[Federal Register Volume 65, Number 173 (Wednesday, September 6, 2000)]
[Notices]
[Pages 54083-54094]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-22779]



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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 August 14, 2000, through August 25, 2000. 
The last biweekly notice was published on August 23, 2000 (65 FR 
51346).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
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 October 6, 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

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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).

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

    Date of amendment request: August 10, 2000.
    Description of amendment request: The requested amendment proposes 
to change the Technical Specifications for operations involving 
positive reactivity addition. The proposed changes revise the Required 
Actions and Limiting Condition for Operation (LCO) Notes to limit the 
introduction of reactivity such that the required SHUTDOWN MARGIN (SDM) 
or refueling boron concentration will remain satisfied.
    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:

    Carolina Power & Light (CP&L) Company has evaluated the proposed 
Technical Specifications change and has concluded that it does not 
involve a significant hazards consideration. The CP&L conclusion is 
in accordance with the criteria set forth in 10 CFR 50.92. The bases 
for the conclusion that the proposed change does not involve a 
significant hazards consideration are discussed below.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change does not involve any physical alteration of 
plant systems, structures or components. The proposed change revises 
ACTIONS in the H. B. Robinson Steam Electric Plant (HBRSEP) Unit No. 
2 Technical Specifications (TS) that require suspending operations 
involving positive reactivity additions and several Limiting 
Condition For Operation (LCO) Notes that preclude reduction in boron 
concentration. The change revises these ACTIONS and LCO Notes to 
limit the introduction of reactivity such that the required SHUTDOWN 
MARGIN (SDM) or refueling boron concentration will still be 
satisfied. The proposed change ensures that the SDM of LCO 3.1.1 and 
minimum boron concentration requirements of LCO 3.9.1 are met. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated in the Safety Analysis Report (SAR) because the 
accident analysis assumptions and initial conditions will continue 
to be maintained.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve any physical alteration of 
plant systems, structures or components. The proposed change, which 
allows positive reactivity additions that do not result in SDM or 
the refueling boron concentration being exceeded, does not introduce 
new failure mechanisms for systems, structures or components not 
already considered in the SAR [Safety Analysis Report]. Therefore, 
the possibility of a new or different kind of accident from any 
accident previously evaluated is not created because no new failure 
mechanisms or initiating events have been introduced.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change will allow positive reactivity additions, 
but the reactivity additions will not result in a[n] SDM or 
refueling boron concentration outside of the associated design basis 
limits. Allowing positive reactivity additions that do not result in 
the SDM or the refueling boron concentration being exceeded 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: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602
    NRC Section Chief: Richard P. Correia

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 request: June 19, 2000
    Description of amendment request: The proposed amendment would 
revise the technical specifications to remove their applicability 
related to the Boron Dilution Protection System (BDPS) after the next 
refueling outage for each unit. During the refueling outages, 
modifications are scheduled to be made which will permit the licensee 
to mitigate a boron dilution event without the use of the BDPS.
    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?
    The only accident potentially impacted by the proposed changes 
is the inadvertent boron dilution event.
    The Boron Dilution Protection System (BDPS) is not considered an 
initiator of any analyzed event. The BDPS performs detection and 
mitigative functions for the

