[Federal Register Volume 65, Number 27 (Wednesday, February 9, 2000)]
[Notices]
[Pages 6402-6417]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-2835]


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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 January 14, 2000, through January 28, 2000. 
The last biweekly notice was published on January 26, 2000 (65 FR 
4268).

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 March 10, 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

[[Page 6403]]

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).

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Dated of amendments request: January 25, 2000.
    Description of amendments request: The proposed amendment requests 
a revision to the definition of Response Time Testing (RTT) for the 
Reactor Protective System (RPS) and Engineered Safety Features 
Actuation System (ESFAS). The revision allows use of either an 
allocated sensor response time or a measured sensor response time for 
pressure sensors used in channels of RPS and ESFAS. The request is 
based on Combustion Engineering NPSD-1167, Revision 1, ``Elimination of 
Pressure Sensor Response Time Testing Requirements--CEOG Task 1070.''
    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 licensing basis change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated in the safety analysis report.
    This change to the licensing basis does not result in a 
condition where the design, material, and construction standards 
that were applicable prior to the change are altered. The same 
Reactor Protective System and Engineered Safety Features Actuation 
System instrumentation is being used; the time response allocations/
modeling assumptions in Updated Final Safety Analysis Report Chapter 
14 analyses remain the same; only the method of verifying time 
response is changed. The proposed change will not modify any system 
interface and could not increase the likelihood of an accident since 
these events are independent of this change. The proposed activity 
will not change, degrade or prevent actions or alter any assumptions 
previously made in evaluating the radiological consequences of an 
accident described in the Updated Final Safety Analysis Report. 
Therefore, the proposed amendment does not result in any increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed licensing basis change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated in the safety analysis report.
    This change does not alter the performance of the pressure and 
differential pressure sensors used in the plant protection systems. 
These sensors will still have their response time verified before 
they are placed in operational service and after any maintenance to 
them that could affect their response time. Changing the method of 
periodically verifying instrument response for certain sensor 
(assuring equipment

[[Page 6404]]

operability) from time response testing to calibration, use of 
actual data, and channel checks will not create any new accident 
initiators or scenarios. Periodic surveillance of these instruments 
will detect significant degradation in the sensor response 
characteristic. Implementation of the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed licensing basis change does not involve a 
significant reduction in margin of safety.
    The total Reactor Protective System and Engineered Safety 
Features Actuation System response time assumed in the safety 
analysis is not affected by this change. The periodic system 
response time verification method for selected pressure and 
differential pressure sensors is modified to allow the use of 
allocated data based on actual test results or other verifiable 
response time data. Verification methods and calibration tests 
assure that any degradation sufficient to significantly affect 
sensor response time will be detected before the total system 
response time exceeds that defined in the safety analysis. 
Therefore, it is concluded that the proposed change does not result 
in a significant reduction in margin with respect to plant safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Marsha Gamberoni, Acting.

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

    Dated of amendment request: November 18, 1999.
    Description of amendment request: The proposed amendment would 
remove license condition 3.H, ``Long Term Program,'' from Facility 
Operating License DPR-35 for the Pilgrim Nuclear Power Station.
    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 does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. No physical changes to the facility will occur 
as a result of this amendment. Work activities will continue to 
receive the appropriate level of review in accordance with Pilgrim 
procedures and practices. The organizational structure and processes 
that control and manage these activities ensure activities are 
prioritized and performed in a manner consistent with plant safety. 
The proposed amendment removes an administrative burden that is no 
longer required.
    (2) The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. No changes to the physical design and operation of the 
plant will occur as a result of this amendment. The processes by 
which activities are planned, prioritized, and controlled are not 
affected. The appropriate level of technical review and management 
oversight will continue to be performed in accordance with existing 
procedures and practices to ensure activities are performed in a 
manner consistent with plant safety.
    (3) The proposed amendment does not involve a significant 
reduction in a margin of safety. As stated earlier, no changes to 
the physical design and/or operation of any plant systems will occur 
as a result of this amendment; therefore, there is no reduction in 
any margins of safety. Work activities will continue to receive the 
appropriate technical review and management oversight to ensure 
activities are prioritized and performed in a manner consistent with 
plant safety. The proposed amendment removes an administrative 
burden that is no longer required.

    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: W. S. Stowe, Esquire, Entergy Nuclear 
Generation Company, 800 Boylston Street, 36th Floor, Boston, 
Massachusetts 02199.
    NRC Section Chief: James W. Clifford.

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

    Dated of amendment request: November 29, 1999
    Description of amendment request: The proposed amendment would 
relocate the requirements associated with the high-steam-generator-
level trip functions of the Reactor Protective System from the 
Technical Specifications to the Technical Requirements Manual.
    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 Steam Generator Level--High function of the RPS [Reactor 
Protection System] is not credited in any accident analyses nor does 
it correspond to any TS [Technical Specification] Safety Limit. The 
high-level function acts to protect the Main Turbine from excessive 
moisture carryover during feedwater transient events. Protection of 
the Main Turbine is not required to adequately assure continued 
reactor safety or the health and safety of the public. Although this 
function may also serve to limit water intrusion into the main steam 
lines and consequential overcooling events, its role in this 
capacity is insignificant, as it does not directly act to secure 
feedwater from the steam generators. This Steam Generator Level--
High function acts only to isolate the Main Turbine from the steam 
generators by causing a reactor trip, which in turn actuates a 
turbine trip. This function does not meet any of the criterions 
listed in 10 CFR 50.36(c)(2) (ii) for inclusion into the technical 
specifications for ANO-2 [Arkansas Nuclear One, Unit 2], and, 
therefore, may be excluded from the TSs. Since no changes are made 
that affect the current operation of this function during its 
relocation to the ANO-2 TRM [Technical Requirements Manual], and 
because this function is not credited in any accident analyses, no 
increase in the probability or consequences of an accident 
previously evaluated is evident.
    Therefore, the proposed changes do 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 changes relocate affected TS requirements 
associated with the Steam Generator Level--High Functions of the RPS 
from the ANO-2 TSs to the ANO-2 TRM. Future revisions to the 
setpoints and values associated with this function will be 
established within the requirements of 10 CFR 50.59 to ensure that 
excessive moisture carryover is prevented in order to protect the 
turbine and steam line loads. The Steam Generator Level--High Trip 
setpoint is not credited in any accident analyses and performs only 
an equipment protection function. The setpoint continues to protect 
the Main Turbine from damage and preserves operating margin to 
accommodate excessive feedwater flow prior to trip.
    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 Steam Generator Level--High Trip setpoint is not credited in 
any accident analyses and performs only an equipment protection 
function. The setpoint continues to protect the Main Turbine from 
damage and preserves operating margin to accommodate excessive 
feedwater flow prior to trip. In addition, turbine failure has been 
previously evaluated at ANO-2 as not to be a significant threat to 
the health and safety of the public.

