind of accident is not created.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The margin of safety is established by the acceptance criteria 
used by NRC. Meeting the acceptance criteria assures that the 
consequences of accidents are within known and acceptable limits. 
The loss-of-coolant accident (LOCA) acceptance criteria are 
unchanged: peak cladding temperature of  2200 degrees 
Fahrenheit; maximum cladding oxidation of 17 percent of 
the total cladding thickness before oxidation; maximum hydrogen 
generation of 1 percent of the hypothetical amount if all 
of the cladding metal were to react; coolable geometry such that the 
core remains amenable to cooling; and long-term cooling to maintain 
core temperature at an acceptably low value and removal of decay 
heat for an extended period.
    These requirements continue to be met with the new M5 fuel rod 
cladding material. The acceptance criteria for Departure from 
Nucleate Boiling (DNB) events has not changed and is still the 95 
percent probability and 95 percent confidence interval that DNB is 
not occurring during the transient. The changes to material 
properties have been evaluated in BAW-10227P-A and all applicable 
acceptance criteria are met. In addition, the proposed revision to 
allow the use of M5 fuel rod cladding will not impact plant 
setpoints that maintain the margin of safety. Based on these 
results, it is concluded that the margin of safety is not 
significantly reduced.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

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

    Date of application request: March 6, 2000 (ULNRC-04197).
    Description of amendment request: The proposed amendment will 
revise Table 3.7.1-1, ``Operable Main Steam Safety Valves [MSSVs] 
versus Maximum Allowable Power,'' of the technical specifications to 
reduce the maximum allowable reactor power for a given number of 
operable MSSVs per steam generator. There are five MSSVs on each of the 
four steam generators for the plant. This change will increase 
restrictions on the operation of the plant to account for (1) 
Westinghouse letter, SCP-99-129, dated July 7, 1999, and (2) 
Westinghouse Nuclear Safety Advisory Letter, NSAL-94-001, dated January 
20, 1994. This change will decrease the setpoint values for the power 
range neutron flux high channels, which are part of the reactor trip 
system (RTS) instrumentation in Table 3.3.1-1, ``Reactor Trip System 
Instrumentation,'' of the TSs, and will result in the reactor being 
shut down at a lower reactor power for a given number of operable MSSVs 
per steam generator. There is also a change to the Required Action

[[Page 17921]]

