[Federal Register Volume 65, Number 144 (Wednesday, July 26, 2000)]
[Notices]
[Pages 46005-46023]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-18771]


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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 July 1, 2000, through July 14, 2000. The 
last biweekly notice was published on July 12, 2000 (65 FR 43040).

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 August 25, 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

[[Page 46006]]

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

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

    Date of amendment request: June 19, 2000 (U-603367).
    Description of amendment request: The proposed amendment would 
allow some emergency diesel generator (EDG) Technical Specification 
surveillance requirements to be performed during plant operation 
instead of during plant shutdown as now required. These EDG 
surveillance tests include load rejection tests and the EDG 24-hour run 
test.
    Basis for proposed no significant hazards consideration 
determination: The NRC staff has performed an analysis of the issue of 
no significant hazards consideration which is presented below:
    1. No changes will be made to the design or operation of the 
emergency diesel generators (EDGs) and the plant electrical 
distribution system will normally be aligned to minimize perturbations 
from the EDG tests during power operation. Additionally, while some 
portions of some surveillance tests will result in a decrease in EDG 
availability during power operation, EDG availability is not 
significantly decreased. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
any accident previously evaluated.
    2. No physical changes will be made to the plant. Electrical 
protective isolation devices will continue to act as before and 
Technical Specification system operability requirements are not being 
changed. Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously evaluated.
    3. No changes will be made to the design or operation of the 
emergency diesel generators (EDGs) and the plant electrical 
distribution system will normally be aligned to minimize perturbations 
from the EDG tests during power operation. Additionally, while some 
portions of some surveillance tests will result in a decrease in EDG 
availability during power operation, EDG availability is not 
significantly decreased. Therefore, the proposed change does not 
significantly reduce a margin of safety.
    Based on its initial review, the NRC staff finds that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the

[[Page 46007]]

amendment request involves no significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius 
LLP, 1800 M Street, NW, Washington, DC 20036.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: June 27, 2000.
    Description of amendment request: The proposed amendment would 
revise Section 3.5.1, ``Safety Injection Tanks (SITs),'' of the 
Palisades Plant Improved Technical Specifications (ITS) as issued by 
the NRC on November 30, 1999 (Amendment No. 189), for implementation on 
or before October 31, 2000. Specifically, Condition A, which currently 
applies to ``One SIT inoperable due to boron concentration not within 
limits,'' would be expanded to include ``OR One SIT inoperable due to 
the inability to verify level or pressure.'' Required Action A.1, which 
currently states ``Restore boron concentration to within limits,'' 
would be changed to state ``Restore SIT to OPERABLE status.'' The 
specified Completion Time for the revised Required Action A.1 would 
remain as 72 hours. Condition B, which applies to ``One SIT inoperable 
for reasons other than Condition A,'' would be changed to specify a 
Completion Time of 24 hours (rather than the current 1 hour) to restore 
the SIT to OPERABLE status. The licensee also forwarded revised pages 
to the Palisades ITS Bases for these proposed changes. Additional 
changes proposed in the licensee's application dated June 27, 2000, 
(which address the Low-Pressure Safety Injection System) are outside 
the scope of this Federal Register (FR) notice and are addressed in a 
separate FR notice.
    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 in Accordance with the Proposed Amendment] Does Not 
Involve a Significant Increase in the Probability or Consequences of 
an Accident Previously Evaluated.
    The Safety Injection Tanks (SITs) are passive components in the 
Emergency Core Cooling System. The SITs 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.
    SITs were designed to mitigate the consequences of Loss of 
Coolant Accidents (LOCA). These proposed changes do not affect any 
of the assumptions used in deterministic LOCA analysis. Hence the 
consequences of accidents previously evaluated do not change. In 
addition, in order to fully evaluate the effect of the SIT Allowable 
Outage Time (AOT) [a.k.a, ``Completion Time'] extension, 
probabilistic safety analysis (PSA) methods were utilized. The 
results of these analyses show no significant increase in the core 
damage frequency or large early release frequency. As a result, from 
a PSA standpoint, there would be no significant increase in the 
consequences of an accident previously evaluated. These analyses are 
detailed in CE NPSD-994, Combustion Engineering Owners Group ``Joint 
Applications Report for Safety Injection Tank AOT/STI [surveillance 
time interval] Extension.''
    The changes pertaining to SIT inoperability based solely on 
instrumentation malfunction do not involve a significant increase in 
the consequences of an accident as evaluated and endorsed by the 
Nuclear Regulatory Commission (NRC) in NUREG-1366, ``Improvements to 
Technical Specifications Surveillance Requirements.'' This 
evaluation is applicable to the Palisades Plant.
    Therefore, these changes do not involve an increase in the 
probability or consequences of any accident previously evaluated.
    [Operation in Accordance with the Proposed Amendment] Does Not 
Create the Possibility of a New or Different Kind of Accident from 
any Previously Evaluated.
    The proposed change does not change the design, configuration, 
or method of operation of the plant. The proposed configuration (one 
SIT out of service) is already allowed. Therefore, this change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    [Operation in Accordance with the Proposed Amendment] Does Not 
Involve a Significant Reduction in the Margin of Safety.
    The proposed changes do not affect the limiting conditions for 
operation or their bases that are used in the deterministic analyses 
to establish the margin of safety. The proposed configuration (one 
SIT out of service) is already allowed. PSA evaluations were used to 
evaluate these changes. The results of these analyses show no 
significant increase in the core damage frequency or large early 
release frequency. These evaluations are detailed in CE NPSD-994. 
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: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: June 27, 2000.
    Description of amendment request: The proposed amendment would 
revise Section 3.5.2, ``ECCS [Emergency Core Cooling System]--
Operating,'' of the Palisades Plant Improved Technical Specifications 
(ITS) as issued by the NRC on November 30, 1999 (Amendment No. 189), 
for implementation on or before October 31, 2000. Specifically, the 
change would extend the Completion Time (a.k.a., allowed outage time or 
AOT) for a single low-pressure safety injection (LPSI) subsystem from 
72 hours to 7 days. The change would apply for operating Modes 1, 2, 
and 3 with the primary coolant system temperature at or above 325 
degrees F. The licensee also forwarded revised pages to the Palisades 
ITS Bases for the proposed change.
    Additional changes proposed in the licensee's application dated 
June 27, 2000, (which address the safety injection tanks) are outside 
the scope of this Federal Register (FR) notice and are addressed in a 
separate FR notice.
    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 in Accordance with the Proposed Amendment] Does Not 
Involve a Significant Increase in the Probability or Consequences of 
an Accident Previously Evaluated.
    The Low Pressure Safety Injection system (LPSI) is part of the 
Emergency Core Cooling System. LPSI components are not accident 
initiators in any accident previously evaluated. Therefore, this 
change does not involve an increase in the probability of an 
accident previously evaluated.
    The LPSI system is primarily designed to mitigate the 
consequences of a large Loss of Coolant Accident (LOCA). These 
proposed changes do not affect any of the assumptions used in 
deterministic LOCA analysis. Hence the consequences of accidents 
previously evaluated do not change. In addition, in order to fully 
evaluate the effect of the LPSI AOT extension, probabilistic safety 
analysis (PSA) methods were utilized. The results of these analyses 
show no significant increase in the core damage frequency. As a 
result, from a PSA standpoint, there would be no significant 
increase in the consequences of an accident previously evaluated. 
These analyses are detailed in CE NPSD-995, Combustion Engineering 
Owners Group ``Joint Applications Report for Low Pressure Safety 
Injection System AOT Extension.''
    Therefore, these changes do not involve an increase in the 
probability or consequences of any accident previously evaluated.