[[Page 6405]]

Events that may result from water intrusion into the main steam 
lines have been previously evaluated and found not to rely upon the 
Steam Generator Level--High Trip function. The relocation of the 
requirements associated with the Steam Generator Level--High 
function from the TSs to the ANO-2 TRM does not change the current 
values and requirements. Since no technical change in the setpoint 
or allowable value is proposed by this submittal and because the 
Steam Generator Level--High function does not meet any of the four 
criterion of 10 CFR 50.36(c)(2)(ii), no significant change to the 
margin of safety is evident.
    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

    Dated of amendment request: November 29, 1999 .
    Description of amendment request: The proposed amendment would 
revise selected Technical Specifications (TSs), Bases, and portions of 
the Safety Analysis Report (SAR) to maintain consistency with the 
transient and accident analyses which evaluated the impact of the 
replacement steam generators (SGs) that are being used for Cycle 15 
operation. TS changes are proposed for the Reactor Protection System 
(RPS) and Engineered Safety Features Actuation System (ESFAS) low 
pressurizer pressure setpoints, the RPS and ESFAS low SG pressure 
setpoints, the RPS and ESFAS low SG level setpoints, the reactor 
coolant flow rate limit, and the high linear power trip setpoints with 
inoperable main steam safety valves (MSSVs). SAR changes would support 
the new TS values and would also include small increases in calculated 
offsite radiological doses using newer, more conservative methods, for 
some non-loss-of-coolant accident events. The doses would remain within 
the 10 CFR Part 100 acceptance criteria. The proposed amendment would 
also make changes to the TSs and Bases that are not directly related to 
the replacement SGs. These changes would revise the allowed outage time 
of the MSSVs in Modes 1 and 2 to allow up to 12 hours to reduce the 
high linear power level-high trip setpoint when one or more MSSVs are 
inoperable, and would revise the action statement in Mode 3 to maintain 
at least two MSSVs operable on each SG.
    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 proposed changes to the ANO-2 [Arkansas Nuclear One, Unit 2] 
TSs are analytically based which change setpoints and procedure 
limits. No physical modifications are required as a result of the 
proposed changes. The RPS/ESFAS setpoint changes provide 
functionally equivalent protection with the RSGs [replacement steam 
generators] as the previous setpoint values provided with the OSGs 
[original steam generators]. Proposed changes in regard to RCS 
[reactor coolant system] flow rate and High Linear Power Trip 
setpoints associated with conditions where MSSVs are inoperable 
represent appropriate restrictions that have resulted from the 
various analyses performed in support of RSG installation. An 
Emergency Core Cooling System (ECCS) performance analysis was 
performed to demonstrate conformance to 10 CFR 50.46 for operation 
with RSGs. For the large break Loss of Coolant Accident (LOCA), the 
most limiting single failure of the ECCS [is no failure to the 
ECCS]. The small break LOCA analysis was reanalyzed using the 
existing Supplement 2 Model (S2M) of the ABB CENP [ABB Combustion 
Engineering Nuclear Power] small break LOCA evaluation model. The 
analysis was performed for 0.03 ft\2\, 0.04 ft\2\, and 0.05 ft\2\ in 
the reactor coolant pump (RCP) discharge leg. The results of both 
analyses demonstrate continued conformance to the ECCS acceptance 
criteria of 10 CFR 50.46. Non-LOCA analyses intended to confirm the 
Chapter 15 events in the ANO-2 SAR were also performed. The analyses 
were performed considering the proposed Safety Limits and the 
Limiting Safety Settings of the TSs and were confirmed to be 
bounding for the affected safety analyses. The results of the non-
LOCA analyses indicate that operation with the RSGs in service is 
acceptable. As a result of the analyses and evaluations performed in 
support of the RSGs, the ANO specific safety parameters and 
regulatory limits are protected. Therefore, the proposed TS changes 
will not significantly increase the probability of an accident 
previously analyzed.
    Loss of Coolant Accidents (LOCAs) and non-LOCA safety analyses 
supporting the proposed changes have been performed and have 
demonstrated conformance with all applicable Licensing Basis 
acceptance criteria. Although calculated radiological doses using 
newer, more conservative methods increase for some non-LOCA events 
(requiring a revision to Chapter 15 of the SAR), the results are 
within the acceptance criteria of 10 CFR 100. Therefore, the 
proposed changes do not involve a significant increase in the 
consequences of an accident previously evaluated.
    Therefore, the proposed changes do 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 changes to the ANO-2 TSs are analytically based and 
require changing plant setpoints and procedural limits. No physical 
modifications are required as a result of the proposed changes. The 
RPS/ESFAS setpoint changes provide functionally equivalent 
protection with the RSGs as the previous setpoint values provided 
with the OSGs. Proposed changes in regard to RCS flow rate and High 
Linear Power Trip setpoints associated with conditions where MSSVs 
are inoperable represent appropriate restrictions that have resulted 
from the various analyses performed in support of RSG installation. 
The additional 8 hours provided for reducing the High Linear Power 
Level trip setpoints is acceptable due to the low probability of an 
event occurring within this period, based on operating experience 
which indicates such a time period is reasonable to complete the 
changes, and to provide consistency with the RSTS [Revised Standard 
Technical Specifications]. Therefore, the proposed TS changes will 
not create the possibility of a new or different kind of accident 
than previously analyzed.
    A review of both LOCA and non-LOCA events was performed which 
confirms that existing licensing basis methodologies have been 
considered and that a new accident event has not been created.
    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

    LOCA and non-LOCA safety analyses supporting the proposed 
changes have been performed and have demonstrated conformance within 
applicable acceptance criteria. With the increased size of the RSGs 
and the change in design characteristics, the bases for the 
setpoints in the ANO-2 TSs are affected. However, based on the new 
analyses and evaluations conducted in support of this license 
amendment, the new TS setpoints provide adequate margin to protect 
established safety and regulatory limits. Although calculated 
offsite radiological doses increase slightly for some non-LOCA 
events documented in Chapter 15 of the ANO-2 SAR, the increases are 
not considered to be significant in that the results remain within 
the 10 CFR 100 acceptance criteria.
    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 6406]]

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. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Dated of amendment request: July 29, 1999, as supplemented by 
letters dated August 8 and August 24, 1999 (NPF-38-220).
    Description of amendment request: The proposed change modifies 
Technical Specifications (TS) 3.8.1.1 and associated Bases by extending 
the Emergency Diesel Generator allowed outage time from 72 hours to ten 
days. Additionally, this proposed change adds Section 6.16, 
``Configuration Risk Management Program'' to the Administrative 
Controls of the TS.
    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. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: The emergency diesel generators (EDGs) are backup 
alternating current power sources designed to power essential safety 
systems in the event of a loss of offsite power. EDGs are not an 
accident initiator in any accident previously evaluated. Therefore, 
this change does not involve an increase in the probability of an 
accident previously evaluated.
    The EDGs provide backup power to components that mitigate the 
consequences of accidents. The proposed changes to allowed outage 
times (AOTs) do not affect any of the assumptions used in 
deterministic safety analyses.
    In order to fully evaluate the EDG AOT extension, probabilistic 
safety analysis methods were utilized. The results of these analyses 
indicate no significant increase in the risk of an accident 
previously evaluated. These analyses are detailed in CE NPSD-996, 
Combustion Engineering Owners Group ``Joint Applications Report for 
Emergency Diesel Generators AOT Extension.''
    The Configuration Risk Management Program is an Administrative 
Program that assesses risk based on plant status. Adding the 
requirement to implement this program for Technical Specification 
3.8.1.1 ACTION b does not affect the probability or the consequences 
of an accident.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: The proposed change does not change the design or 
configuration of the plant. No new method of plant operation is 
involved.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: The proposed changes do not affect the Technical 
Specification limiting conditions for operation or their bases which 
support the deterministic analyses used to establish the margin of 
safety. Evaluations used to support the requested Technical 
Specification changes have been demonstrated to be either risk 
neutral or risk beneficial depending on precise plant conditions. 
These evaluations are detailed in CE NPSD-996.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.