A.1 for Limiting Condition for Operation (LCO) 3.7.1, ``Main Steam 
Safety Valves (MSSVs).'' The licensee has administrative controls in 
place to ensure that the proposed reduced maximum allowable reactor 
power values are in effect at the plant.
    In addition to the changes to LCO 3.7.1 above, the licensee also 
proposed to correct two format errors in the actions for LCO 3.7.1. The 
first correction is to add a separating line between Conditions A and 
B; the second correction is to move the word ``(continued)'' above the 
bottom line for Condition B. Neither of these corrections have any 
affect on the requirements stated in LCO 3.7.1. The licensee also 
showed the changes to the Bases of LCO 3.7.1 that are related to the 
proposed amendment including two editorial corrections to the Bases.
    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 operability of the MSSVs ensures that the secondary side 
system pressure is limited to within 110% of its design pressure 
during the most severe anticipated system operational transient, 
which is the Loss of Load/Turbine Trip Event. As stated in FSAR 
[Callaway final safety analysis report] 15.2.3.3, these events do 
not present a hazard to the integrity of the reactor core, the 
reactor coolant system, or the main steam system. The Power Range 
Neutron Flux High Reactor Trip function and the MSSVs are designed 
to mitigate the consequences of the Loss of Load/Turbine Trip event. 
The Loss of Load event is initiated as a result of an electrical 
system disturbance and the Turbine trip event is initiated as a 
result of a signal derived from the turbine emergency trip fluid 
pressure transmitters and turbine stop valve limit switches.
    The Power Range Neutron Flux High Reactor Trip function and the 
MSSVs ensure that the FSAR Loss of Load/Turbine Trip analyses are 
bounding for cases when not all of the MSSVs are operable. Technical 
Specification Table 3.7.1-1 controls the Power Range Neutron Flux 
High Setpoints when a MSSV is found to be inoperable. The controls 
under this proposed change, which are more restrictive than the ones 
in Technical Specification Table 3.7.1-1, do not install or modify 
any plant equipment. The revised Power Range Neutron Flux High 
Setpoints with inoperable MSSVs proposed under this change are 
bounded by the reactor trip setpoints currently provided in Table 
3.7.1-1. In addition the functionality of plant equipment is 
unaffected by the proposed change.
    Therefore, these 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 ensure that the FSAR Loss of Load/Turbine 
Trip analyses are bounding for cases when not all of the MSSVs are 
operable. Furthermore, the changes do not result in any previously 
incredible accidents becoming credible. No additional equipment is 
being [added to the plant or] credited in the mitigation of any 
[FSAR] Chapter 15 accident events, and the proposed changes do not 
invalidate any previous conclusions.
    Thus, the changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Using the Power Range Neutron Flux High Setpoints with 
inoperable MSSVs provided by Westinghouse (Reference 2 [in the 
licensee's application letter]) in lieu of the ones calculated using 
the equation provided in the Current Technical Specifications Bases, 
results in more conservative reactor trip setpoints. This increases 
the margin of safety. The margin of safety as determined in the 
basis for the Technical Specification is not reduced.
    Therefore, the 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: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Section Chief: Stephen Dembek.

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

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

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs 
Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

    Date of amendment request: February 18, 2000.
    Description of amendment request: The amendment requests approval 
to Baltimore Gas and Electric Company's (BGE's) operating license that 
the new identified failure mode is acceptable on the basis that BGE 
will assure on every shift that safety-related loads are sufficiently 
available to Diesel Generator 1A to ensure the minimum load is met.
    Date of publication of individual notice in Federal Register: March 
7, 2000 (65 FR 12038).
    Expiration date of individual notice: April 6, 2000.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: July 26, 1999, as supplemented 
January 20, 2000.
    Brief description of amendment: The proposed amendment would revise 
Technical Specifications associated with the degraded voltage trip and 
the under-frequency reactor trip surveillance tests.
    Date of publication of individual notice in Federal Register: 
February 28, 2000 (65 FR 10565).
    Expiration date of individual notice: March 29, 2000.

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

    Date of application of amendments: January 27, 2000.
    Brief description of amendments: The amendments revised the 
Facility Operating Licenses by (a) deleting the license conditions that 
have been fulfilled by actions that have been completed, (b) changing 
the license conditions that have been superseded by the current plant 
status, and (c) incorporating other administrative changes.
    Date of publication of individual notice in Federal Register: 
February 8, 2000 (64 FR 6243).
    Expiration date of individual notice: March 9, 2000.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the

[[Page 17922]]

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, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

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

    Date of application for amendment: December 16, 1999.
    Brief description of amendment: The amendment allowed a one-time 
extension of some Technical Specification surveillance intervals to 
support elimination of a planned spring 2000 midcycle outage. The 
surveillances would be extended to no later than November 30, 2000.
    Date of issuance: March 17, 2000.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment No.: 125.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 12, 2000 (65 FR 
1921).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 17, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 25, 1999 (U-603282).
    Brief description of amendment: The amendment revised the Technical 
Specification allowable values for the reactor protection system 
electric power monitoring assembly overvoltage and undervoltage trip 
setpoints.
    Date of issuance: March 21, 2000.
    Effective date: Immediately upon date of issuance and shall be 
implemented within 30 days.
    Amendment No.: 126.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 12, 2000 (65 FR 
1919).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 21, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 13, 1999.
    Brief description of amendment: The amendment revised Sections 
2.a., 2.c.(3) and 2.c.(7) of the Facility Operating License to delete 
already completed license conditions or update out-of-date reporting 
references, and made a change to the Bases of Technical Specification 
3.1.1 regarding the pressurizer safety valves lift setpoint.
    Date of issuance: March 14, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 222.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35206).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 14, 2000.
    No significant hazards consideration comments received: No.