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inadvertent boron dilution event. Therefore, the proposed changes 
have no impact on the probability of an event previously analyzed. 
Therefore, the proposed change does not involve a significant 
increase in the probability of occurrence of an accident previously 
evaluated.
    The proposed changes impact the consequences of an inadvertent 
dilution event due to the new requirement to manually reposition the 
Chemical and Volume Control System (CVCS) valves that isolate the 
boron dilution sources and that re-start boration of the Reactor 
Coolant System (RCS) in Modes 3, 4, and 5 (i.e., Hot Standby, Hot 
Shutdown, and Cold Shutdown, respectively). The revised detection 
and mitigation methodology being proposed achieves the same basic 
function as the existing BDPS, i.e., to prevent a return to critical 
during an inadvertent boron dilution event. The proposed changes 
will provide an improved response to the inadvertent boron dilution 
event compared to the BDPS, and thereby will prevent a return to 
critical. Therefore, the proposed changes do not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes to manually isolate potential dilution 
sources and to re-start boration of the RCS do not create the 
potential for a new or different kind of accident because the change 
results in plant configurations that have always been allowed. In 
conjunction with these proposed changes, enhancements to plant 
hardware, revisions to procedures, and administrative controls will 
be implemented. The proposed enhancements to plant hardware include 
the addition of two new redundant Volume Control Tank (VCT) high 
level alarms, which are passive in nature (i.e., do not provide any 
control function), and therefore do not create the possibility of a 
new or different kind of accident. Administrative controls and 
revisions to procedures will increase the operator's awareness of a 
potential boron dilution event and will provide the steps necessary 
to respond to a boron dilution event. As a result, the 
administrative controls and revisions to procedures do not create 
the possibility of a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The design criterion and margin of safety for the existing BDPS 
is that the inadvertent boron dilution event is terminated within a 
specified period prior to the complete loss of shutdown margin. This 
criterion will continue to be satisfied following implementation of 
the proposed changes. The proposed changes were evaluated to ensure 
that the plant operators prevent criticality in Modes 3, 4 and 5 
following an inadvertent boron dilution event, based on the revised 
analytical methodology previously discussed with the NRC and found 
to be feasible as documented in a letter from L. R. Wharton (U.S. 
NRC) to Licensees (Commonwealth Edison, Texas Utilities Electric, 
Union Electric, Wolf Creek Nuclear Operating Corporation, and 
Westinghouse), ``Utility Subgroup Technical Approach to Modify or 
Delete the Boron Dilution Mitigation System,'' dated February 8, 
1993. The proposed method of detecting and mitigating this event has 
been shown by the analysis supporting this Technical Specifications 
change request to prevent a return to critical following an 
inadvertent boron dilution event, and meets the same NRC acceptance 
criteria as specified in the Standard Review Plan (SRP), NUREG-0800, 
Section 15.4.6, ``Chemical and Volume Control System Malfunction 
That Results in a Decrease in Boron Concentration in the Reactor 
Coolant (PWR),'' dated July 1981, as applicable to the existing 
BDPS. Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: June 1, 2000
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 3.6.16 Reactor Building; and TS 5.5.11 
Ventilation Filter Testing Program. It will also revise Bases Sections 
3.6.10, 3.6.16, 3.7.12, and 3.7.13. The amendments will: (1) Enhance 
the ability to determine that reactor building annulus outside air 
inleakage is within the maximum assumed design value used in the dose 
analyses. Administrative limits are currently imposed at Catawba to 
limit inleakage in order to ensure that the dose analyses remain 
conservative. The amendments also request changes for the Unit 2 
Annulus Ventilation System (AVS) in-place penetration and bypass 
leakage criteria in TS 5.5.11. This portion of the amendments affects 
TS Bases 3.6.10, TS 3.6.16 and Bases, and TS 5.5.11; (2) Describe the 
alignment the Auxiliary Building Filtered Ventilation Exhaust System 
(ABFVES) filtered exhaust units should be tested in and request 
appropriate TS 5.5.11 limits in order to ensure that the ABFVES will 
continue to meet its design basis functions. Similar to Item 1 above, 
the amendments also request changes for the Unit 2 ABFVES in-place 
penetration and bypass leakage criteria in TS 5.5.11. This portion of 
the amendments affects TS Bases 3.7.12 and TS 5.5.11; and (3) Modify 
the TS Bases for the Fuel Handling Ventilation Exhaust System (FHVES) 
and similar to Items 1 and 2 above, the amendments also request changes 
for the Unit 2 FHVES in-place penetration and bypass leakage criteria 
in TS 5.5.11. This portion of the amendments affects TS Bases 3.7.13 
and TS 5.5.11.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The following discussion is a summary of the evaluation of the 
changes contained in this proposed amendment against the 10 CFR 
50.92(c) requirements to demonstrate that all three standards are 
satisfied. A no significant hazards consideration is indicated if 
operation of the facility in accordance with the proposed amendment 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated, or
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. Involve a significant reduction in a margin of safety.

First Standard

    Implementation of this amendment would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. Neither the AVS, nor the ABFVES, nor the FHVES 
is capable of initiating any accident. The AVS, ABFVES, and FHVES, 
which are responsible for maintaining an acceptable environment in 
the annulus, the auxiliary building, and the fuel building during 
normal and accident conditions, will continue to function as 
designed, and in accordance with all applicable TS. The design and 
operation of the systems are not being modified by this proposed 
amendment. Therefore, there will be no impact on any accident 
probabilities or consequences.

Second Standard

    Implementation of this amendment would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. No new accident causal mechanisms are created 
as a result of NRC approval of this amendment request. No changes 
are being made to the plant which will introduce any new accident 
causal mechanisms. This amendment request does not impact any plant 
systems that are accident initiators and does not impact any safety 
analyses.

[[Page 54086]]

Third Standard

    Implementation of this amendment would not involve a significant 
reduction in a margin of safety. Margin of safety is related to the 
confidence in the ability of the fission product barriers to perform 
their design functions during and following an accident situation. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment system. The performance of these fission 
product barriers will not be impacted by implementation of this 
proposed amendment. The performance of the AVS, the ABFVES, and the 
FHVES in response to normal and accident conditions will not be 
impacted by this proposed amendment. The changes to the AVS 
surveillances will provide for a better method to ensure that the 
assumptions of the dose analyses are met. There is no risk 
significance to this proposed amendment, as no reduction in system 
or component availability will be incurred. No safety margins will 
be impacted.
    Based upon the preceding discussion, Duke has concluded that the 
proposed amendment does not involve a significant hazards 
consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: August 10, 2000.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to allow an alternate storage 
configuration of fuel assemblies adjacent to the walls within Region 1 
of the spent fuel pool.
    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 probability of fuel handling accidents (dropped assemblies, 
misplaced/misloaded assemblies, etc.) is not changed by utilizing 
the previously described vacant spaces that are face adjacent to the 
SFP [spent fuel pool] walls in Region I [Region 1] to store design 
basis assemblies that are less reactive than RI A [Region 1 
Configuration A] type assemblies. Fuel assemblies of different types 
are presently stored face adjacent to these walls. This proposal 
will allow additional assemblies to be located face adjacent to the 
Region I SFP walls and does not effect the precursors to any 
postulated spent fuel pool accidents.
    The consequences of an accident different than that previously 
analyzed additionally remains unchanged. Evaluations have 
demonstrated that the fuel handling accident reactivity values will 
remain less than the 0.95 Keff acceptance criteria in the 
event of a fuel handling accident, assuming an initial SFP boron 
concentration of 1000 ppm. The boron concentration limit is 
additionally bounded by ANO-2 [Arkansas Nuclear One, Unit 2] TS 
[Technical Specification] Limiting Condition for Operation (LCO) 
3.9.12.c which limits SFP boron to greater than 1600 ppm at all 
times.
    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.
    As discussed previously, the proposed SFP configuration will not 
result in exceeding the acceptance criteria of 0.95 Keff 
during normal or accident conditions. Since fuel assemblies are 
currently located along the Region I SFP walls, no new or different 
kind of accident than that previously evaluated exists. Locations 
required to be vacant will remain physically blocked. In the event 
that a ``misloading'' type accident occurs in this region, 
evaluations have shown that the fuel handling accident reactivity 
values will remain well below 0.95 Keff when initial SFP 
boron concentrations are at or above 1000 ppm, which is 
significantly less than the TS boron limit of 1600 ppm.
    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.
    As previously discussed, the proposed configuration will not 
result in exceeding the 0.95 Keff acceptance criteria 
during normal operations that assume zero concentration of boron at 
the maximum water density in the SFP or during accident conditions 
that assume an initial SFP boron concentration of at least 1000 ppm. 
Furthermore, ANO-2 TS 3.9.12.c requires SFP boron to be maintained 
greater than 1600 ppm at all times. Fuel assemblies are presently 
stored along the Region I SFP walls; therefore, storing additional 
assemblies along these same walls will not significantly reduce the 
margin to safety since it has been shown that the current CSA 
[criticality safety analysis] remains valid.
    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, Winston and Strawn, 
1400 L Street, NW., Washington, DC 20005-3502
    NRC Section Chief: Robert A. Gramm