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

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Dated of amendment request: July 29, 1999, as supplemented by 
letter dated August 24, 1999 (NPF-38-221).
    Description of amendment request: The proposed change modifies 
Technical Specifications (TS) 3.6.2.1 to extend the allowable outage 
time to seven days for one Containment Spray System (CSS) train 
inoperable. A new ACTION has been added to provide a shutdown 
requirement for the inoperability of two CSSs. Additionally, the 
APPLICABILITY is being changed to provide an end state of MODE 4. 
Associated TS Bases changes are included.
    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. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: The Containment Spray System (CSS) is part of the 
Containment Depressurization and Cooling System. Inoperable CSS 
components are not accident initiators in any accident previously 
evaluated. Therefore, this change does not involve an increase in 
the probability of any accident previously evaluated.
    The CSS system is primarily designed to mitigate the 
consequences of a Loss of Coolant Accident (LOCA) or Main Steam Line 
Break (MSLB). These proposed changes do not affect any of the 
assumptions used in the deterministic LOCA or MSLB analyses. Hence 
the consequences of accidents previously evaluated do not change.
    In order to fully evaluate the CSS AOT [Allowed Outage Time] 
extension, probabilistic safety assessment (PSA) methods were 
utilized. The results of these analyses show no significant increase 
in the core damage frequency. These analyses are detailed in report 
CE NPSD-1045, ``Modifications To The Containment Spray System, and 
Low Pressure Safety Injection System Technical Specifications.''
    The Configuration Risk Management Program is an Administrative 
Program that assesses risk based on plant status. Adding the 
requirement to implement this program for Technical Specification 
3.6.2.1 does not affect the probability or the consequences of an 
accident.
    Analyzed events are assumed to be initiated by the failure of 
plant structures, systems or components. Allowing an extended AOT or 
changing the APPLICABILITY does not increase the probability that a 
failure leading to an analyzed event will occur. The CSS components 
are passive until an actuation signal is generated. This change does 
not increase the failure probability of the CSS components. As such, 
the probability of occurrence for a previously analyzed accident 
[is] not significantly increased.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: The proposed change does not change the design or 
configuration of the plant. No new equipment is being introduced, 
and installed equipment is not being operated in a new or different 
manner. There is no change being made to the parameters within which 
the plant is operated, and the setpoints at which protective or 
mitigative actions are initiated are unaffected by this change. No 
alteration in the procedures which ensure the plant remains within 
analyzed limits is being proposed, and no change is being made to 
the procedures relied upon to respond to an off-normal event. As 
such, no new failure modes are being introduced. The proposed change

[[Page 6407]]

will only provide the plant some flexibility in the AOT and 
chang[es] the APPLICABILITY. The change does not alter assumptions 
made in the safety analysis and licensing basis. Therefore, the 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: The proposed changes do not affect the limiting 
conditions for operation or their bases used in the deterministic 
analysis to establish the margin of safety. PSA evaluations were 
used to evaluate these changes. These evaluations demonstrate that 
the changes involve no significant increase in risk. These 
evaluations are detailed in report CE NPSD-1045. The margin of 
safety is established through equipment design, operating 
parameters, and the setpoints at which automatic actions are 
initiated. None of these are adversely impacted by the proposed 
change. Sufficient equipment remains available to actuate upon 
demand for the purpose of mitigating a transient event. The proposed 
change, which allows operation to continue for up to 7 days with 
components inoperable in one CSS train, is acceptable based on the 
remaining CSS components providing 100% of the required CSS flow. 
The reduced potential for a self-induced plant transient resulting 
from unit shutdown required for a second inoperable CSS train is 
minimized. Therefore, the change does not involve a significant 
reduction in the margin of safety, and is offset by minimizing the 
potential for a self-induced plant transient.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.

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

Entergy Operations Inc., Docket No. 50-382, Waterford Steam 
Electric Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: October 18, 1999.
    Description of amendment request: The proposed change modifies 
Technical Specification (TS) 3.6.2.2 Limiting Condition for 
Operation to allow Waterford Steam Electric Station, Unit 3 to 
operate with two independent trains of containment cooling, 
consisting of one cooler per train, operable during modes 1, 2, 3, 
and 4. Associated changes to the TS Bases have been proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will the operation of the facility in accordance with these 
proposed changes involve a significant increase in the probability 
or consequence of an accident previously evaluated?
    Response: The proposed change to Technical Specification (TS) 
3.6.2.2 reduces the number of Containment Fan Coolers (CFC) from two 
to one required to be operable in each train of the Containment 
Cooling System for modes 1, 2, 3, and 4. This change does not create 
any new system interactions and has no impact on operation or 
function of any system or equipment in a way that could cause an 
accident. The CFCs are not an initiator of any events nor affect any 
accident initiators of any events analyzed in Chapter 15 of the 
UFSAR [Updated Final Safety Analysis Report]. Therefore this change 
will not impact the probability of occurrence of an accident.
    The results of the reanalysis of the limiting Loss of Coolant 
Accident (LOCA) and Main Steam Line Break (MSLB) accidents show that 
the consequences of an accident previously evaluated are not 
increased by the change in the required number of operable CFCs. The 
limiting accidents affected by the proposed changes are identified 
below:
    The peak containment pressure following the limiting LOCA 
(Double Ended Hot Leg Slot Break with minimum safety injection flow) 
was determined to be 35.2 psig [pounds per square inch, gauge] as 
compared to the current licensing basis limiting LOCA (Double Ended 
Suction Leg Slot Break with minimum safety injection flow) peak 
pressure of 43.1 psig.
    The peak containment pressure at 24 hours following the start of 
the limiting LOCA (Double Ended Discharge Leg Slot Break with 
minimum safety injection flow) and the operation of one containment 
spray train and one partially flooded CFC operable was determined to 
be 15.5 psig with a peak pressure of 33.27 psig as compared to the 
current licensing basis of 14.9 psig with a peak pressure of 42.9 
psig. The current licensing basis limiting LOCA is the Double Ended 
Suction Leg Slot Break with maximum safety injection flow and the 
operation of one containment spray train and two operable CFCs.
    The peak containment pressure following the limiting MSLB (102% 
power with failure of one containment heat removal train consisting 
of one containment spray pump and one CFC operable) was determined 
to be 42.68 psig as compared to the current licensing basis peak 
pressure of 42.9 psig. The current licensing basis limiting MSLB is 
75% power with the failure of one train of containment heat removal 
system consisting of one containment spray train and two operable 
CFCs.
    The peak containment equipment qualification temperature 
following the limiting MSLB (102% power with the failure of one MSIV 
to close) was determined to be 397.4  deg.F as compared to the 
current licensing basis peak temperature of 409.1  deg.F. The 
current licensing basis limiting MSLB is 102% power with two CFCs 
per train operable and the failure of one train of containment 
spray.
    These values above demonstrate that the containment design basis 
pressure and equipment qualification temperature of 44 psig and 
413.5  deg.F, respectively, are not exceeded and the containment 
pressure at 24 hours after start of the limiting LOCA is less than 
50% of the peak pressure.
    The results of the containment response analysis discussed above 
satisfy the following NRC Staff Standard Review Plan (SRP) section 
6.2.1.1.A guidance document acceptance criteria for a PWR 
[Pressurized Water Reactor] dry containment.
    The peak calculated containment pressure following a Loss of 
Coolant Accident (LOCA) or Main Steam Line Break (MSLB) should be 
less than the containment design pressure.
    To satisfy the requirements of GDC [General Design Criteria] 38 
to rapidly reduce the containment pressure, the containment pressure 
should be reduced to less than 50% of the peak calculated pressure 
for the design basis LOCA within 24 hours after the postulated 
accident.
    Thus, revising the containment cooling system TS to require only 
one operable CFC per train results in acceptable containment 
response and therefore, will not adversely impact the consequences 
of accidents previously evaluated.
    Therefore, the proposed changes will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Will the operation of the facility in accordance with these 
proposed changes create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    Response: The proposed change to reduce the number of operable 
Containment Fan Coolers (CFC) from two to one in each train of the 
Containment Cooling System for modes 1, 2, 3, and 4 does not alter 
the operation of the CFCs. Although only one of the two CFCs per 
train is required to be operable, the manner in which the CFCs 
perform their safety function is not changed. All four CFCs (two per 
train) will be maintained operable to the extent possible to provide 
the greatest defense in depth and operating flexibility.
    This proposed change does not involve a change in plant design, 
nor does it involve any potential initiating events that would 
create any new or different kind of accident. This proposed change 
does not alter the way in which the plant is operated in a manner 
that would create a new or different accident. Therefore, since no 
hardware modifications will be made, the proposed change will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Will the operation of the facility in accordance with these 
proposed changes involve a significant reduction in a margin of 
safety?
    Response: The proposed change revises TS 3.6.2.2, Containment 
Cooling System. This change revises the required number of fan 
coolers from two fan coolers per train to one fan cooler per train. 
As described in the containment depressurization and cooling