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs 
Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

    Date of application for amendment: November 18, 1999.
    Brief description of amendment: The amendment incorporated a change 
in the pressure-temperature curves in the Calvert Cliffs Nuclear Power 
Plant, Unit No. 1 Technical Specifications. Baltimore Gas and Electric 
Company changed the fluence level for which the curves are valid from 
2.61  x  10\19\ n/cm\2\ to 4.49  x  10\19\ n/cm\2\.
    Date of issuance: March 20, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 234.
    Facility Operating License No. DPR-53: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70078).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 20, 2000.
    No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket No. 50-374, LaSalle County Station, 
Unit 2, LaSalle County, Illinois

    Date of application for amendment: February 21, 2000.
    Brief description of amendment: The amendment changed Technical 
Specification Surveillance Requirement 4.0.5.f to allow the required 
examination of weld RH-2005-29 to be deferred until the next scheduled 
refueling outage or December 31, 2000, whichever is earlier.
    Date of issuance: March 22, 2000.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment No.: 123.
    Facility Operating License No. NPF-18: The amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (65 FR 11809 dated March 6, 2000). The notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by April 5, 2000, but indicated that if the Commission makes a 
final no significant hazards consideration determination, any such 
hearing would take place after issuance of the amendment. The 
Commission's related evaluation of the amendment,

[[Page 17923]]

finding of exigent circumstances, and final no significant hazards 
consideration determination are contained in a Safety Evaluation dated 
March 22, 2000.

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

    Date of application for amendment: December 17, 1999, as 
supplemented January 26, 2000.
    Brief description of amendment: The amendment revises Technical 
Specification Surveillance Requirement 3.6.1.3.9 to allow a 
representative sample of reactor instrumentation line excess flow check 
valves to be tested every 18 months, instead of testing each excess 
flow check valve every 18 months.
    Date of issuance: March 14, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 137.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4270) The January 26, 2000, letter 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 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 14, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 3, 1999, as 
supplemented by letter dated January 14, 2000.
    Brief description of amendments: The amendments revised the 
Technical Specifications Surveillance Requirements (SR) 3.8.1.13 and SR 
3.8.1.14 for emergency diesel generators at Catawba Nuclear Station. 
Specifically, these SR may now be performed at any operational power 
level for Catawba Nuclear Station. In addition, in November 3, 1999, 
application, licensee requested that the power factor requirements be 
deleted from SR 3.8.1.9, and 3.8.1.14. However, licensee withdrew the 
power factor deletion part of the request for Catawba Nuclear Station, 
Units 1 and 2, in a letter dated January 14, 2000.
    Date of issuance: March 16, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1-185; Unit 2-177.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 1, 1999 (64 FR 
67332).
    The January 14, 2000, letter provided additional clarifications 
that did not enlarge the scope of the previous no significant hazard 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 16, 2000.
    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: November 3, 1999, as 
supplemented by letters dated January 14 and February 17, 2000.
    Brief description of amendments: The amendments revise the 
following Technical Specifications Surveillance Requirements (SR): (1) 
SR 3.8.1.9 to allow performance of the diesel generator (DG) load 
rejection test at any operational power level and to delete the power 
factor requirements, (2) SR3.8.1.10 to allow performance of the DG full 
load rejection test at any power level, and (3) SR 3.8.1.14 to allow 
performance of the 24-hr DG run at any operational power level and 
delete the power factor requirement. No plant modification is involved.
    Date of issuance: March 15, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1-192; Unit 2-173.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 1, 1999 (64 FR 
67333).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 15, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application of amendments: January 27, 2000.
    Brief description of amendments: The amendments revised the 
Facility Operating Licenses by (a) Deleting the license conditions that 
have been fulfilled by actions that have been completed, (b) Changing 
the license conditions that have been superseded by the current plant 
status, and (c) Incorporating other administrative changes.
    Date of Issuance: March 13, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1-311; Unit 2-311; Unit 3-311.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Facility Operating Licenses and License 
Conditions.
    Date of initial notice in Federal Register: February 8, 2000 (65 FR 
6243).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 13, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 29, 1999.
    Brief description of amendment: This amendment authorizes a 
revision to the post loss-of-coolant accident (LOCA) dose calculations 
described in the River Bend Station (RBS) Updated Safety Analysis 
Report (USAR). The analyses are being updated to account for several 
changes that were determined by the licensee to involve an unreviewed 
safety question in accordance with title 10 of the Code of Federal 
Regulations, section 50.59(a)(2)(i). Specifically, the licensee 
requested the following changes to the RBS USAR, Sections 6.2.3 and 
15.6.5:
    Increase of the positive pressure period of the secondary 
containment following a design basis accident to 195.5 seconds from 189 
seconds.
    Decrease of the suppression pool water volume to 1.2E5 ft\3\ from 
1.35E5 ft\3\ for use in the post-LOCA dose calculation.
    Change to the engineered safety feature (ESF) liquid leakage model 
adding the leakage resulting from a gross failure of a passive 
component outside of primary containment.
    Direct release of ESF leakage through the Standby Gas Treatment 
System to the environment without hold up in the auxiliary building.
    Date of issuance: March 17, 2000.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.