[[Page 46008]]

    [Operation in Accordance with the Proposed Amendment] Does Not 
Create the Possibility of a New or Different Kind of Accident from 
any Previously Evaluated.
    The proposed changes do not change the design, configuration, or 
method of operation 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 changes do not alter assumptions made in the safety 
analysis and licensing basis. Therefore, these changes do not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    [Operation in Accordance with the Proposed Amendment] Does Not 
Involve a Significant Reduction in the Margin of Safety.
    The proposed changes do not affect the limiting conditions for 
operation or their bases used in the deterministic analyses to 
establish the margin of safety. PSA evaluations were used to 
evaluate these changes. These evaluations demonstrate that the 
changes are either risk neutral or risk beneficial. These 
evaluations are detailed in CE NPSD-995. 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 changes. 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: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: Claudia M. Craig.

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 amendment request: June 21, 2000.
    Description of amendment request: The proposed amendments would 
modify the Emergency Feedwater System (EFW) section of the Updated 
Final Safety Analysis Report.
    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) Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    No. The EFW System is utilized to mitigate the consequences of 
an accident. Failure of the EFW System is not a precursor to any 
accident evaluated in the UFSAR [Updated Final Safety Analysis 
Report].
    The UFSAR change proposes additional exceptions to the ability 
of the EFW system to mitigate specific events coupled with a single 
failure. These exceptions are appropriate, because diverse systems 
(i.e., the SSF [standby shutdown facility] ASW [auxiliary service 
water] System or EFW System from another unit) are available to 
mitigate the defined transient/accident and the probability of the 
defined transient/accident occurring is small.
    The proposed UFSAR changes do not involve any adverse impact on 
containment integrity, radiological release pathways, fuel design, 
filtration systems, main steam relief valve setpoints, or radwaste 
systems. In addition, it does not create any new radiological 
release pathways.
    Therefore, it is concluded that the proposed changes will not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated:
    No. The EFW System is utilized to mitigate the consequences of 
an accident. Failure of the EFW System is not a precursor to any 
accident evaluated in the UFSAR. The proposed UFSAR changes do not 
physically effect the plant, nor do they increase the risk of a unit 
trip or reactivity excursion. This proposed change does not 
introduce any new accident precursors. Therefore, these proposed 
changes do not create the possibility of any new or different kind 
of accident.
    (3) Involve a significant reduction in a margin of safety.
    No. A Probabilistic Risk Assessment (PRA) evaluation of the 
single failures identified in a failure modes and effects analysis 
performed for the EFW System concluded that there are no single 
active failures that contribute significantly to core damage 
frequency.
    The UFSAR change proposes additional exceptions to the ability 
of the EFW system to mitigate specific events coupled with a single 
failure. These exceptions are appropriate, because the probability 
of the defined transient/accident occurring is small, and diverse 
systems (i.e., the SSF ASW System or EFW System from another unit) 
are available to mitigate the defined transient/accident.
    The proposed UFSAR changes do not involve: (1) a physical 
alteration of the plant; (2) the installation of new or different 
equipment; or (3) any impact on the fission product barriers or 
safety limits.
    Therefore, it is concluded that the proposed UFSAR changes will 
not result in a significant decrease 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: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: June 29, 2000.
    Description of amendment request: The current Arkansas Nuclear One, 
Unit 2 (ANO-2) Technical Specification (TS) 3.6.2.3 states: ``Two 
independent containment cooling groups shall be OPERABLE with at least 
one operational cooling unit in each group.'' The proposed change will 
modify this wording as follows: ``Two independent containment cooling 
groups shall be OPERABLE with two operational cooling units in each 
group.'' In addition, the proposed amendment would change the 
surveillance requirements contained in TS 4.6.2.3.a. At the present 
time, TS 4.6.2.3.a. would allow a containment cooler group with a 
minimum service water flow rate of 1250 gpm to be declared operable if 
one of the two cooling units and associated fan is operable. As a 
result of this proposed change, the surveillance requirements will be 
modified to require both cooling units per group to be operable for the 
containment cooler group to be operable.
    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 containment cooling units do not have the ability to cause 
an accident, however, they do serve to mitigate containment accident 
conditions. The new MSLB [Main Steam Line Break] and LOCA [Loss of 
Coolant Accident] analyses contain the same assumptions relating to 
containment heat removal as the original analyses, i.e., at least 
one containment building cooling unit in conjunction with one train 
of CSS [containment spray system] is adequate for containment heat 
removal. During 2R14 [Unit 2, 14th refueling outage] the containment