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

    Date of amendment request: May 12, 2000, as supplemented June 19, 
2000.
    Description of amendment request: The proposed amendment would 
revise the Beaver Valley Power Station, Units 1 and 2 (BVPS-1 and 2), 
calculated doses and associated descriptions/information listed in the 
Updated Final Safety Analysis Reports (UFSARs) for the Design Basis 
Accidents (DBAs). An evaluation of all of the BVPS-1 and 2 dose 
calculations was completed which reviewed the input parameter values, 
the input assumptions, and the methodologies used. Some of the input 
parameter values, input assumptions and methodologies used in the DBA 
dose calculations were revised. The resultant DBA dose calculation 
revisions necessitate associated revisions to the UFSARs. Additionally, 
some changes would be made in response to Generic Letter 99-02. For 
BVPS-1, the requested amendment would affect the analyses for the 
following DBAs: loss of offsite AC power, fuel-handling accident, 
accidental release of waste gas, steam generator tube rupture, major 
secondary system pipe rupture, rod cluster control assembly ejection, 
single reactor coolant pump locked rotor, and loss of reactor coolant 
from small ruptured pipes/loss-of-coolant accidents. For BVPS-2, the 
requested amendment would affect the analyses for the following DBAs: 
steam system piping failures, loss of AC power, reactor coolant pump 
shaft seizure, rod cluster control assembly ejection, failure of small 
lines carrying primary coolant outside containment, steam generator 
tube rupture, loss-of-coolant accidents, waste gas system failure, and 
fuel-handling accidents.
    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?
    Following a reevaluation of the calculated dose values for BVPS 
Unit 1 and Unit 2

[[Page 54087]]

design basis accidents (DBAs) as described in their respective 
[Updated Final Safety Analysis Report] UFSAR, several calculated 
dose values were identified to be increased. These increases were 
small and remained within the applicable DBA previously approved 
regulatory limit.
    The increases for each DBA were as a result of revised plant 
data being used in the dose calculation, revised calculation 
assumptions, or new methodology. These changes were not the result 
of plant hardware changes. The changes were intended to ensure that 
accurate, current and conservative licensing basis information and 
assumptions were used for DBA dose analyses. The UFSAR changes are 
proposed to reflect the revised analyses results for the Unit 1 and 
Unit 2 UFSAR.
    Since the calculated DBA radiological doses remain within the 
applicable DBA previously approved regulatory limit, these 
calculated dose values do not involve a significant increase in the 
probability or consequences of an accident as previously evaluated 
in the BVPS Unit 1 and Unit 2 UFSAR.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    BVPS Unit 1 and Unit 2 calculations which are used to determine 
DBA calculated dose values were revised. The changes were as a 
result of revised plant data being used in the dose calculation, 
revised calculation assumptions or new methodology. The changes were 
intended to ensure that accurate, current and conservative licensing 
basis information and assumptions were used for DBA dose analyses. 
The DBA events themselves remain the same postulated events as 
previously described within the BVPS Unit 1 and Unit 2 UFSARs. The 
revised dose calculations do not create the possibility of a new or 
different kind of accident from the DBA accidents previously 
evaluated in the UFSAR. These changes were not the result of plant 
hardware changes. The changes were only in the calculations. The 
UFSAR changes are proposed to reflect the revised analyses['] 
results for the Unit 1 and Unit 2 UFSAR.
    3. Does the change involve a significant reduction in a margin 
of safety?
    This amendment request addresses only proposed changes to the 
Unit 1 and Unit 2 UFSAR, which was determined to involve an 
Unreviewed Safety Question pursuant to 10 CFR 50.59. This request 
does not propose modifying any Technical Specification criteria. 
This request proposes that several calculated dose values for BVPS 
Unit 1 and Unit 2 DBAs be increased following a reevaluation of 
their design basis calculations. These proposed increases are small 
and remained within the applicable DBA previously approved 
regulatory limit. Thus, the proposed changes to the UFSAR which 
originated from revised BVPS DBA dose calculations [do] not involve 
a significant reduction in the margin of safety for BVPS Unit 1 and 
Unit 2 because the Technical Specifications will not be altered and 
the increase in calculated dose values is small and remains within 
regulatory approved limits.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Marsha Gamberoni