[[Page 17924]]

    Amendment No.: 111.
    Facility Operating License No. NPF-47: The amendment revised the 
USAR.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70084).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 17, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 18, 1999.
    Brief description of amendment: This amendment removes license 
condition 3.H, ``Long Term Program,'' from Facility Operating License 
No. DPR-35 for the Pilgrim Nuclear Power Station.
    Date of issuance: March 13, 2000.
    Effective date: As of its date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 183.
    Facility Operating License No. DPR-35: Amendment revised the 
License.
    Date of initial notice in Federal Register: February 9, 2000 (65 FR 
6404).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 13, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 5, 1999, as supplemented 
January 31, 2000.
    Brief description of amendment: This amendment modifies the 
licensing basis for the on-site fuel storage requirements for the 
emergency diesel generators. Various sections of the technical 
specifications were amended to reflect the new licensing basis.
    Date of issuance: March 17, 2000.
    Effective date: As of its date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 184.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 2, 1999 (64 FR 
29708).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 17, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 3, 1998, as 
supplemented by letter dated October 7, 1999.
    Brief description of amendment: This amendment authorizes revision 
of the Grand Gulf Nuclear Station Updated Final Safety Analysis Report 
for implementation of a limited scope application of the alternative 
accident source term described in NUREG-1465. The amendment allows a 
change in the minimum time assumed for the onset of fission product 
release from perforated fuel rods following a postulated design basis 
loss-of-coolant accident.
    Date of issuance: March 22, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No: 143.
    Facility Operating License No. NPF-29: The amendment changes the 
Grand Gulf Nuclear Station design basis by revising the Updated Final 
Safety Analysis Report.
    Date of initial notice in Federal Register: December 1, 1999 (64 FR 
67333).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 22, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 2, 1999.
    Brief description of amendment: This amendment revises the 
Technical Specifications (TSs) to (1) Relocate the requirements of TS 
3/4.1.2.8, Reactivity Control Systems--Borated Water Sources--Shutdown, 
in its entirety, to the DBNPS Updated Safety Analysis Report (USAR) 
Technical Requirements Manual (TRM); (2) Relocate the requirements of 
TS 3/4.1.2.9, Reactivity Control Systems--Borated Water Sources--
Operating, to the USAR TRM, except for portions applicable to the 
Borated Water Storage Tank, which have been deleted because they are 
redundant to the existing provisions of TS 3/4.5.4, Emergency Core 
Cooling Systems--Borated Water Storage Tank; (3) Modify TS 3/4.1.2.1, 
Reactivity Control Systems--Borated Water Sources--Shutdown, by 
deleting references to TS 3.1.2.8; (4) Incorporate corresponding 
changes to the TS index; and (5) Incorporate corresponding changes to 
the TS Bases.
    Date of issuance: March 14, 2000.
    Effective date: Immediately upon date of issuance and shall be 
implemented within 120 days.
    Amendment No.: 238.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70086).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 14, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 8, 1998.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) 5.3.1, ``Design Features--Reactor Core--Fuel 
Assemblies,'' and TS Bases Section 2.1, ``Safety Limits.'' The 
amendment permits the use of the Framatome Cogema Fuels ``M5'' advanced 
alloy for fuel rod cladding and fuel assembly spacer grids.
    Date of issuance: March 15, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 239.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 7, 1998 (63 FR 
53961).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 15, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 23, 1999, as 
supplemented March 9, 2000.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) surveillance testing of the safety-
related ventilation system charcoal to meet the actions requested in 
Generic Letter 99-02, ``Laboratory Testing of Nuclear-Grade Activated 
Charcoal,'' dated June 3, 1999. Other systems impacted include the 
emergency containment filtering system,