[[Page 46009]]

coolers will be modified by adjusting the fan pitch, which will 
reduce fan flow as well as the post DBA [Design Basis Accident] 
motor horsepower. To offset this lower containment cooler fan 
airflow rate, two cooling units per group will be required. The 
resulting heat removal capacity with two containment cooling units 
in service at the new blade pitch position is greater than the 
required heat removal assumed in the LOCA and MSLB analyses.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.

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

    Assuming the single failure of a loss of one group of 
components, the remaining group with two cooling units will continue 
to be available. The modification to the fan blade pitch will result 
in a lower air flow rate through each containment cooler. However, 
the requirement for two units per group to be operable provides 
adequate heat removal capacity for containment uprate conditions. 
Therefore, the heat removal capacity assumed in the Containment 
Uprate analysis remain valid. The previous ability to credit either 
cooler unit provided additional design margin whereby the required 
redundancy is still provided by this change.
    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 Containment Cooling System ensures that (1) the containment 
air temperature will be maintained within limits during normal 
operation, and (2) adequate heat removal capacity is available when 
operated in conjunction with the containment spray systems during 
post-LOCA conditions. The modification planned during 2R14 will 
result in a lower air flow rate through each cooling unit and a 
corresponding reduction in heat transfer capability of each cooling 
unit. However, the safety margin is still maintained by requiring 
both cooler units to be operable and thus providing adequate heat 
removal capacity to remain below the containment design pressure.
    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.

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

    Date of amendment request: June 9, 2000.
    Description of amendment request: The proposed amendment would 
reduce the bypass valve (BV) cycling frequency of SR 3.7.7.1 from 31 
days to 92 days. This will make the test frequency for the BVs 
consistent with the testing frequency for the other Main Turbine Valves 
(e.g. Main Turbine Control, Stop, and Combined Intermediate Valves). 
The 92-day frequency is also consistent with the typical testing 
frequency for stroking safety-related valves under TS 5.5.6, In-Service 
Testing Program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The current TS SR 3.7.7.1 requires that the BVs be cycled once 
every 31 days to demonstrate that the BVs are mechanically OPERABLE 
(free to move). DAEC in-house operating experience has shown that 
the BVs have reliable equipment performance in that they 
consistently pass the valve cycling test at both the 31-day and 92-
day frequency. A test frequency of 92 days is sufficient to ensure 
the reliability of the BVs. The DAEC is analyzed for certain 
transient events with the assumption that the MTBS is out-of-service 
(e.g. turbine trips, generator load rejects, feedwater flow 
controller failure at maximum demand). Continued plant operation is 
allowed in cases of inoperable MTBS provided the more restrictive 
MCPR limit is applied (LCO 3.7.7). Margin to the MCPR Safety Limit 
is bounded by the analyzed failure of the MTBS. Should the BV fail a 
cycling test, the TS required actions would be taken accordingly. 
Therefore, this proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    There are no modifications made to the MTBS (including BVs) or 
system operations in this proposed TS amendment. The only change is 
the BV's cycling frequency from 31 days to 92 days. The proposed TS 
amendment does not alter the OPERABILITY requirements or performance 
characteristics of the MTBS. The reduced BV cycling frequency 
reduces the need for reactivity changes and pressure perturbations 
on the reactor. This proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    The only change by this proposed TS amendment is the frequency 
of the BV's cycling test from 31 days to 92 days. The OPERABILITY 
requirement and functional characteristics of the MTBS remain 
unchanged. DAEC in-house operating experience has demonstrated that 
a 92-day test frequency provides reasonable assurance that the BVs 
remain OPERABLE. The BV's response times are used in determining the 
effect on the MCPR. The surveillance tests that ensure the MTBS 
meets the system's automatic actuation requirements (SR 3.7.7.2) and 
response time limits (SR 3.7.7.3) are not affected by this proposed 
TS amendment and will continue to be performed at the current TS 
frequency. Therefore, this 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: Al Gutterman, Morgan, Lewis & Bockius, 1800 
M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: June 14, 2000.
    Description of amendment request: Alliant Energy Corporation (AEC) 
plans to merge and consolidate another utility it owns, Interstate 
Power Company (IPC), with IES Utilities Inc., effective January 1, 
2001. The name of the surviving corporation, IES, would be changed to 
Interstate Power and Light Company (IP&L).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    No physical or operational changes to the DAEC will result from 
changing the corporate name. The DAEC will continue to be operated 
in the same manner with the same organization. Therefore, the 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different

[[Page 46010]]

kind of accident from any accident previously evaluated.
    No physical or operational changes to the DAEC will result from 
changing the corporate name. The DAEC will continue to be operated 
in the same manner with the same organization. Therefore, the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    No physical or operational changes to the DAEC will result from 
changing the corporate name. The DAEC will continue to be operated 
in the same manner with the same organization. Therefore, 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: Al Gutterman, Morgan, Lewis & Bockius, 1800 
M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Claudia M. Craig.