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone 
Nuclear Power Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: July 31, 2000
    Description of amendment request: The amendment would change 
Technical Specifications 3.8.1.1, ``Electrical Power Systems--A.C. 
Sources--Operating,'' and 3.8.1.2, ``Electrical Power Systems--A.C. 
Sources--Shutdown.'' The index and the Bases for these Technical 
Specifications will be modified as a result of the proposed changes. 
The proposed changes will allow certain emergency diesel generator 
(EDG) surveillance requirements to be performed when the plant is 
operating instead of shut down as currently required. Additional 
changes will remove EDG accelerated testing and special reporting 
requirements, and the surveillance requirement to perform EDG 
inspections. EDG inspections will still be performed as recommended by 
the manufacturer. The proposed changes will not adversely impact the 
type and amounts of effluents that may be released off site.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff's analysis is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed Technical Specification changes are associated with 
the surveillance requirements for the Emergency Diesel Generators 
(EDGs) and will not affect the ability of the EDGs to perform their 
intended safety function. Therefore, the proposed Technical 
Specification changes will not result in 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.
    Since there are no changes in components, component operation, or 
system operation, this change does not create the possibility of an 
accident of a different type.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes will have no adverse effect on plant operation 
or equipment important to safety. The plant response to the design 
basis accidents will not change and the accident mitigation equipment 
will continue to function as assumed in the design basis accident 
analysis. Therefore, there will be no significant reduction in a margin 
of safety.
    Based on the staff's analysis, it appears that the three standards 
of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut
    NRC Section Chief: James W. Clifford

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

    Date of amendment request: June 14, 2000
    Description of amendment request: The proposed amendments would 
revise Vogtle's Surveillance Requirements (SR) 3.8.1.9 and 3.8.1.14 to 
reduce the emergency Diesel Generator (EDG) loading requirements from 
6800 kW and 7000 kW to 6500 kW and 
7000 kW. These changes will make the above SRs consistent 
with SR 3.8.1.3 and 3.8.1.13 which are in the current Technical 
Specifications (TS). In addition, the proposed amendments would revise 
TS section 5.6.7, ``EDG Failure Report'', to correct a typographical 
error.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed change to section 5.6.7 is administrative only 
since it does nothing more than correct a typographical error. The 
proposed changes to the DG loading requirements specified in SRs 
3.8.1.9 and

[[Page 54088]]

3.8.1.14 have no impact on or relationship to the probability of any 
of the initiating events assumed for the accidents previously 
evaluated. Therefore, the proposed changes do not involve a 
significant increase in the probability of any accident previously 
evaluated. Furthermore, since the proposed loading requirements 
bound the maximum expected loading for the DGs, SRs 3.8.1.9 and 
3.8.1.14 will continue to demonstrate that the DGs are capable of 
performing their safety function. Since the proposed changes do not 
adversely affect the capability of the DGs to perform their safety 
function, the outcome of the accidents previously evaluated (i.e., 
radiological consequences) will not be affected. Therefore, the 
proposed changes do not involve a significant increase in the 
consequences of any accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    No. The proposed change to section 5.6.7 is administrative only 
since it does nothing more than correct a typographical error. The 
proposed changes to the DG loading requirements specified in SRs 
3.8.1.9 and 3.8.1.14 will not introduce any new equipment or create 
new failure modes for existing equipment. Other than the reduced 
loading requirements for the DGs, the proposed changes will not 
affect or otherwise alter plant operation. The DGs will remain 
capable of performing their safety function. No other safety related 
or important to safety equipment will be affected by the proposed 
changes. Therefore, the proposed changes will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    No. The proposed change to section 5.6.7 is administrative only 
since it does nothing more than correct a typographical error. The 
proposed changes reduce the loading requirements of SRs 3.8.1.9 and 
3.8.1.14. The new loading requirements bound the maximum expected 
loading of the DGs under the worst case scenario, and they are 
consistent with the regulatory guidance found in Regulatory Guide 
(RG) 1.9, Revision 3, ``Selection, Design, and Qualification of 
Diesel-Generator Units Used as Standby (Onsite) Electric Power 
Systems at Nuclear Power Plants,'' July 1993. Reduction in wear and 
tear should inherently increase the reliability of the DGs. 
Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

Conclusion

    Based on the above evaluation, the proposed changes do not 
involve a significant hazard as defined in 10 CFR 50.92.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216
    NRC Section Chief: Richard L. Emch, Jr.