[[Page 17925]]

post accident containment vent system, and the control room emergency 
ventilation system.
    Date of issuance: March 21, 2000.
    Effective date: March 21, 2000.
    Amendment Nos.: 205 and 199.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the TS.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70089). The March 9, 2000, submittal provided clarifying information 
that did not change the scope of the original request or change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 21, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 19, 1999, as 
supplemented February 23, 2000.
    Brief description of amendment: Changes the Crystal River Unit 3 
Technical Specifications (TS) to incorporate the requirements of 10 CFR 
50.55a relating to containment inspections.
    Date of issuance: March 16, 2000.
    Effective date: March 16, 2000.
    Amendment No.: 191.
    Facility Operating License No. DPR-72: Amendment revised the TS.
    Date of initial notice in Federal Register: October 20, 1999 (64 FR 
56530). The February 23, 2000, supplement did not affect the original 
no significant hazards consideration determination, or expand the scope 
of the amendment request as originally noticed.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 16, 2000.
    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: December 3, 1998.
    Brief description of amendments: The amendments incorporate the 
Distribution Ignition System requirements into the Unit 1 and Unit 2 
Technical Specifications.
    Date of issuance: March 15, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 242 and 223.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4279).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 15, 2000.
    No significant hazards consideration comments received: No

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

    Date of amendment request: March 31, 1999.
    Brief description of amendment: The change modifies Cooper Nuclear 
Station's Technical Specifications, Section 5.3.1, ``Unit Staff 
Qualifications.'' The change endorses the provisions of Regulatory 
Guide 1.8, Revision 2, ``Qualification and Training of Personnel for 
Nuclear Power Plants,'' for the shift supervisor, senior operator, 
licensed operator, shift technical advisor, and radiological manager.
    Date of issuance: March 15, 2000.
    Effective date: March 15, 2000, to be implemented within 30 days.
    Amendment No.: 181.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24197).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 15, 2000.
    No significant hazards consideration comments received: No.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: November 23, 1999, as 
supplemented December 7, 1999.
    Brief description of amendment: The amendment updates the list of 
documents which describe the analytical methods used to determine the 
core operating limits specified in Technical Specification 6.9.1.8b.
    Date of issuance: March 17, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 242.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4284).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 17, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 27, 1999.
    Brief description of amendments: Revised the technical 
specifications to clarify several administrative requirements, delete 
redundant requirements, and correct typographical errors, and are 
considered administrative in nature.
    Date of issuance: March 14, 2000.
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment Nos.: Unit 1-139; Unit 2-102.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 17, 1999 (64 
FR 62714).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 14, 2000.
    No significant hazards consideration comments received: No.