Northeast Nuclear Energy Company, et al., Docket No. 50-245, Millstone 
Nuclear Power Station, Unit No. 1, New London County, Connecticut

    Date of amendment request: June 6, 2000.
    Description of amendment request: The proposed amendment would 
revise the Millstone Nuclear Power Station, Unit No. 1 license to 
modify or remove license conditions and confirmatory orders to reflect 
the permanently defueled condition of the unit. 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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The purpose of the proposed changes is to revise the Millstone 
Unit No. 1 Operating License to only address conditions and 
requirements that are germane to the permanently shutdown and 
defueled condition. Since Millstone Unit No. 1 has permanently 
ceased operation and will be maintained in a defueled condition, 
many provisions of the license related to the operation of the plant 
are no longer appropriate. Elimination of the unnecessary 
requirements and statements allows the plant staff to focus on those 
requirements which continue to be appropriate to the existing plant 
conditions. The proposed changes do not affect the only design basis 
accident that continues to be applicable (i.e., the fuel handling 
accident). Therefore, the changes do not increase the probability or 
consequences of any previously evaluated accident.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The purpose of the proposed changes is to revise the Millstone 
Unit No. 1 Operating License to only address conditions and 
requirements that are germane to the permanently shutdown and 
defueled condition. Since Millstone Unit No. 1 has permanently 
ceased operation and will be maintained in a defueled condition, 
many provisions of the license related to the operation of the plant 
are no longer appropriate. Elimination of the unnecessary 
requirements and statements allows the plant staff to focus on those 
requirements which continue to be appropriate to the existing plant 
conditions. The proposed changes do not affect storage of spent 
fuel. Therefore, the proposed changes do not create a different kind 
of accident from those previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The purpose of the proposed changes is to revise the Millstone 
Unit No. 1 Operating License to only address conditions and 
requirements that are germane to the permanently shutdown and 
defueled condition. Since Millstone Unit No. 1 has permanently 
ceased operation and will be maintained in a defueled condition, 
many provisions of the license related to the operation of the plant 
are no longer appropriate. Elimination of the unnecessary 
requirements and statements allows the plant staff to focus on those 
requirements which continue to be appropriate to the existing plant 
conditions. The proposed changes do not affect storage of spent 
fuel. 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: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: Michael T. Masnik

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

    Date of amendment request: June 28, 2000.
    Description of amendment request: The proposed changes to the 
Technical Specifications (TSs) are associated with Section 3/4.7.6, 
``Control Room Emergency Ventilation System.'' Specifically, TS 3.7.6.1 
will be revised to add a footnote that the Control Room boundary can be 
opened intermittently under administrative control, and add a new Modes 
1 through 4 action requirement that will allow 24 hours to restore the 
Control Room boundary. In addition, various editorial changes 
associated with action requirement format and letter designations are 
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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The action requirements for the Control Room Emergency 
Ventilation System have been changed to address the impact a loss of 
boundary integrity has on the associated system. The proposed 
changes to the action requirements will not cause an accident. 
Allowing the Control Room boundary to be opened intermittently under 
administrative controls will have no adverse impact on the 
consequences of the design basis accidents since the administrative 
controls will be able to rapidly restore boundary integrity when 
required. Allowing 24 hours to restore the Control Room boundary in 
Modes 1 through 4 could result in an increase in the consequences of 
a design basis accident to the Control Room personnel. However, 
considering the low probability of a design basis accident occurring 
during this time, the proposed allowed outage time is reasonable to 
allow the boundary integrity to be restored before requiring a plant 
shutdown.
    These changes are consistent with Technical Specification 
3.6.5.2, ``Containment Systems--Enclosure Building,'' which allows 
normal entry and egress through associated access openings 
(Surveillance Requirement 4.6.5.2.1) and 24 hours to restore 
Enclosure Building integrity, and with generic industry guidance 
(NUREG-1432, Technical Specification 3.7.11, TSTF-287, Rev. 5).
    The proposed changes to address format issues will not result in 
any technical changes to the current requirements.
    The proposed Technical Specification changes will have no 
adverse effect on plant operation or the operation of accident 
mitigation equipment, and will not significantly impact the 
availability of accident mitigation equipment. The plant response to 
the design basis accidents will not change. In addition, the 
equipment covered by this specification is not an accident initiator 
and can not cause an accident. Therefore, the proposed changes will 
not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed Technical Specification changes do not impact any 
system or

[[Page 46011]]

component which could cause an accident. The proposed changes will 
not alter the plant configuration (no new or different type of 
equipment will be installed) or require any unusual operator 
actions. The proposed changes will not alter the way any structure, 
system, or component functions, and will not significantly alter the 
manner in which the plant is operated. There will be no adverse 
effect on plant operation or accident mitigation equipment. The 
proposed changes do not introduce any new failure modes. Also, the 
response of the plant and the operators following an accident will 
not be significantly different as a result of these changes. In 
addition, the accident mitigation equipment affected by the proposed 
changes is not an accident initiator. Therefore, the proposed 
changes will not create the possibility of a new or different kind 
of accident from any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes to Technical Specification 3.7.6.1 are 
consistent with Technical Specification 3.6.5.2 which allows normal 
entry and egress through associated access openings (SR 4.6.5.2.1) 
and 24 hours to restore Enclosure Building integrity, and with 
generic industry guidance (NUREG-1432, Technical Specification 
3.7.11, TSTF-287, Rev. 5). If the Control Room boundary is not 
operable, the proposed action requirements will require timely 
restoration of the boundary or the plant will be placed in a 
configuration where there is no adverse impact associated with the 
loss of Control Room boundary integrity. The proposed allowed outage 
time provides a reasonable time for repairs before requiring a plant 
shutdown, and reflects the low probability of an event occurring 
while the boundary is inoperable. The proposed shutdown times, which 
are consistent with times already contained in the Millstone Unit 
No. 2 Technical Specifications and with generic industry guidance 
(NUREG-1432), will allow an orderly shutdown to be performed.
    The proposed changes to address format issues will not result in 
any technical changes to the current requirements. These proposed 
changes will not adversely impact any of the design basis accidents 
or the associated accident mitigation equipment.
    The proposed changes will have no adverse effect on plant 
operation or equipment important to safety. The plant response to 
the design basis accidents will not change and the accident 
mitigation equipment will continue to function as assumed in the 
design basis accident analyses. Therefore, there will be no 
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: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: James W. Clifford.