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: August 11, 2000 (TS-400).
    Description of amendment request: The proposed amendment would 
change the Units 2 and 3 Technical Specifications to revise the testing 
frequency for certain isolation valves of a type known as excess flow 
check valves (EFCV). The proposed testing frequency would allow a 
representative sample to be tested every 24 months, such that each EFCV 
is tested at least once every 120 months. The current specification 
requires that each EFCV be tested at least once every 24 months.
    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 current excess flow check valve (EFCV) frequency requires 
that each reactor instrument line EFCV be tested every 24 months. 
The EFCVs are designed to automatically close upon excessive 
differential pressure including failure of the down stream piping or 
instrument and will reopen when appropriate. This proposed change 
will allow a reduction in the number of EFCVs that are verified 
tested every 24 months, to approximately 20 percent of the valves 
each cycle. BFN and industry operating experience demonstrates high 
reliability of these valves. Neither the EFCVs or their failure is 
capable of initiating a previously evaluated accident. Therefore, 
there is no increase in the probability of occurrence of an accident 
previously evaluated.
    The instrument lines going to the Reactor Coolant Pressure 
boundary with EFCVs installed have flow restricting devices upstream 
of the EFCV. The consequences of a unisolable failure of an 
instrument line has been previously evaluated and meets the intent 
of NRC Safety Guide 11. The offsite exposure has been calculated to 
be substantially below the limits of 10 CFR 100. Additionally, 
coolant lost from such a break is inconsequential compared to the 
makeup capabilities of normal and emergency makeup systems. Although 
not expected to occur as a result of this change, the effects of a 
postulated failure of an EFCV to isolate and instrument line break 
as a result of reduced testing are bounded by TVA analysis.
    Therefore, the proposed amendment 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 proposed change reduces the number of EFCVs tested each 
operating cycle. No other changes to the TS are being proposed. BFN 
and industry operating experience demonstrates that these valves are 
highly reliable, a proposed reduction in test frequency is bounded 
by previous evaluation of a line rupture. The change will not alter 
the operation of process variables, structures, systems or 
components described in the BFN Updated Final Safety Analysis 
Report. Therefore, reduction in the number of EFCVS tested each 
cycle does not result in the possibility of a new or different kind 
of accident.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The consequences of an unisolable rupture of an instrument line 
has been previously evaluated and meets the intent of NRC Safety 
Guide 11. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    Therefore, the proposed revised surveillance frequency does not 
adversely affect the public health and safety, and does not involve 
any significant safety hazards.

    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 I0H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

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

    Date of application for amendments: August 4, 2000 (TS 99-20)
    Brief description of amendments: The proposed amendments would 
change the Sequoyah Nuclear Plant (SQN) Technical Specifications (TS), 
Section 6.2.2, to change the title of various shift members and to 
change the Shift Technical Advisor requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the

[[Page 54089]]

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 title change of Shift Operations Supervisor to Shift Manager 
is administrative. The elimination of TS 6.2.2.b and Table 6.2-1 is 
considered an administrative change. These two items contain similar 
requirements as those contained in 10 CFR 50.54(m)(2)(iii), 10 CFR 
50.54(m)(2)(i), and 10 CFR 50.54(k). These sections are considered a 
duplicate of the requirements contained in the Code of Federal 
Regulations. This request also eliminates the title of Shift 
Technical Advisor (STA) but will not eliminate or reduce licensee 
responsibilities in this area. This request is based on an NRC 
policy statement, contained in Generic Letter 86-04, that supports 
the transition of engineering expertise from the STA position to 
another individual on shift who possesses the mandated education 
qualifications. The proposed administrative and organizational 
changes do not result in any 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.
    As described above, the proposed changes are administrative and 
organizational in nature and cannot create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    As described above, the proposed changes are administrative and 
organizational in nature. The proposed changes are based on approved 
NRC guidance. The margin of safety is, therefore, not reduced.

    The NRC 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.

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.

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

    Date of amendment request: July 21, 2000.
    Description of amendment request: The amendment requests approval 
to remove a shutdown requirement with regard to the relief valve 
position indication system in Section 3.13 of the Technical 
Specifications.
    Date of publication of individual notice in Federal Register: 
August 2, 2000 (65 FR 47520).
    Expiration date of individual notice: September 1, 2000.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for a Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, 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. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

    Date of amendment request: July 21, 2000.
    Description of amendment request: The proposed amendment revises 
the Oyster Creek Nuclear Generating Station Technical Specifications 
Section 3.13 to remove a shutdown requirement with regard to the relief 
valve position indication system.
    Date of issuance: August 21, 2000.
    Effective Date: As of date of issuance to be implemented within 30 
days.
    Amendment No.: 214.
    Facility Operating License No. DPR-16: This amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (65 FR 47520) August 2, 2000. That 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 September 1, 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 finding of 
exigent circumstances, state consultation, and final determination of 
no significant hazards consideration determination are contained in a 
Safety Evaluation dated August 21, 2000.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, N.W., Washington, D.C. 20036-5869.
    NRC Section Chief: Marsha Gamberoni.

[[Page 54090]]

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 0-530, Palo Verde Nuclear Generating Station, Units Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of application for amendments: June 6, 2000.
    Brief description of amendments: The amendments revise the 
information in Figure 3.5.5-1, ``Minimum Required RWT Volume in TS 
3.5.5, Refueling Water Tank (RWT),'' for the three units. The 
amendments relocate design information to the Bases of the TSs, 
truncate the lower end of the RWT limit curve at 210  deg.F, retitle 
the right-hand ordinate from ``minimum useful volume required in the 
RWT'' to ``RWT Volume,'' and delete the two footnotes and the 
references to the footnotes.
    Date of issuance: August 18, 2000.
    Effective date: August 18, 2000, to be implemented within 45 days 
of the date of issuance.
    Amendment Nos.: Unit 1-127, Unit 2-127, Unit 3-127.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: July 12, 2000 (65 FR 
43043).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 18, 2000.
    No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket No. 50-237, Dresden Nuclear Power 
Station, Unit 2, Grundy County, Illinois