Power Authority of The State of New York, Docket No. 50-286, Indian 
Point Nuclear Generating Unit No. 3, Westchester County, New York

    Date of application for amendment: September 9, 1996, as 
supplemented June 6, 1997, and June 7, 1999.
    Brief description of amendment: The amendment removes the 
requirement for the Plant Operating Review Committee review of the fire 
protection program and implementing procedures.
    Date of issuance: March 13, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 201.
    Facility Operating License No. DPR-64: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 1, 1999 (64 FR 
67339).
    No significant hazards consideration comments received: No.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 13, 2000.
    No significant hazards consideration comments received: No.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of application for amendments: October 24, 1997, as 
supplemented,

[[Page 17926]]

January 8, September 21, and December 22, 1998; and January 7, February 
17, June 21, and August 23, 1999, and February 7, 2000.
    Brief description of amendments: These amendments revise the Salem 
Technical Specifications (TSs), Section 3/4.7.7, ``Auxiliary Building 
Exhaust Air Ventilation System,'' to require two auxiliary building 
ventilation system (ABVS) supply fans, and three ABVS exhaust fans to 
be operable, and clarify administrative controls.
    Date of issuance: March 21, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 228 and 209.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 17, 1997 (62 
FR 66140).
    The January 8, September 21, and December 22, 1998; and January 7, 
February 17, June 21 and August 23, 1999; and February 7, 2000, letters 
provided clarifying information that did not change the staff's initial 
proposed no significant hazards consideration determination or expand 
the application beyond the scope of the original notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 21, 2000.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of application or amendments: November 30, 1999.
    Brief description of amendments: The amendments revise Technical 
Specifications and associated Bases to Surveillance Requirement 
3.8.1.12 to remove the restriction which prevents performance of the 
diesel generator 24-hour run while operating in either Mode 1 or 2.
    Date of issuance: March 15, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1-218; Unit 2-159.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73098)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 15, 2000.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of application for amendments: October 15, 1999.
    Brief description of amendments: The amendments revise the Safety 
Limit Minimum Critical Power Ratios (SLMCPR) in Technical Specification 
2.1.1.2 to reflect the results of a cycle-specific calculations for 
Unit 1 Cycle 19 and Unit 2 Cycle 16. The calculations were performed 
using the new NRC-approved methodology for determining SLMCPRs.
    Date of issuance: March 22, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1-219; Unit 2-160.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 17, 1999 (64 
FR 62715 and 64 FR 62716).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 22, 2000.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of application for amendments: April 6, 1999.
    Brief description of amendments: The amendments revise the 
Technical Specifications to allow an increase of 168 fuel assemblies in 
the storage capacity of Unit 1's spent fuel pool and an increase of 88 
fuel assemblies in the storage capacity of Unit 2's spent fuel pool.
    Date of issuance: March 23, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1-220; Unit 2-161.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 4, 1999 (64 FR 
23877).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 23, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 12, 1999 (PCN-505).
    Brief description of amendments: The amendments revise Technical 
Specification 5.5.2.13, ``Diesel Fuel Oil Testing Program.''
    Date of issuance: March 20, 2000.
    Effective date: March 20, 2000, to be implemented within 30 days of 
issuance.
    Amendment Nos.: Unit 2--167; Unit 3-158.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: December 1, 1999 (64 FR 
67339).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 20, 2000.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry 
Nuclear Plant, Units 2 and 3, Limestone County, Alabama