Northeast Nuclear Energy Company, et al., Docket Nos. 50-336 and 50-
423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London 
County, Connecticut

    Date of amendment request: June 26, 2000.
    Description of amendment request: The proposed changes to the 
Technical Specifications (TSs) are associated with the Reactivity 
Control Systems section. Specifically, the surveillance requirements 
associated with the frequency for determining the operability of each 
rod not fully inserted in the core will be revised from once every 31 
days to once every 92 days for Units 2 and 3. In addition, the 
surveillance requirement associated with the frequency of testing the 
Control Element Assembly Deviation Circuit will be revised from once 
every 31 days to once every 92 days for Unit 2.
    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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to Millstone Unit Nos. 2 and 3 Specification 
4.1.3.1.2 will revise the frequency for determining the operability 
of each rod that is not fully inserted in the core from once every 
31 days to once every 92 days. The proposed change in the frequency 
does not change any of the assumptions used in the safety analyses. 
On the other hand, the decrease in surveillance frequency will 
reduce the potential for reactor trips and the unnecessary 
challenges to the safety systems associated with the performance of 
the surveillance. Additionally, NNECO [Northeast Nuclear Energy 
Company] performed Millstone Unit Nos. 2 and 3 specific evaluations 
of the effect of changing the frequency of rod movement test from 31 
days to 92 days on Core Damage Frequency (CDF). These evaluations 
concluded that the change in test frequency from 31 days to 92 days 
has no adverse impact on CDF (the estimated potential risk 
associated with tripping the reactor as a result of this high risk 
surveillance is about 1.31E-8/yr for Millstone Unit No. 2 and 4.28E-
8/yr for Millstone Unit No. 3) and is therefore acceptable. 
Therefore, this change will not significantly increase the 
probability or consequences of an accident previously evaluated.
    The proposed change in the frequency of testing the CEA 
Deviation Circuit in Millstone Unit No. 2 Specification 4.1.3.1.3 
from once every 31 days to once every 92 days does not change any of 
the assumptions used in the safety analysis. On the other hand, the 
decrease in surveillance frequency will reduce the reactor trips and 
the unnecessary challenges to the safety systems associated with the 
performance of the surveillance. Additionally, the Deviation Circuit 
has excellent testing history and increasing the surveillance 
interval from 31 days to 92 days will have no adverse effect on its 
overall reliability. The Nuclear Regulatory Commission approved this 
increase in surveillance interval as part of TSTF [Technical 
Specifications Task Force] -127.[ ] Therefore, this change will not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes will not alter [the] configuration of the 
plants (no new or different type of equipment will be installed) or 
require any new or unusual operator actions. They do not alter the 
way any structure, system, or component functions and do not alter 
the manner in which the plants are operated. Therefore, the proposed 
changes will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes in the surveillance frequency do not change 
any of the assumptions used in the safety analyses. Additionally, 
NNECO performed Millstone Unit Nos. 2 and 3 specific evaluations of 
the effect of changing the frequency of rod movement test from 31 
days to 92 days on CDF. These evaluations concluded that the change 
in test frequency from 31 days to 92 days has no adverse impact on 
CDF and is therefore acceptable. Therefore, the proposed changes 
will not result in a significant reduction in a margin of safety.
    As described above, this License Amendment Request does not 
involve a significant increase in the probability of an accident 
previously evaluated, does not involve a significant increase in the 
consequences of an accident previously evaluated, does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated, and does not result in a significant 
reduction in a margin of safety. Therefore, NNECO has concluded that 
the proposed changes do not involve an SHC [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: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: James W. Clifford.

[[Page 46012]]

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Docket No. 
50-277, Peach Bottom Atomic Power Station, Unit No. 2, York County, 
Pennsylvania