[[Page 6408]]

system Technical Specification Bases, the containment cooling system 
is designed to maintain the post accident containment peak pressure 
below its design value of 44 psig. The system is also designed to 
reduce the containment pressure by a factor of [two] from its post-
accident peak within 24 hours.
    The analyses that have been performed to support this Technical 
Specification change have shown that the peak containment pressure 
remains below 44 psig, the 24-hour containment pressure is less than 
half the peak pressure, and the containment peak temperature remains 
below the maximum temperature of 413.5  deg.F provided in the Bases 
for Technical Specifications 3.6.2.1 and 3.6.2.2. In comparison of 
the current safety margins to the safety margins that would exist if 
the proposed changes were in effect, the results of the analyses, 
illustrated below, show an increase in the margin of safety for 
containment pressure and equipment qualification temperature 
following the associated limiting LOCA and MSLB.
    The peak containment pressure following the limiting LOCA was 
determined to be 35.2 psig as compared to the current licensing 
basis limiting LOCA peak pressure of 43.1 psig.
    The peak containment pressure at 24 hours following the start of 
the limiting LOCA was determined to be 15.5 psig with a peak 
pressure of 33.27 psig as compared to the current licensing basis of 
14.9 psig with a peak pressure of 42.9 psig.
    The peak containment pressure following the limiting MSLB was 
determined to be 42.68 psig as compared to the current licensing 
basis peak pressure of 42.9 psig.
    The peak containment equipment qualification temperature 
following the limiting MSLB was determined to be 397.4  deg.F as 
compared to the current licensing basis peak temperature of 409.1 
deg.F.
    This proposed change does not adversely impact a margin of 
safety, involve a change in plant design, or have any affect on the 
plant protective barriers. Therefore, the proposed change will not 
involve a significant reduction in the margin of safety.

    The Nuclear Regulatory Commission 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: N. S. Reynolds, Esquire, Winston & Strawn 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: December 13, 1999.
    Description of amendment request: The licensee proposes to change 
the license to delete an expired license condition and to make some 
editorial and administrative changes to correct or clarify the license.
    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 changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. In addition, the proposed changes do not 
affect the manner in which the plant responds in normal operation, 
transient or accident conditions nor do they change any of the 
procedures related to operations of the plant. The proposed changes 
do not alter or prevent the ability of structures, systems and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the acceptance limits 
assumed in the Updated Final Safety Analysis Report (UFSAR). The 
proposed changes are administrative and editorial in nature and only 
correct, update and modify the Operation License.
    The proposed changes do not affect the source term, containment 
isolation or radiological release assumptions used in evaluating the 
radiological consequences of an accident previously evaluated in the 
Seabrook Station UFSAR. Further, the proposed changes do not 
increase the types and amounts of radioactive effluent that may be 
released offsite, nor significantly increase individual or 
cumulative occupational/public radiation exposures.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes are administrative in nature and only 
correct, update and clarify the Seabrook Station Operating License. 
The proposed changes do not modify the facility nor do they modify 
the manner in which the plant will be operated nor do they affect 
the plant's response to normal, transient or accident conditions. 
The changes do not introduce a new mode of plant operation. The 
plant's design basis are not revised and the current safety analyses 
will remain in effect and the plant will continue to be operated in 
accordance with the existing Technical Specifications. Therefore, 
the proposed changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes are administrative and editorial changes to 
the Seabrook Station Operating License that do not revise the 
Technical Specifications or the bases for the Technical 
Specifications. The safety margins established through Limiting 
Conditions for Operation, Limiting Safety System Settings and Safety 
Limits as specified in the Technical Specifications are not revised 
nor is the plant design or its method of operation revised by the 
proposed changes. Since there will be no changes to the physical 
design or operation of the plant, 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 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.

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

    Date of amendment request: March 18, 1998.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications 2.15(4) and 2.15(5) to identify (1) 
all indication functions and control functions required for the 
alternate (remote) shutdown system (alternate shutdown panel and 
auxiliary feedwater panel), (2) panel locations of the functions, and 
(3) the number of channels required.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes to Technical Specifications Section 2.15(4) 
and 2.15(5) identify functions, instruments, and controls along with 
their location and the number of required channels. The new 
Technical Specifications section addresses the regulatory 
requirements for equipment required for Alternative and Dedicated 
Shutdown Capability per 10 CFR Part 50 Appendix R. It will ensure 
that proper Limiting Conditions for Operation are entered for 
equipment or functional inoperability. There are no physical 
alterations being made to the Alternate Shutdown Panel and the 
Auxiliary Feedwater Panel or related systems. Therefore, the 
proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

[[Page 6409]]

    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes will not result in any physical alterations 
to the Alternate Shutdown Panel or the Auxiliary Feedwater Panel, or 
any plant configuration, systems, equipment, or operational 
characteristics. There will be no changes in operating modes, or 
safety limits, or instrument limits. With the proposed changes in 
place, Technical Specifications retain requirements for the 
Alternate Shutdown Panel and the Auxiliary Feedwater Panel. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes clarify the regulatory requirements for the 
Alternative and Dedicated Shutdown Capability as defined by 10 CFR 
Part 50, Appendix R. The proposed changes will not alter any 
physical or operational characteristics of the Alternate Shutdown 
Panel and the Auxiliary Feedwater Panel and their associated systems 
and equipment. Therefore, the proposed changes do not involve a 
reduction in a margin of safety.

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

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 amendment request: November 29, 1999.
    Description of amendment request: The amendment would adopt the 
``Standard Test Method for Nuclear Grade Activated Carbon'' for 
charcoal filter laboratory testing with certain exceptions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: The proposed license amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. The adoption of the new test method 
and acceptance criteria of ASTM [American Society for Testing and 
Materials] D3803-1989, with the exceptions as identified in the 
Technical Specifications, for activated charcoal filters does not 
involve any modifications to the plant, will not require changes to 
how the plant is operated nor will it affect the operation of the 
plant. Adoption of these provisions ensures compliance with the new 
test standard of ASTM D 3803-1989. Adoption of new test method will 
not cause an accident and therefore cannot involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response: The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. The adoption of the new test method and 
acceptance criteria of ASTM D 3803-1989, with the exceptions as 
identified in the Technical Specifications, for activated charcoal 
filters does not involve any modifications to the plant, will not 
require changes to how the plant is operated nor will it affect the 
operation of the plant. Adoption of new test method will not cause 
an accident and therefore cannot create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Does the proposed license amendment involve a significant 
reduction in a margin of safety?
    Response: The proposed license amendment does not involve a 
significant reduction in a margin of safety. The use of outdated 
test protocols or inappropriate test conditions can lead to an 
overestimation of the charcoal filters' ability to adsorb 
radioiodine following an accident. The adoption of the new test 
method and testing criteria of ASTM D 3803-1989, with the exceptions 
as identified in the Technical Specifications, for activated 
charcoal filters would ensure at least a safety factor of two is 
maintained. Thus, the proposed change would not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Section Chief: Marsha Gamberoni, Acting.
Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York
    Date of amendment request: November 19, 1999.
    Description of amendment request: The proposed amendment would 
revise Sections 4.7.A and 4.11.B of the Appendix A Technical 
Specifications (TSs) to the James A. FitzPatrick Operating License to 
adopt the surveillance test methods and performance criteria detailed 
in NRC Generic Letter 99-02 for laboratory testing of nuclear-grade 
charcoal. The proposed amendment also would reduce the minimum 
allowable Standby Gas Treatment System (SGTS) and Control Room 
Emergency Ventilation Air Ventilation Supply System (CREVASS) charcoal 
filter efficiencies specified in the TSs to those assumed in the 
updated radiological dose calculations.
    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:

    Operation of the FitzPatrick plant in accordance with the 
proposed amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92(c) since it would not:
    (1) Involve an increase in the probability or consequences of an 
accident previously evaluated.
    These changes are not modifications to the plant. They will not 
require changes to how the plant is operated, nor will they affect 
the operation of the plant.
    Changes to these test methods will not cause an accident and 
therefore cannot increase the probability of an accident.
    Calculated radiological doses increase as a result of reductions 
in assumed charcoal efficiencies, but remain within regulatory 
limits. Radiological doses at the site boundary and low population 
zone are less than 25 percent of 10 CFR 100 criteria. Post-accident 
doses to control room operators are also within regulatory limits.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    These changes are not modifications to the plant, nor will they 
require changes to how the plant is operated. Changes to these test 
methods will not cause an accident, and therefore cannot create the 
possibility of a new or different kind of accident.
    (3) Involve a significant reduction in a margin of safety.
    Evidence has been presented which contend that activated 
charcoal testing performed to test standards other than ASTM 
[American Society for Testing and Materials] D3803-1989 may be 
inaccurate and may overestimate its adsorption capabilities. The 
adoption of ASTM D3803-1989 charcoal performance standards and 
plant-specific test parameters ensures adequate safety margins.

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

[[Page 6410]]

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: Marsha Gamberoni, Acting.

Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: December 20, 1999.
    Description of amendment request: This application for an amendment 
to the James A. FitzPatrick Technical Specifications (TS) proposes a 
change to the Main Steam Isolation Valve (MSIVs) closure scram 
setpoint. The proposed amendment changes the MSIV closure scram Trip 
Level Setting from <ls-thn-eq>10% to <ls-thn-eq>15% valve closure.
    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 will not significantly increase the 
probability or consequences of any previously evaluated accidents.
    This proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
This proposed change to the MSIV scram trip setpoint from 90% open 
to 85% open results in the MSIV closure scram event having a slight 
delay in the initiation of the reactor scram. The MSIV scram event 
is not an accident initiator and will not increase the probability 
or consequence of any accident previously evaluated. An evaluation 
of events concluded that the MSIV direct scram remains a relatively 
low consequence event and has no effect on any operating limits. The 
limiting event analyzed for the reactor vessel overpressure event in 
the James A. FitzPatrick Updated Final Safety Analysis Report 
(UFSAR) is the MSIV closure terminated by a high neutron flux scram, 
which does not take credit for the MSIV closure valve position 
scram. In addition, an evaluation of the Main Steam Line Break 
outside containment event concludes that there is no impact on PCT 
[Peak Clad Temperature] or break flow and that the results presented 
in the UFSAR are bounding. An evaluation of the impact on 
containment was also made. The containment response is evaluated for 
much more severe events such as a Loss of Coolant Accident (LOCA) or 
stuck open Safety Relief Valve (SRV), thus the change in MSIV scram 
setpoint has no impact on the containment analysis. Therefore, 
changing the MSIV closure scram setpoint from 90% open to 85% open 
does not significantly increase the probability or consequences of 
any previously evaluated accidents.
    2. The proposed change will not create the possibility of a new 
or different kind of accident.
    This proposed change to the MSIV scram trip setpoint from 90% 
open to 85% open results in the MSIV closure scram event having a 
slight delay in the initiation of the reactor scram, and does not 
introduce a new or different kind of accident previously analyzed. 
An evaluation of the event determined that the MSIV closure scram 
remains a relatively low consequence event and has no effect on any 
operating limits. The limiting event analyzed for the reactor vessel 
overpressure event in the UFSAR is the MSIV closure terminated by a 
high neutron flux scram, which does not take credit for the MSIV 
closure valve position scram. The proposed change will not create 
the possibility of a new or different kind of accident.
    3. The proposed change will not involve a significant reduction 
in a margin of safety.
    This proposed change to the MSIV scram trip setpoint from 90% 
open to 85% open results in the MSIV closure scram event having a 
slight delay in the initiation of the reactor scram. An evaluation 
of the event determined that the MSIV closure scram remains a 
relatively low consequence event and has no effect on any operating 
limits. In addition, an evaluation of the Main Steam Line Break 
outside containment event concludes that there is no impact on PCT 
or break flow and that the results presented in the UFSAR are 
bounding. An evaluation of the impact on containment was also made. 
The containment response is evaluated for much more severe events 
such as a LOCA or stuck open SRV, thus the change in MSIV scram 
setpoint has no impact on the containment analysis. Changing the 
MSIV valve position scram setpoint from <ls-thn-eq>10% to 
<ls-thn-eq>15% of valve closure will allow the limit switches to be 
positioned such that both scram and indicating limit switches can be 
coordinated and provide accurate and reliable valve position 
indication in the control room. Therefore, changing the MSIV closure 
scram setpoint from 90% open to 85% open does not involve a 
significant reduction in a margin of safety and provides a net 
benefit to plant operations.
    The proposed change will not increase the probability or 
consequences of any previously analyzed accident, introduce any new 
or different kind of accident previously evaluated, or significantly 
reduce existing margin to safety. Therefore, the proposed license 
amendment will not involve a significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: Marsha Gamberoni, Acting.

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: August 30, 1999.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 5.2.2 in order to raise the level of the 
approval authority for deviations from the guidelines provided to 
minimize unit staff overtime.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed license amendment would strengthen the 
administrative controls that permit plant personnel to work beyond 
those limits outlined in the TS's. As a result, there will be 
greater scrutiny on the amount of overtime being utilized to perform 
safety-related function. Therefore, it has been determined that 
operation of the facility in accordance with the proposed amendment 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    There will be no physical changes to the systems, components or 
structure of the facility as a result of this proposed license 
amendment. The initial assumptions of the design accident analyses 
will be unaffected. Therefore, operation of the facility in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    This amendment raises the administrative level of management 
approval required for overtime in excess of the limits outlined in 
the TS. Therefore, operation of the facility in accordance with the 
proposed amendment will not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,

[[Page 6411]]

NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: Richard L. Emch, Jr.