    Date of application for amendments: September 28, 1999, as 
supplemented February 4, 2000 (TS-399).
    Brief description of amendments: The Technical Specifications (TS) 
have been changed to increase the allowable leakage for any one of the 
four main steam line (MSL) penetrations from 11\1/2\ to 100 standard 
cubic feet per hour (scfh), and to establish a 150 scfh limit on the 
maximum allowable combined leakage of all four MSL penetrations.
    Date of issuance: March 14, 2000.
    Effective date: March 14, 2000.
    Amendment Nos.: 263 and 223.
    Facility Operating License Nos. DPR-52 and DPR-68: Amendments 
revised the TS.
    Date of initial notice in Federal Register: November 3, 1999 (64 FR 
59807). The supplemental letter dated February 4, 2000, contained 
clarifying information that did not change the initial no significant 
hazards determination.

[[Page 17927]]

    The Commission's related evaluation of the amendment is contained 
in an Environmental Assessment dated February 22, 2000, and a Safety 
Evaluation dated March 14, 2000.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of application for amendment: September 30, 1999, as 
supplemented February 29, 2000.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) analytical methods for core operating limits to 
implement an analysis supporting a more negative moderator temperature 
coefficient for the end-of-cycle, rated thermal power condition.
    Date of issuance: March 14, 2000.
    Effective date: March 14, 2000.
    Amendment No.: 20.
    Facility Operating License No. NPF-90: Amendment revises the TS.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4291). The supplemental letter dated February 29, 2000, contained 
clarifying information and did not change the initial proposed No 
Significant Hazards Consideration Determination or expand the 
application beyond the scope of the original notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 14, 2000.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of application for amendment: June 25, 1999, as supplemented 
January 25, 2000.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TS) to apply the Westinghouse generic best estimate 
large break loss-of-coolant accident analysis methodology, using the 
WCOBRA/TRAC code to the Watts Bar Unit 1 plant.
    Date of issuance: March 17, 2000.
    Effective date: March 17, 2000.
    Amendment No.: 21.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 9, 2000 (65 FR 
611). The January 25, 2000, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination or expand the application beyond the scope 
of the original notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 17, 2000.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of application for amendment: November 15, 1999 (TS 99-16).
    Brief description of amendment: The amendment changes the 
methodology and frequency for sampling the ice condenser ice bed 
(stored ice) and adds a new Technical Specification (TS) and associated 
Bases to change the methodology and frequency for sampling requirements 
for all ice additions to the ice bed.
    Date of issuance: March 21, 2000.
    Effective date: March 21, 2000.
    Amendment No.: 22.
    Facility Operating License No. NPF-90: Amendment revises the TSs.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70092).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 21, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 28, 1999, as 
supplemented December 21, 1999.
    Brief description of amendments: The amendments remove the 
operability and surveillance requirements of Technical Specifications 
(TS) Section 3/4.6.4.3, ``Waste Gas Charcoal Filter System,'' from the 
TS and relocate them to the Technical Requirements Manual.
    Date of issuance: March 13, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 222 and 203.
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: February 9, 2000 (65 FR 
6412).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 13, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 1, 1999.
    Brief description of amendments: These amendments reflect a change 
to Technical Specification Section 15.5.4. The amendments remove one of 
the two separate methods for verifying the acceptability of reactor 
fuel for placement and storage in the spent fuel pool and new fuel 
storage vault.
    Date of issuance: March 20, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 194 and 199.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 28, 1999 (64 FR 
40911).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 20, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 15, 1999.
    Brief description of amendments: This amendment changes the control 
rod surveillance interval in TS Table 15.4.1-2, Item 10, ``Partial 
movement of all rods,'' from once ``Every 2 weeks'' to ``Quarterly.'' 
This change implements the recommendation of NRC Generic Letter 93-05, 
``Line Item Technical Specifications Improvements to Reduce 
Surveillance Requirements for Testing During Power Operation.''
    Date of issuance: March 22, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 195 and 200.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73103).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 22, 2000.
    No significant hazards consideration comments received: No.