    Date of application for amendment: June 14, 2000.
    Description of amendment request: The licensee requests that the 
Peach Bottom Atomic Power Station (PBAPS), Unit 2, Technical 
Specifications (TS) contained in Appendix A to the Operating License be 
amended to: (1) Revise TS 2.1.1.2 to reflect changes in the Safety 
Limit Minimum Critical Power Ratios (SLMCPRs) due to the cycle-specific 
analysis performed by Global Nuclear Fuel (formerly General Electric 
Nuclear Energy (GENE)) for PBAPS, Unit 2, Cycle 14, which includes the 
use of the GE-14 product line, (2) delete the cycle-specific footnote 
for the SLMCPRs contained in TS 2.1.1.2 (``Reactor Core SLs''), and (3) 
update a reference contained in TS 5.6.5.b.2 (``Core Operating Limits 
Report'') which documents an analytical method used to determine the 
core operating limits. Basis for proposed no significant hazards 
consideration determination: As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    (1) The proposed TS changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The derivation of the cycle specific SLMCPRs for incorporation 
into the TS, and its use to determine cycle specific thermal limits, 
has been performed using the methodology discussed in ``General 
Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13, 
and U.S. Supplement, NEDE-24011-P-A-13-US, August 1996, and 
Amendment 25. Amendment 25 was approved by the NRC in a March 11, 
1999 safety evaluation report.
    The basis of the SLMCPR calculation is to ensure that greater 
than 99.9% of all fuel rods in the core avoid transition boiling if 
the limit is not violated. The new SLMCPRs preserve the existing 
margin to transition boiling. The GE-14 fuel is in compliance with 
Amendment 22 to ``General Electric Standard Application for Reactor 
Fuel,'' NEDE-24011-P-A-13, and U. S. Supplement, NEDE-24011-P-A-13-
US, August, 1996 (GESTAR-II), which provides the fuel licensing 
acceptance criteria. The probability of fuel damage will not be 
increased as a result of these changes. Additionally, as a result of 
the use of the GE-14 product line, no dose calculations are being 
adversely impacted. Therefore, the proposed TS changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    In addition to the change to the SLMCPRs, the footnote to TS 
2.1.1.2 is being deleted. The footnote associated with TS 2.1.1.2 
was originally included to ensure that the SLMCPR value was only 
applicable for the identified cycle. Since that time, Amendment 25 
has been approved. Therefore, this footnote is no longer necessary. 
The footnote was for information only, and has no impact on the 
design or operation of the plant. A similar change was previously 
approved for PBAPS, Unit 3, as discussed in the NRC safety 
evaluation (Amendment No. 233), dated October 5, 1999. The deletion 
of the footnote associated with TS 2.1.1.2 is an administrative 
change that does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The reference to the Revision 1 ARTS/MELLLA analysis contained 
in TS 5.6.5.b.2 is being updated to a Revision 2 analysis, to 
reflect changes that were previously approved by the NRC as 
documented in the safety evaluation report dated August 10, 1994 
(Amendment No. 192 for PBAPS, Unit 2). This is an administrative 
change which will ensure that the references contained in the PBAPS, 
Unit 2 Technical Specifications are accurate and consistent with 
other licensing documents. No technical changes are occurring which 
have not been previously approved by the NRC. A similar change was 
previously approved for PBAPS, Unit 3, as discussed in the NRC 
safety evaluation (Amendment No. 233), dated October 5, 1999. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The SLMCPR is a TS numerical value, calculated to ensure that 
transition boiling does not occur in 99.9% of all fuel rods in the 
core if the limit is not violated. The new SLMCPRs are calculated 
using NRC approved methodology discussed in ``General Electric 
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13 (GESTAR-
II), and U.S. Supplement, NEDE-24011-P-A-13-US, August 1996, and 
Amendment 25. Additionally, the GE-14 fuel is in compliance with 
Amendment 22 to ``General Electric Standard Application for Reactor 
Fuel,'' NEDE-24011-P-A-13, and U.S. Supplement, NEDE-24011-P-A-13-
US, August, 1996 (GESTAR-II), which provides the fuel licensing 
acceptance criteria. The SLMCPR is not an accident initiator, and 
its revision will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    Additionally, this proposed change will delete footnotes 
contained in TS 2.1.1.2 as the result of the NRC approval of 
analysis associated with Amendment 25. The proposed change also 
updates the ARTS/MELLLA analysis reference contained in TS 
5.6.5.b.2. This revision contains information which was previously 
approved by the NRC. Similar changes were previously approved for 
PBAPS, Unit 3, as discussed in the NRC safety evaluation (Amendment 
No. 233), dated October 5, 1999. Therefore, the deletion of the 
footnote associated with TS 2.1.1.2, and the updating of the 
reference contained in TS 5.6.5.b.2 are administrative changes that 
do not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    There is no significant reduction in the margin of safety 
previously approved by the NRC as a result of: (1) The proposed 
changes to the SLMCPRs, which includes the use of GE-14 fuel, (2) 
the proposed change that will delete the footnote to TS 2.1.1.2, and 
(3) updating the ARTS/MELLLA analysis reference contained in TS 
5.6.5.b.2. The new SLMCPRs are calculated using methodology 
discussed in ``General Electric Standard Application for Reactor 
Fuel,'' NEDE-24011-P-A-13 (GESTAR-II), and U.S. Supplement, NEDE-
24011-P-A-13-US, August 1996, and Amendment 25. The SLMCPRs ensure 
that greater than 99.9% of all fuel rods in the core will avoid 
transition boiling if the limit is not violated when all 
uncertainties are considered, thereby preserving the fuel cladding 
integrity. Therefore, the proposed TS changes will not involve a 
significant reduction in the margin of safety previously approved by 
the NRC.
    Additionally, the proposed changes that delete the footnote to 
TS 2.1.1.2, and update the revision to the ARTS/MELLLA analysis 
reference contained in TS 5.6.5.b.2, are administrative changes that 
will not significantly reduce the margin of safety previously 
approved by the NRC.