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: October 4, 1999.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 5.5.6, ``Prestressed Concrete 
Containment Tendon Surveillance Program,'' to incorporate three 
exceptions to Regulatory Guide (RG) 1.35, Revision 2, 1976. The 
exceptions concern the number of tendons detensioned, inspection of 
concrete adjacent to vertical tendons, and the time during which areas 
adjacent to tendons are inspected.
    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?
    No. The proposed change only clarifies TS requirements for the 
containment tendon surveillance program. The proposed clarification 
has been previously reviewed and approved by the NRC staff with 
Amendments 23 and 4, and is consistent with current regulatory 
guidance. As such, the proposed change is essentially administrative 
in nature. The containment tendon surveillance program has no impact 
on the probability of any accident initiators, and it will continue 
to ensure containment structural integrity. Therefore, the proposed 
change does 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 previously evaluated?
    No. The proposed change only clarifies TS requirements for the 
containment tendon surveillance program. The proposed clarification 
has been previously reviewed and approved by the NRC staff with 
Amendments 23 and 4, and is consistent with current regulatory 
guidance. As such, the proposed change is essentially administrative 
in nature. Plant design and operation will not be changed, and no 
other safety related or important to safety equipment is affected by 
the proposed change. Therefore, the proposed change will not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety.
    No. The proposed change only clarifies TS requirements for the 
containment tendon surveillance program. The proposed clarification 
has been previously reviewed and approved by the NRC staff with 
Amendments 23 and 4, and is consistent with current regulatory 
guidance. As such, the proposed change is essentially administrative 
in nature. The containment prestressing system will continue to 
perform its function to ensure containment structural integrity. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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 No. 50-390 Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of amendment request: June 25, 1999 (TS 98-016).
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) and TS Bases to reflect 
application of the Westinghouse generic Best Estimate Large Break Loss-
of-Coolant Accident Analysis methodology using the WCOBRA/TRAC computer 
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, 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 changes involve use of the Best Estimate Large 
Break loss-of-coolant accident (LOCA) analysis methodology and 
associated technical specification changes. Accumulator water level 
set points will be revised from [greater than or equal to] 7717 
gallons and [less than or equal to] 8004 gallons to [greater than or 
equal to] 7630 gallons and [less than or equal to] 8000 gallons to 
provide the plant with an increased operating range. The plant 
conditions assumed in the analysis, including the accumulator water 
level instrumentation changes, are bounded by the design conditions 
for all equipment in the plant.
    Therefore, there will be no increase in the probability of a 
LOCA. The consequences of a LOCA are not being increased, since it 
is shown that the emergency core cooling system (ECCS) is designed 
so that its calculated cooling performance conforms to the criteria 
contained in 10 CFR 50.46, Paragraph b. The small break LOCA 
analysis assumes only a nominal accumulator water level which is the 
same nominal value assumed in this analysis, therefore, the small 
break LOCA analysis is unaffected by the increase in the accumulator 
range. Also, the increased safety analysis range in accumulator 
water volume (+/-15 cubic feet) has an insignificant effect on the 
containment related analyses.
    The post-LOCA containment sump boron calculation assumes a 
minimum accumulator volume which bounds (is smaller than) the 1005 
cubic feet (7518 gallons) value supported by the Best Estimate Large 
Break LOCA analysis. Also, the hot leg switchover calculation models 
a maximum accumulator volume which is not bounded by the 1095 cubic 
feet (8191 gallons) maximum value supported by the Best Estimate 
Large Break LOCA analysis. However, an evaluation concludes that the 
Watts Bar hot leg switchover time is unaffected by the difference in 
maximum volumes.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    No new modes of plant operation are being introduced by the new 
analysis or by the changes in instrumentation setpoints for 
accumulator water level. The parameters assumed in the analysis are 
within the design limits of existing plant equipment. Therefore, the 
proposed change does not 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.
    It has been shown that the analytic technique used in the 
analysis realistically describes the expected behavior of the WBN 
Unit 1 reactor system during a postulated loss of coolant accident. 
Uncertainties have been accounted for as required by 10 CFR 50.46. 
The physical setpoint changes to accumulator water level 
instrumentation are bounded by the uncertainty evaluation addressing 
accumulator water level. A sufficient number of loss of coolant 
accidents with different break sizes, different locations, and other 
variations in properties have been considered to provide assurance 
that the most severe postulated LOCAs were evaluated. It has been 
shown by the analysis that there is a high level of probability that 
all criteria contained in 10 CFR 50.46, Paragraph b are met.

    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,

[[Page 6412]]

400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: December 14, 1999.
    Description of amendment request: This proposed change would revise 
the Vermont Yankee (VY) Technical Specifications (TS) by relocating the 
procedural details of the Radiological Effluent Technical 
Specifications (RETS) to the Offsite Dose Calculation Manual (ODCM). 
The TS would also be revised to relocate procedural details associated 
with solid radioactive wastes to the Process Control Program (PCP). In 
addition, the TS definition for ``solidification'' would be relocated 
to the VY Technical Requirements Manual (TRM).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes do not affect accident initiators or 
precursors and do not alter the design assumptions, conditions, 
configuration of the facility, or the manner in which the plant is 
operated. The proposed changes do not alter or prevent the ability 
of structures, systems, or components to 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 are administrative in 
nature and do not change the level of programmatic controls and 
procedural details relative to radiological effluents.
    Implementation of programmatic controls for RETS already in TS 
will assure that the applicable regulatory requirements pertaining 
to the control of radioactive effluents will continue to be 
maintained. 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. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed changes do not alter the design assumptions, 
conditions, configuration of the facility, or the manner in which 
the plant is operated. The proposed changes have no impact on 
component or system interactions. The proposed changes are 
administrative in nature and do not change the level of programmatic 
controls and procedural details relative to radiological effluents.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated for Vermont Yankee.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    There is no impact on 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 as a result 
of the proposed changes. The proposed changes are administrative in 
nature and do not change the level of programmatic controls and 
procedural details relative to radiological effluents. A comparable 
level of administrative controls will continue to be applied to 
those specifications being relocated to the ODCM, PCP, or TRM.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: October 28, 1999, as supplemented 
December 21, 1999.
    Description of amendment request: The proposed changes will 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.
    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 waste gas decay tank rupture is highly unlikely, as the waste 
gas decay tanks are designed and constructed to stringent quality 
control standards, are provided with pressure relief valves to 
prevent overpressurization, are missile-shielded by installation 
below grade, and have their gaseous contents controlled to prevent 
potentially explosive mixtures. The entire gaseous content of the 
waste gas decay tank is assumed to be released to the atmosphere as 
a ground-level release * * *. The waste gas charcoal filter system 
is not credited for any mitigation of the release in the accident 
analysis for a waste gas decay tank rupture. In addition, the 
releases associated with a waste gas decay tank rupture are bounded 
by the existing LOCA [loss-of-coolant accident] releases. 
Specifically, operation of the North Anna Power Station in 
accordance with the proposed Technical Specification changes will 
not:
    [1.] Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Relocating the operability and surveillance requirements for The 
Waste Gas Charcoal System to the TRM [Technical Requirements Manual] 
does not change the operation of the plant. The plant and the 
radioactive gas waste system will not be operated differently. No 
new accident initiators are established as a result of the proposed 
changes. Therefore, the probability of occurrence is not increased 
for any accident previously evaluated.
    Relocating the operability and surveillance requirements for The 
Waste Gas Charcoal Filter to the TRM does not [a]ffect the gaseous 
releases to the environment, which are controlled by the ODCM 
[Offsite Dose Calculation Manual]. Additionally, no credit for these 
filters is taken in the accident analysis for Waste Gas Decay Tank 
rupture. Therefore, there is no increase in the consequences of any 
accident previously analyzed.
    [2.] Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed changes do not affect the operation of the plant. 
The gaseous waste systems will not be operated differently as a 
result of the proposed changes. No new accident or event initiators 
are created moving the operability and surveillance requirements for 
The Waste Gas Charcoal Filter to the TRM. Therefore, the proposed 
changes do not create the possibility of any accident or malfunction 
of a different type.
    [3.] Involve a significant reduction in the margin of safety as 
defined in the bases on any Technical Specifications.
    The proposed changes have no effect on any safety analyses 
assumptions. The waste gas charcoal filters are not used to mitigate 
the consequence[s] [of] a Waste Gas Decay Tank rupture. The accident 
analysis assumes total release of the radioactiv[ity] in the Waste 
Gas Decay Tank in the accident analysis. Therefore, the proposed 
changes do not result in a reduction in a margin of safety.

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

[[Page 6413]]

involves no significant hazards consideration.
    Attorney for licensee: Mr. Donald P. Irwin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Section Chief: Richard L. Emch Jr.