    Date of application for amendment: April 30, 1999.
    Brief description of amendment: The amendment revised the 
expiration date of the operating license to allow 40 years of operation 
from the original date of issuance of the Provisional Operating 
License.
    Date of issuance: August 24, 2000.
    Effective date: August 24, 2000.
    Amendment No.: 178.
    Facility Operating License No. DPR-19: The amendment revised the 
Facility Operating License. Date of initial notice in Federal Register: 
March 22, 2000 (65 FR 15376).
    The Commission's related evaluation of the amendment is contained 
in an Environmental Assessment dated June 1, 2000, and a Safety 
Evaluation dated August 24, 2000.
    No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, WNP-2, Benton County, Washington

    Date of application for amendment: November 18, 1999, as 
supplemented by letter dated June 7, 2000.
    Brief description of amendment: The amendment changes Technical 
Specification 5.5.7, ``Ventilation Filter Testing Program (VFTP)'' to 
include the requirement for laboratory testing of engineered safety 
feature ventilation system charcoal samples per American Society for 
Testing and Materials D3803-1989 and the application of a safety factor 
of 2.0 to the charcoal filter efficiency assumed in the plant design-
basis dose analyses.
    Date of issuance: August 25, 2000.
    Effective date: August 25, 2000.
    Amendment No.: 167.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73088).
    The June 7, 2000, supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noticed, and did not change the staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 25, 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: April 9, 1998, as supplemented by 
letters dated January 13, 1999, and June 28, 2000.
    Brief description of amendment: The amendment consists of changes 
to the River Bend Station (RBS) Facility Operating License, paragraph 
2.C(13). The amendment allows RBS to operate with final feedwater 
temperature reduction in order to extend the fuel cycle by maintaining 
the core thermal power at or close to rated power, thus delaying the 
start of normal coastdown. The January 13, 1999, letter provided a 
revised proprietary version of the licensee's analysis submitted in its 
original April 9, 1998, application and the June 28, 2000, letter 
provided additional information to support staff review of the original 
application, and did not affect the initial finding of no significant 
hazards consideration determination dated May 20, 1998 (63 FR 27762).
    Date of issuance: August 22, 2000.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment No.: 112.
    Facility Operating License No. NPF-47: The amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: May 20, 1998 (63 FR 
27762).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 22, 2000.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc. Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Units 1 and 2, Pope County, Arkansas

    Date of amendment request: July 14, 1999, as supplemented by 
letters dated February 24, 2000, and July 17, 2000.
    Brief description of amendments: The proposed amendments delete 
requirements from the Technical Specifications to maintain a Post 
Accident Sampling System (PASS). Licensees were required to implement 
PASS upgrades as a result of NUREG-0737, ``Clarification of TMI [Three 
Mile Island Nuclear Station] Action Plan Requirements,'' and Regulatory 
Guide 1.97, Revision 3, ``Instrumentation for Light Water Cooled 
Nuclear Power Plants to Assess Plant and Environmental Conditions 
During and Following an Accident.'' Implementation of these upgrades 
were an outcome of the Nuclear Regulatory Commission's lessons learned 
from the accident that occurred at TMI, Unit 2. The staff has concluded 
that the information obtained using PASS is not required for the 
development of protective action recommendations or for core damage 
assessment.
    Date of issuance: August 17, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 208 and 218
    Facility Operating License Nos. DPR-51 and NPF-6: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43773). The supplements dated February 24 and July 17, 2000, did not 
change the scope of the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 17, 2000.
    No significant hazards consideration comments received: No

[[Page 54091]]

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio

    Date of application for amendment: June 1, 2000, as supplemented by 
letter dated June 30, 2000.
    Brief description of amendment: The amendment approves a proposed 
modification that changes the Perry Nuclear Power Plant as described in 
the Updated Safety Analysis Report by installing inflatable seals that 
surround the Emergency Service Water (ESW) alternate intake sluice 
gates. This modification is necessary so that the licensee may use 
inflatable seals to minimize leakage of warm water into the ESW forebay 
from the Service Water discharge and thus maintain the ESW temperature 
below the design limit.
    Date of issuance: August 22, 2000
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 114
    Facility Operating License No. NPF-58: This amendment authorizes 
revision of the Updated Safety Analysis Report.
    Date of initial notice in Federal Register: June 14, 2000 (65 FR 
37414) The supplemental information contained clarifying information 
and did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register Notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 22, 2000.
    No significant hazards consideration comments received: No

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa

    Date of application for amendment: May 10, 1999, as supplemented 
April 6, April 26, and June 5, 2000.
    Brief description of amendment: Changes Technical Specifications to 
establish the actions to be taken for an inoperable ``Standby Filter 
Unit'' (SFU) System due to a degraded control building boundary.
    Date of issuance: August 11, 2000
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 233
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38029). The April 6, April 26, and June 5, 2000, submittals provided 
additional clarifying information that did not change the initial 
proposed no significant hazards consideration determination or expand 
the scope of the application beyond the initial notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 11, 2000.
    No significant hazards consideration comments received: No