    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: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101
    NRC Section Chief: James W. Clifford

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: February 4, 2000.
    Description of amendment request: The proposed amendment to the 
Indian Point 3 Technical Specifications (TSs) proposes to revise the 
main steam line break (MSLB) analysis to correct the assumption for 
non-isolable feedwater and also to revise assumptions regarding boron 
in the safety injection piping and assumptions regarding shutdown 
margin.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 46013]]

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 Indian Point 3 in accordance with the proposed 
amendment would not involve a significant hazards consideration as 
defined in 10 CFR 50.92 since it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes include revised assumptions in the TS to 
correct non-conservative TS and revised TS with respect to the peak 
calculated containment pressure for a postulated MSLB. The changes 
take credit for existing boron in the SI [Safety Injection] system 
and existing shutdown margin, perform surveillance to verify the 
boron concentration, and revise the containment testing program to 
reflect a minimum test pressure that must bound the peak calculated 
pressure. These changes cannot increase the probability of an 
accident previously evaluated since they do not change plant 
operations and are not related to accident initiators. These changes 
will not increase the consequences of an accident previously 
evaluated since they do not change system operation to mitigate any 
accident and the use of a minimum containment test pressure ensures 
the TS required testing bounds the calculated peak calculated [sic] 
pressure.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes include revised assumptions in the TS to 
correct non-conservative TS and revised TS with respect to the peak 
calculated pressure. The changes take credit for existing boron in 
the SI system and existing shutdown margin, perform surveillance to 
verify the boron, and revise the containment testing program to 
reflect a minimum test pressure that must bound the peak calculated 
pressure. These changes do not physically alter the plant since they 
take credit for existing plant conditions and the physical act of 
sampling meets system design and Technical Specification 
requirements. Therefore, these changes do not create the possibility 
of a new or different type of accident from those previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes include revised assumptions in the TS to 
correct non-conservative TS and revised TS with respect to the peak 
calculated pressure. The changes take credit for existing boron in 
the SI system and existing shutdown margin, perform surveillance to 
verify the boron, and revise the containment testing program to 
reflect a minimum test pressure that must bound the peak calculated 
pressure. These changes do not involve a significant reduction in 
the margin of safety since the credited boron is part of the 
existing system design that has not been credited since the BIT 
[Boron Injection Tank] tank retirement. The credited shutdown margin 
is typical of the excess shutdown margin resulting from cycle 
specific core design.

    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 considerations. 
Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Section Chief: Marsha Gamberoni, Acting.

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

    Date of amendment request: March 13, 2000.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) Table 3.3-6, ``Radiation Monitoring 
Instrumentation,'' to change the Containment Gaseous Activity Monitor 
(R12A) alarm/trip setpoint for the Containment Purge and Pressure 
Relief system isolation for Mode 6 (Refueling) operation. Specifically, 
the existing setpoint of less than or equal to two times background 
would be changed to ``Set at less than or equal to 50% of the 10 CFR 
[Part] 20 concentration limits for gaseous effluents released to 
unrestricted areas.'' The proposed amendment will also specify an upper 
setpoint limit that is not presently required by the existing TS 
requirement. In addition, the associated TS Bases section would be 
revised to address the proposed change.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. The proposed change will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    During Mode 6 operation, the Containment Gaseous Activity 
Monitor R12A serves to monitor the gaseous activity concentration in 
the containment atmosphere, and provides an alarm and isolation of 
the Containment Purge and Pressure Relief system in response to high 
gaseous activity that would result from a Fuel Handling Accident 
inside containment. As such, the Containment Gaseous Activity 
Monitor is not considered as an initiator of any accident previously 
evaluated. Therefore, the proposed change would not affect the 
probability of an accident previously evaluated.
    The proposed setpoint would allow an alarm/trip setpoint to be 
higher than the current TS requirements. As a result, it would be 
expected that the consequences of an accident previously evaluated 
could possibly increase. However, the proposed setpoint value would 
isolate the Containment Purge and Pressure Relief system prior to 
reaching the 10 CFR Part 20 concentration limits for gaseous 
effluents released to unrestricted areas. The 10 CFR Part 20 limits 
are equivalent to the radio-nuclide concentrations which, if inhaled 
or ingested continuously over the course of a year, would produce at 
total effective dose equivalent of 0.05 rem (50 millirem or 0.5 
millisieverts). These restrictions are intended to minimize and 
limit the amount of dose received by individual members of the 
public during normal operations, and are considerably more 
restrictive than the 10 CFR Part 100 limits. The proposed change 
would not be considered a significant increase in the consequences 
of an accident previously evaluated because the revised setpoint 
would isolate the appropriate release path and maintain doses well 
below 10 CFR Part 100 limits. Therefore, the proposed change will 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change to the Containment Gaseous Activity Monitor 
alarm/trip setpoint will not create any new accident causal 
mechanisms. Plant operation will not be affected by the proposed 
amendments and no new failure modes will be created. Thus, the 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change will not involve a significant reduction 
in a margin of safety.
    An evaluation of a fuel handling accident inside containment has 
been performed by the licensee that demonstrates that the limits of 
10 CFR Part 100 would not be exceeded even though no containment 
isolation was assumed. The analysis assumed that all airborne 
activity reaching the containment atmosphere would exhaust to the 
environment within two hours (no containment isolation) and 
concluded that the exclusion area boundary doses were well within 
the limits of 10 CFR Part 100. The analysis