[[Page 17928]]

Wisconsin Public Service Corporation,Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: June 22, 1999, as supplemented 
on December 2, 1999, and January 17, 2000.
    Brief description of amendment: The amendment extends the 
application of the length-based pressure boundary definition (L-
criterion) for the Westinghouse mechanical hybrid expansion joints in 
sleeved steam generator tubes to the end of operating cycle 24.
    Date of issuance: March 15, 2000.
    Effective date: Immediately upon its date of issuance and is to be 
implemented within 30 days of the date of issuance.
    Amendment No.: 146.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4266).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 15, 2000.
    No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: October 21, 1999.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.4.10, Pressurizer Safety Valves [PSV], of the 
improved Technical Specifications (TSs) issued March 31, 1999. The 
amendment reduced the safety valve set pressure in Limiting Condition 
for Operation (LCO) 3.4.10 and decreased the setpoint in Surveillance 
Requirement (SR) 3.4.10.1. The PSV setpoint and setpoint tolerance were 
changed from 2485 psig 1% to 2460 psig 2% in 
the LCO. The tolerance of 1% in the SR for resetting the 
setpoint after testing, it needed, was not changed.
    Date of issuance: March 23, 2000.
    Effective date: March 23, 2000, and shall be implemented before the 
restart from refueling outage 11, which is the next refueling outage 
scheduled to begin October 2000.
    Amendment No.: 133.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 17, 1999 (64 
FR 62718).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 23, 2000.
    No significant hazards consideration comments received: No.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room).
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By May 5, 2000, 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

[[Page 17929]]

and a petition for leave to intervene. Requests for a hearing and a 
petition for leave to intervene shall be filed in accordance with the 
Commission's ``Rules of Practice for Domestic Licensing Proceedings'' 
in 10 CFR Part 2. Interested persons should consult a current copy of 
10 CFR 2.714 which is available at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room). If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) The nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) The possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of amendment request: February 18, 2000, as supplemented March 
8, 2000.
    Description of amendment request: The amendment changes current 
Technical Specification (TS) 4.9a.2 and improved TS 3.7.5 and its 
associated bases to remove requirements associated with the backup 
steam supply to turbine-driven auxiliary feedwater pump P-8B.
    Date of issuance: March 14, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days, except that implementation with respect to the improved 
TSs shall be on or before October 31, 2000.
    Amendment No. 190.
    Facility Operating License No. DPR-20: Amendment revises the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes (65 FR 11089, March 1, 2000). The notice 
provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. No comments have been received.
    The notice also provided for an opportunity to request a hearing by 
March 31, 2000, but indicated that if the Commission makes a final NSHC 
determination, any such hearing would take place after issuance of the 
amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated March 
14, 2000.
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: Claudia M. Craig.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of application for amendment: February 25, 2000.
    Brief description of amendment: The amendment revised Technical 
Specification Table 3.3.2-1, ``Engineered Safety Feature Actuation 
System Instrumentation'' to provide a one-time exception, until the 
next time the turbine is removed from service, from the requirement to 
perform response time testing for the solenoid valve 1-FSV-47-027. The 
amendment also supersedes the Notice of

[[Page 17930]]

Enforcement Discretion granted on February 23, 2000, and confirmed by 
letter dated February 25, 2000 (00-6-004).
    Date of issuance: March 22, 2000.
    Effective date: March 22, 2000.
    Amendment No.: 23.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes (65 FR 11348 dated March 2, 2000). The notice 
provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. No comments have been received. The notice also 
provided for an opportunity to request a hearing by March 15, 2000, but 
indicated that if the Commission makes a final NSHC determination, any 
such hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and final determination of NSHC are contained in 
a Safety Evaluation dated March 22, 2000.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

    Dated at Rockville, Maryland, this 29th day of March 2000.
    For The Nuclear Regulatory Commission.
John A. Zwolinski,
Director Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-8211 Filed 4-4-00; 8:45 am]
BILLING CODE 7590-01-P