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

    Dated of amendment request: November 29, 1999.
    Description of amendment request: The proposed changes will modify 
the Technical Specifications in Section 4.7.7.1 for the Control Room 
Emergency Habitability System and Section 4.7.8.1 for the Safeguards 
Area Ventilation System. The changes will require laboratory testing of 
the charcoal filter carbon to be consistent with American Society for 
Testing and Materials (ASTM) Standard D3803-1989.
    Basis for proposed no significant hazards consideration 
determination: In 10 CFR 50.92 three criteria are provided to determine 
whether a proposed license amendment involves a significant hazards 
consideration. No significant hazards consideration is involved if 
operation of the facility 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. 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:

    Virginia Electric and Power Company has reviewed the 
requirements of 10 CFR 50.92 as they relate to the proposed changes 
for the North Anna Units 1 and 2 and determined that a significant 
hazards consideration is not involved. The proposed Technical 
Specification changes adopt the nuclear-grade charcoal testing 
requirements of ASTM D3803-1989 and do not affect the design or 
operation of the plant. The changes also do not involve any physical 
modification to the plant or result in a change in a method of 
system operation. The adoption of the 1989 edition of ASTM D3803 
provides assurance that testing of nuclear-grade activated charcoal 
of ventilation systems is being performed with a suitable standard 
to ensure that charcoal adsorbers are capable of performing their 
required safety function and that the regulatory requirements 
regarding onsite and offsite dose consequences continue to be 
satisfied. The changes do not create an unreviewed safety question.
    (a) The proposed changes modify surveillance testing 
requirements and do not affect plant systems or operation and 
therefore do not increase the probability or the consequences of an 
accident previously evaluated. The proposed surveillance 
requirements adopt ASTM D3803-1989 as the laboratory method for 
testing samples of the charcoal adsorber in response to NRC's 
Generic Letter 99-02. This method of testing charcoal adsorbers has 
been approved by the NRC as an acceptable method for determining 
methyl iodide removal efficiency. Since the charcoal adsorbers are 
used to mitigate the consequences of an accident, the more accurate 
the test, the better assurance we have that we remain within our 
accident analysis assumptions. The laboratory test acceptance 
criteria contain a safety factor to ensure that the efficiency 
assumed in the accident analysis is still valid at the end of the 
operating cycle. There is no change in the method of plant 
operation, system performance or system design. (b) The proposed 
changes do not create the possibility of an accident or malfunction 
of a different type than any evaluated previously.
    The proposed changes modify surveillance testing requirements 
and do not impact plant systems or operations and therefore do not 
create the possibility of an accident or malfunction of a different 
type than evaluated previously. The proposed surveillance 
requirements adopt ASTM D3803-1989 as the laboratory method for 
testing samples of the charcoal adsorber. This change is in response 
to NRC's request in response to their Generic Letter 99-02. There is 
no change in the method of plant operation or system design. There 
are no new or different accident scenarios, transient precursors, 
nor failure mechanisms that will be introduced.
    (c) The proposed changes modify surveillance test requirements 
and do not impact plant systems or operations and therefore do not 
significantly reduce the margin of safety. The revised surveillance 
requirements adopt ASTM D3803-1989 as the laboratory method for 
testing samples of the charcoal adsorber. The 1989 edition of this 
standard imposes very stringent requirements for establishing the 
capability of new and used activated carbon to remove radio-labeled 
methyl iodide from air and gas streams. The results of this test 
provide a more conservative estimate of the performance of nuclear-
graded activated carbon used in all nuclear power plant HVAC 
[heating, ventilation, and air conditioning] systems for the removal 
of radioiodine. The laboratory test acceptance criteria contain a 
safety factor to ensure that the efficiency assumed in the accident 
analysis is still valid at the end of the operating cycle.
    This analysis demonstrates that the proposed amendment to The 
North Anna Units 1 and 2 Technical Specifications does not involve a 
significant increase in the probability or consequences of a 
previously evaluated accident, does not create the possibility of a 
new or different kind of accident and 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 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. Donald P. Irwin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Section Chief: Richard L. Emch Jr.

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.

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

    Dated of amendment request: April 11, 1996, as supplemented April 
6, 1998, March 22 and July 29, 1999 (PCN-460).
    Brief description of amendment request: The proposed amendments 
would revise the San Onofre Units 2 and 3 Technical Specification (TS) 
related to the containment isolation valves. Specifically, the licensee 
proposed a revision to TS 3.6.3 to extend the completion times for 
Section D.1 and D.2 valves from 4 hours to the applicable limiting 
condition for operation time pertaining to the engineered safety 
feature system in which the valve is installed.
    Dated of publication of individual notice in Federal Register: 
January 19, 2000 (65 FR 2993), as corrected January 26, 2000 (65 FR 
4265).
    Expiration date of individual notice: February 18, 2000.

[[Page 6414]]

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 amendment request: January 2, 1998, as supplemented 
December 13, 1999 (PCN-482).
    Brief description of amendment request: The proposed amendments 
would revise the San Onofre Units 2 and 3 Technical Specification (TS) 
relating to the Auxiliary Feedwater (AFW) System. Specifically, the 
licensee proposed to revise TS 3.7.5 to add a note that states: ``The 
steam driven AFW pump is OPERABLE when running and controlled manually 
to support plant start-ups, plant shut-downs, and AFW pump and valve 
testing.''
    Date of publication of individual notice in Federal Register: 
January 19, 2000 (65 FR 2991), as corrected January 26, 2000 (65 FR 
4265).
    Expiration date of individual notice: February 18, 2000.

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 amendment request: January 11, 1999, as supplemented 
November 29, 1999 (PCN-499).
    Brief description of amendment request: The proposed amendments 
would revise the San Onofre Units 2 and 3 Technical Specification (TS) 
3.7.6, ``Condensate Storage Tank (CST T-121 and T-120)'' to change the 
minimum inventory of water maintained in the condensate storage tank 
(T-120) from 280,000 gallons to 360,000 gallons during plant operation 
Modes 1, 2 and 3.
    Date of publication of individual notice in Federal Register: 
January 18, 2000 (65 FR 2648).
    Expiration date of individual notice: February 17, 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).

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

    Date of application for amendment: September 28, 1999.
    Brief description of amendment: The amendment revises Technical 
Specification Surveillance Requirement 3.7.6.2 ``Component Cooling 
Water (CCW) System'' to change the CCW pump automatic start actuation 
signal basis from Engineered Safety Feature Actuation Signal to Loss-
of-Power Diesel Generator.
    Date of issuance: January 21, 2000.
    Effective date: January 21, 2000.
    Amendment No.: 186.
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 3, 1999 (64 FR 
59798).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 21, 2000.
    No significant hazards consideration comments received: No.

Consumers Energy Company, Docket No. 50-155, Big Rock Point Plant, 
Charlevoix County, Michigan

    Date of amendment request: May 11, 1999, as supplemented June 3 and 
July 28, 1999.
    Brief description of amendment: This amendment deletes from the 
Defueled Technical Specifications (DTS) subsection 1.16, ``SITE 
BOUNDARY,'' Figure 5.1-1, the Big Rock Point (BRP) Site Map, and DTS 
5.1.1 paragraph numbering and removes certain site-specific information 
from DTS 5.1, which describes the BRP site. The amendment also makes 
editorial changes to DTSs 6.6.2.5. g, h, and j, and 6.6.2.6.b. because 
of the changes associated with DTSs 1.16 and 5.1 and Figure 5.1-1 
described above. Most of the information removed or deleted from the 
DTSs can be found in the BRP Final Hazards Summary Report.
    Date of Issuance: January 13, 2000.
    Effective Date: January 13, 2000, to be implemented within 45 days 
from date of issuance. Implementation includes incorporation of the 
site boundary information, as discussed in the staff's safety 
evaluation enclosed with this amendment, into the next Final Safety 
Analysis Report (i.e. the updated Final Hazards Summary Report for the 
Big Rock Point Nuclear Plant) update in accordance with the schedule in 
10 CFR 50.71(e).
    Amendment No.: 121.
    Facility Operating License No. DPR-6. The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 16, 1999 (64 FR 
32288).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 13, 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: April 5, 1999 as supplemented by 
letters dated May 27, July 6, October 7, and November 22, 1999.
    Brief description of amendments: The amendments revised the 
Technical Specifications to incorporate Topical Report DPC-NE-3005-P, 
``Thermal Hydraulic Transient Analysis Methodology.''
    Date of Issuance: January 18, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.