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

    Date of application for amendment: June 7, 2000
    Brief description of amendment: The amendment revised the Technical 
Specifications, Section 3.10.8, ``SHUTDOWN MARGIN (SDM) Test -- 
Refueling,'' correcting an administrative error introduced when 
Amendment No. 92, dated March 2, 2000, was issued.
    Date of issuance: August 24, 2000
    Effective date: As of the date of issuance to be implemented 
concurrently with Amendment No. 92.
    Amendment No.: 93
    Facility Operating License No. NPF-69: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 16, 2000 (65 FR 
37807)
    The staff's related evaluation of the amendment is contained in a 
Safety Evaluation dated August 24, 2000.
    No significant hazards consideration comments received: No

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone 
Nuclear Power Station, Unit No. 3, New London County, Connecticut

    Date of application for amendment: February 1, 2000, as 
supplemented on April 13, 2000
    Brief description of amendment: The amendment temporarily suspends 
the technical (TSs) requirements for TSs 3.7.7 and 3.7.8 in order to 
conduct testing of the cable spreading room that will pressurize the 
area to a pressure that exceeds the adjacent control room envelope 
area.
    Date of issuance: August 22, 2000
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 181
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 31, 2000 (65 FR 
34748)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 22, 2000.
    No significant hazards consideration comments received: No

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

    Date of application for amendment: October 29, 1999, as 
supplemented March 14 and April 25, 2000
    Brief description of amendment: The amendment conforms the license 
to reflect the transfer of possession under Operating License No. DPR-
22 to a newly formed utility operating company subsidiary of Northern 
States Power Company merged with New Century Energies, Inc., as 
approved by Order of the Commission dated May 12, 2000.
    Date of issuance: August 18, 2000
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 111
    Facility Operating License No. DPR-22. Amendment revised the 
Operating License.
    Date of initial notice in Federal Register: February 10, 2000 (65 
FR 6641)
    The March 14 and April 25, 2000, supplements were within the scope 
of the initial application as originally noticed.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 12, 2000.
    No significant hazards consideration comments received: No

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

    Date of application for amendment: February 29, 2000, as 
supplemented July 10, 2000
    Brief description of amendment: The amendment (1) approves 
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, (2) revises the Technical Specifications (TSs) to 
reflect modifications to the EFT system that eliminate the need for 
additional test exceptions, (3) revises the TSs to be consistent with 
the guidance of NRC Generic Letter 99-02, and (4) revises the TSs to 
include operability requirements for the EFT system during operations 
that could result in a fuel handling accident.
    Date of issuance: August 18, 2000
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.

[[Page 54092]]

    Amendment No.: 112
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in  Federal Register: April 5, 2000 (65 FR 
17917)
    The July 10, 2000, supplemental letter provided clarifying 
information that was within the scope of the original application 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 August 18, 2000.
    No significant hazards consideration comments received: No

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, and Docket No. 72-10, 
Prairie Island Independent Spent Fuel Storage Installation, Goodhue 
County, Minnesota

    Date of application for amendments: October 29, 1999, as 
supplemented March 14 and April 25, 2000.
    Brief description of amendments: The amendments conform the 
licenses to reflect the transfer of possession under Operating Licenses 
Nos. DPR-42 and DPR-60 and Materials License No. SNM-2506 to a newly 
formed utility operating company subsidiary of Northern States Power 
Company merged with New Century Energies, Inc., as approved by Order of 
the Commission dated May 12, 2000.
    Date of issuance: August 18, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 154 and 145.
    Facility Operating Licenses Nos. DPR-42 and DPR-60 and Materials 
License No. SNM-2506: Amendments revised the Operating Licenses and 
Materials License.
    Date of initial notice in Federal Register: February 10, 2000 (65 
FR 6642)
    The March 14 and April 25, 2000, supplements were within the scope 
of the initial application as originally noticed.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 12, 2000.
    No significant hazards consideration comments received: No

PECO Energy Company, Public Service Electric and Gas Company Delmarva 
Power and Light Company; and Atlantic City Electric Company, Docket 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: August 11, 1999, as 
supplemented June 29, 2000.
    Brief description of amendments: The Updated Final Safety Analysis 
Report (USFAR) was updated to reflect credit for use of a limited 
amount of containment overpressure in calculations of net positive 
suction head available for emergency core cooling pumps.
    Date of issuance: August 14, 2000.
    Effective date: As of Date of issuance.
    Amendments Nos.: 233 and 237.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
authorized changes to the UFSAR.
    Date of initial notice in Federal Register: April 19, 2000 (65 FR 
21038). The June 29, 2000, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated August 14, 2000.
    No significant hazards consideration comments received: No

Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek 
Generating Station, Salem County, New Jersey

    Date of application for amendment: June 4, 1999, as supplemented 
October 22, 1999.
    Brief description of amendment: The amendment revises the license 
and Technical Specifications to reflect changes related to the transfer 
of the license for the Hope Creek Generating Station, to the extent 
held by Public Service Electric and Gas Company, to PSEG Nuclear 
Limited Liability Company.
    Date of issuance: August 21, 2000
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 129
    Facility Operating License No. NPF-57: This amendment revised the 
License and the Technical Specifications.
    Date of initial notice in  Federal Register: June 30, 1999 (64 FR 
35193). The October 22, 1999, supplement provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the scope of the original 
Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 16, 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: April 13, 2000
    Brief description of amendments: The amendments deleted Technical 
Specification (TS) 3/4.1.3.2.2 which is related to shutdown and control 
rod group demand position indication in Modes 3, 4, and 5.
    Date of issuance: August 17, 2000
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days of issuance.
    Amendment Nos.: 232 and 213
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 28, 2000 (65 FR 
39960)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 17, 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: June 4, 1999, as supplemented 
October 22, 1999.
    Brief description of amendments: The amendment revises the license 
and Technical Specifications to reflect changes related to the transfer 
of the license for the Salem Nuclear Generating Station, Unit Nos. 1 
and 2, to the extent held by Public Service Electric and Gas Company, 
to PSEG Nuclear Limited Liability Company.
    Date of issuance: August 21, 2000
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment Nos.: 233 and 214
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the License and Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35192). The October 22, 1999, supplement provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the scope of the original 
Federal Register notice.

[[Page 54093]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 16, 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: January 24, 2000, as 
supplemented April 19 and May 31, 2000.
    Brief description of amendments: The amendments revise the 
radiological effluent technical specifications (RETS) and 
administrative controls requirements (i.e., Sections 3/4.3, 
Instrumentation, 3/4.11, Radioactive Effluents, 3/4.12, Radiological 
Environmental Monitoring, 6.0, Administrative Controls, and the table 
of contents and definitions) in the Technical Specifications (TSs) by 
implementing programmatic controls for RETS in the administrative 
controls section and relocating procedural details of the RETS, with 
various changes, to the offsite dose calculation manual (ODCM) or to 
the process control program (PCP). The proposed changes follow the 
guidance and requirements in NRC Generic Letter 89-01, ``Implementation 
of Programmatic Controls in the Technical Specifications for 
Radiological Effluent Technical Specifications (RETS) in the 
Administrative Controls Section of the Technical Specifications and the 
Relocation of Procedural Details of RETS to the Offsite Dose 
Calculation Manual or to the Process Control Program,'' that was issued 
in 1989. There is also the change to add the word ``oxygen'' to the 
title of ``Radioactive Gaseous Effluent Monitoring Instrumentation.''
    Date of issuance: August 24, 2000
    Effective date: August 24, 2000
    Amendment Nos.: 234 and 215
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 1, 2000 (65 FR 
11094) The supplemental letters dated April 19 and May 31, 2000, 
provided clarification that did not alter the scope of the proposed 
action or the initial no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 24, 2000.
    No significant hazards consideration comments received: No

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

    Date of application for amendments: August 24, 1999, as 
supplemented on December 29, 1999, and June 16, 2000
    Brief description of amendments: The amendments revised Technical 
Specification 3.3.2 ``Engineered Safety Features Actuation System 
(ESFAS) Instrumentation'' to relax the slave relay test frequency from 
quarterly to every refueling not to exceed 18 months.
    Date of issuance: August 22, 2000
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 114 and 92
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 22, 2000 (65 FR 
15386). The supplemental letters dated December 29, 1999, and June 16, 
2000, provided clarifying information only, and did not change the 
scope of the August 24, 1999, application nor the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 22, 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: March 6, 2000
    Brief description of amendment: Revised the Technical Specification 
(TS) and associated Bases for Limiting Condition for Operation 3.9.4, 
``Refueling Operations--Containment Penetrations,'' to allow the 
containment personnel airlock doors and certain containment 
penetrations to be open during refueling activities under appropriate 
administrative controls.
    Date of issuance: August 24, 2000
    Effective date: August 24, 2000
    Amendment No.: 26
    Facility Operating License No. NPF-90: Amendment revises the TS.
    Date of initial notice in Federal Register: May 17, 2000 (65 FR 
31361)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 24, 2000.
    No significant hazards consideration comments received: No

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

    Date of application for amendments: June 22, 2000, as supplemented 
July 25, 2000
    Brief description of amendments: The amendments revise the 
Technical Specifications Sections 3.4.1.4, 3.4.1.6, 4.4.1.4, and 
4.4.1.6.1; add Sections 4.4.1.6.4 and 4.4.1.6.5; and revise Bases 
Section 3/4.4.1 for Units 1 and 2. These changes will allow for the 
implementation of a vacuum-assisted backfill technique when returning 
an isolated Reactor Coolant System (RCS) loop to service, and provide 
the necessary controls for temperature and boron concentration of the 
isolated RCS loop to ensure the required shutdown margin is maintained.
    Date of issuance: August 25, 2000
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 223 and 204
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments change 
the Technical Specifications.
    Date of initial notice in Federal Register: July 26, 2000 (65 FR 
46019). The letter dated July 25, 2000, contained clarifying 
information only, and did not change the initial no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 25, 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: May 19, 2000, as supplemented 
August 3, 2000
    Brief description of amendments: These amendments eliminate one of 
the license conditions and associated implementation dates from 
Appendix C to the licenses. The license condition required the licensee 
to submit a license amendment application and supporting radiological 
dose analyses demonstrating compliance with General Design Criterion 19 
dose limits without reliance on potassium iodide.
    Date of issuance: August 15, 2000
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 198 and 203

[[Page 54094]]

    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Operating Licenses.
    Date of initial notice in Federal Register: June 6, 2000 (65 FR 
35966)
    The August 3, 2000, supplemental letter provided clarifying 
information that was within the scope of the original application and 
did not change the staff's initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 15, 2000.
    No significant hazards consideration comments received: No

    Dated at Rockville, Maryland, this 30th day of August 2000.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-22779 Filed 9-5-00; 8:45 am]
BILLING CODE 7590-01-P