[[Page 6415]]

    Amendment Nos.: Unit 1-309; Unit 2-309; Unit 3-309.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35202), November 3, 1999 (64 FR 59801).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 18, 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: September 29, 1999.
    Brief description of amendments: The amendments revise the 
Containment Inservice Inspection Program Technical Specifications 
related to the containment leakage testing program and the pre-stressed 
concrete containment tendon surveillance program.
    Date of Issuance: January 18, 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--310; Unit 2--310; Unit 3--310.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: November 17, 1999 (64 
FR 62707).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 18, 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: December 16, 1999.
    Brief description of amendment: The license amendment revises the 
River Bend Station Technical Requirements Manual, Section TR 3.9.14, 
and adds a temporary exception to the current prohibition for travel of 
loads in excess of 1200 pounds over fuel assemblies in the spent fuel 
storage pool. The exception allows the licensee to move the spent fuel 
pool (SFP) watertight gates, which separate the SFP from the cask and 
lower transfer pools, in order to perform repairs on the gates and 
watertight seals prior to the end of Refueling Outage 9. Updated Safety 
Analysis Report (USAR) Sections 9.1.2.2.2 and 9.1.2.3.3 are also 
changed to reflect the proposed exception.
    Date of issuance: January 13, 2000.
    Effective date: The license amendment is effective as of its date 
of issuance and shall be implemented in the next periodic update to the 
USAR and TRM in accordance with 10 CFR 50.71(e). Implementation of the 
amendment is the incorporation into the USAR and TRM update, the 
changes to the description of the facility as described in the 
licensee's application dated December 16, 1999, as supplemented by 
letters dated December 21, 1999, and January 10, 2000, and evaluated in 
the staff's Safety Evaluation attached to this amendment.
    Amendment No.: 108.
    Facility Operating License No. NPF-47: The amendment revises the 
Technical Requirements Manual and Updated Safety Analysis Report.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes (64 FR 71511 dated December 21, 1999). 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 January 20, 
2000, but indicated that if the Commission made 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 NSHC determination are contained in a 
Safety Evaluation dated January 13, 2000.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 13, 2000.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.

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

    Date of amendment request: December 16, 1999
    Brief description of amendment: The proposed change would amend 
Technical Specification 4.18.5.b to allow tube 110/60 to remain in 
service through the current operating cycle (cycle 16) with two axial 
indications that have potential through-wall depths greater than the 
plugging limit. The axial indications are located in the roll 
transition region and are contained within the upper tubesheet.
    Date of issuance: January 13, 2000.
    Effective date: As of the date of issuance.
    Amendment No.: 203.
    Facility Operating License No. DPR-51: Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes (64 FR 73080 dated December 29, 1999). 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 January 28, 
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 NSHC determination are contained in a 
Safety Evaluation dated January 13, 2000.
    Attorney for Licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

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: July 20, 1998, as supplemented 
by letter dated June 29, 1999.
    Brief description of amendment: The amendment incorporates the 
Technical Specification changes necessary for implementation of the 
Boiling Water Reactor Owners' Group Reactor Stability Long-Term 
Solution, Enhanced Option 1-A.
    Date of issuance: January 19, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No: 141.
    Facility Operating License No. NPF-29: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46432).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 19, 2000.
    No significant hazards consideration comments received: No.

[[Page 6416]]

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: October 12, 1999.
    Brief description of amendments: The amendments would revise the 
Technical Specifications (TSs) Surveillance Requirement 4.6.2.2.d for 
the spray additive system to relocate the details associated with the 
acceptance criteria and test parameters to the associated TSs Bases. 
Additionally, certain administrative text format changes were made.
    Date of issuance: January 19, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 240 and 221.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 3, 1999 (64 FR 
59804).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 19, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 11, 1998, as 
supplemented by letters dated January 14, and August 5, 1999.
    Brief description of amendments: The amendments revise the combined 
Technical Specifications (TS) for the Diablo Canyon Power Plant, Unit 
Nos. 1 and 2 to revise TS 6.8.4f., ``Containment Polar and Turbine 
Building Cranes,'' to control the operation of the containment polar 
cranes in jet impingement zones during Modes 1, 2, 3, and 4.
    Date of issuance: January 12, 2000.
    Effective date: January 12, 2000.
    Amendment Nos.: Unit 1-137; Unit 2-137.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 21, 1999 (64 FR 
19561).
    The August 5, 1999, 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 amendments is contained 
in a Safety Evaluation dated January 12, 2000.
    No significant hazards consideration comments received: No.

Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: September 29, 1999, as 
supplemented December 7, 1999.
    Brief description of amendment: This amendment revises the 
Technical Specifications to allow, on a one-time basis only, the Power 
Authority of the State of New York to extend the allowed out-of-service 
time for the Residual Heat Removal Service Water (RHRSW) System from 7 
days to 11 days. This amendment is only applicable during installation 
of Modification 99-095 to the ``A'' RHRSW Strainer.
    Date of issuance: January 28, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 259.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 20, 1999 (64 FR 
56532).
    The December 7, 1999, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 28, 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: August 25, 1999.
    Brief description of amendments: The amendments authorize the 
licensee to perform single-cell charging of operable safety-related 
batteries by using non-Class 1E single-cell battery chargers, with 
proper electrical isolation. The single-cell chargers would be used to 
restore individual cell float voltage to the normal TS limit.
    Date of issuance: January 24, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 226 and 207.
    Facility Operating License Nos. DPR-70 and DPR-75.: Amendments 
revised the licenses.
    Date of initial notice in Federal Register: September 22, 1999 (64 
FR 51349).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 24, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 28, 1999.
    Brief description of amendment: The amendment revises the Ginna 
Station Improved Technical Specifications associated with the Reactor 
Coolant System Leakage Detection Instrumentation.
    Date of issuance: January 19, 2000.
    Effective date: January 19, 2000.
    Amendment No.: 76.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43778)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 19, 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: April 19, 1999, as supplemented 
by letter dated November 1, 1999.
    Brief description of amendments: The amendments revised the 
Technical Specifications Surveillance Requirement (SR) 3.3.5.2 and 
associated Bases to allow the loss of voltage and degraded voltage trip 
setpoints to be treated as ``nominal'' values.
    Date of issuance: January 19, 2000
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 111--Unit 1; 89--Unit 2.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 1, 1999 (64 FR 
67340).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 19, 2000.
    No significant hazards consideration comments received: No.

[[Page 6417]]

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

    Date of amendment request: September 8, 1999, as supplemented 
November 9, 1999.
    Brief description of amendments: The amendments revised Technical 
Specification 3/4.8.1, ``A.C. Sources, Operating,'' and associated 
Bases, by deleting the 18-month surveillance to subject the standby 
diesel generator to inspections in accordance with procedures prepared 
in conjunction with its manufacturer's recommendations. The 
surveillance requirements have been relocated to the Technical 
Requirements Manual.
    Date of issuance: January 14, 2000.
    Effective date: January 14, 2000, to be implemented within 30 days.
    Amendment Nos.: Unit 1--121 ; Unit 2--109
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 1, 1999 (64 FR 
67341).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 14, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: September 8, 1999, as supplemented by 
letter dated November 9, 1999.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3/4.8.1, ``A.C. Sources, Operating,'' and associated 
Bases, by eliminating the requirement for accelerated testing of the 
standby diesel generators and the associated reporting requirements. 
The TS Index was also revised to reflect these changes.
    Date of issuance: January 14, 1999.
    Effective date: January 14, 1999.
    Amendment Nos.: Unit 1--122 ; Unit 2--110.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 3, 1999 (64 FR 
59806).
    The November 9, 1999, supplement provided additional clarifying 
information that was within the scope of the original application and 
Federal Register notice 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 January 14, 2000.
    No significant hazards consideration comments received: No

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

    Date of application for amendment: October 21, 1999.
    Brief description of amendment: The amendment corrects two textual 
errors and changes the designation of a referenced figure.
    Date of Issuance: January 11, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 183.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 17, 1999 (64 
FR 62717).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated January 11, 2000.
For the Nuclear Regulatory Commission.

No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 2nd day of February 2000.

For the Nulcear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-2835 Filed 2-8-00; 8:45 am]
BILLING CODE 7590-01-P