[[Page 46014]]

also demonstrated that the control room doses following the fuel 
handling accident inside containment would be within General Design 
Criterion 19 limits. Therefore, the changes proposed by the licensee 
will not involve a significant reduction in a margin of safety.
    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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: May 31, 2000.
    Description of amendment request: The proposed amendment would 
establish new charcoal filter testing requirements for the Auxiliary 
Building Ventilation (ABV) System, the Control Room Envelope Air 
Conditioning System (CREACS), and the Fuel Handling Ventilation (FHV) 
System consistent with the requirements delineated in Generic Letter 
99-02, ``Laboratory Testing of Nuclear-Grade Activated Charcoal,'' 
dated June 3, 1999. Specifically, the surveillance requirements 
associated with Limiting Conditions for Operation (LCOs) 3.7.6.1, 
3.7.7.1, and 3.9.12 would specify American Society for Testing and 
Materials (ASTM) D3803-1989, ``Standard Test Method for Nuclear-Grade 
Activated Carbon,'' as the testing methodology.
    The May 31, 2000, amendment request would replace Public Service 
Electric and Gas (PSE&G) Company's original November 24, 1999, 
application to change Salem Units 1 and 2 Technical Specifications (TS) 
surveillance requirements associated with the laboratory testing of 
charcoal samples for the ABV, CREACS, and FHV systems. Additional 
information associated with the November 24, 1999, submittal was 
provided on February 10, 2000. However, PSE&G has requested that the 
November 24, 1999, application be withdrawn.
    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 TS change does not involve any physical changes to 
plant structures, systems or components (SSC). The FHV, CREACS and 
ABV systems will continue to function as designed. The FHV, CREACS 
and ABV systems are designed to mitigate the consequences of an 
accident, and therefore, cannot contribute to the initiation of any 
accident. The proposed TS surveillance requirement changes implement 
testing methods that more appropriately demonstrate charcoal filter 
capability and establish acceptance criteria, which ensure that 
Salem's design basis assumptions are appropriately met. In addition, 
this proposed TS change will not increase the probability of 
occurrence of a malfunction of any plant equipment important to 
safety, since the manner in which the FHV, CREACS and ABV systems 
are operated is not affected by these proposed changes. The proposed 
surveillance requirement acceptance criteria ensure that the FHV, 
CREACS and ABV safety functions will be accomplished. Therefore, the 
proposed TS changes would not result in a significant increase of 
the consequences of an accident previously evaluated, nor do they 
involve an increase in the probability of an accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed TS changes do not involve any physical changes to 
the design of any plant SSC. The design and operation of the FHV, 
CREACS and ABV systems are not changed from that currently described 
in Salem's licensing basis. The FHV, CREACS and ABV systems will 
continue to function as designed to mitigate the consequences of an 
accident. Implementing the proposed charcoal filter testing methods 
and acceptance criteria does not result in plant operation in a 
configuration that would create a different type of malfunction to 
the FHV, CREACS and ABV systems than any previously evaluated. In 
addition, the proposed TS changes do not alter the conclusions 
described in Salem's licensing basis regarding the safety related 
functions of these systems.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes contained in this submittal implement TS 
requirements that: (1) Are consistent with the requirements 
delineated in Generic Letter 99-02; (2) implement testing methods 
that adequately demonstrate charcoal filter capability; and (3) 
establish appropriately conservative acceptance criteria. The 
charcoal filter efficiencies specified in the proposed surveillance 
requirements apply a safety factor of 2 to the efficiencies used in 
the design basis dose analysis. There are no increases to the 
currently approved offsite dose releases or the control room 
operator doses as a result of these surveillance requirement 
changes. Therefore, the proposed TS change will not result in 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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: June 14, 2000.
    Description of amendment request: The proposed license amendment 
would allow the use of the Best Estimate Analyzer For Core Operations--
Nuclear (BEACON) system at Salem to perform core power distribution 
measurements. BEACON is a core power distribution monitoring and 
support system based on a three dimensional nodal code. The system is 
used to provide data reduction for incore neutron flux maps, core 
parameter analysis and follow, and core prediction. The licensee has 
stated that BEACON will be used at Salem to augment the functionality 
of the flux mapping system when thermal power is greater than 25% of 
rated thermal power for the purpose of performing power distribution 
surveillance testing.
    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 provide a different method for measuring 
the core power distribution parameters and relocate[s] manufacturing 
and measurement uncertainty values from the TS [Technical 
Specifications] to the core operating limits report (COLR). The [TS] 
power distribution limits themselves are not changed and will 
continue to be measured and verified to be within limits as required 
by the current TS surveillances. The cycle-specific core operating 
limits, although not in TS, will be followed in the operation of the 
Salem Generating Station. The proposed amendment continues to 
require the same actions to be taken when or if limits are exceeded 
as are required by current TS.

[[Page 46015]]

    Each accident analysis addressed in the Salem Updated Final 
Safety Analysis Report (UFSAR) will be examined with respect to 
changes in cycle-dependent parameters, which are obtained from 
application of the NRC [Nuclear Regulatory Commission]-approved 
reload design methodologies, to ensure that the transient evaluation 
of new reloads are bounded by previously accepted analyses. This 
examination, which will be performed per requirements of 10 CFR 
50.59, ensures that future reloads will not involve an increase in 
the probability or consequences of an accident previously evaluated.
    The method of measuring core power distribution parameters and 
the location of manufacturing and measurement uncertainty values are 
not initiators of any previously evaluated accidents and has no 
influence or impact on the consequences those accidents. Therefore, 
the changes do not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    No safety-related equipment, safety function, or plant operation 
will be altered as a result of the proposed changes. The cycle 
specific variables are calculated using the NRC-approved methods and 
submitted to the NRC to allow the Staff to continue to trend the 
values of these limits. The TS will continue to require operation 
within the required core operating limits and appropriate actions 
will be taken when or if limits are exceeded. The change will not 
introduce any new accident initiators. Therefore, the change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes provide a different method for measuring 
the core power distribution parameters and relocates manufacturing 
and measurement uncertainty values from the TS to the COLR. The 
proposed method for measuring the core power distribution parameters 
has been verified by Westinghouse and reviewed and approved by the 
NRC. Appropriate measures exist to control the values of the 
manufacturing and measurement uncertainties. The proposed amendment 
continues to require operation within the core limits, as obtained 
from NRC-approved reload design methodologies. Appropriate actions 
that [are] required to be taken when or if limits are violated 
remain unchanged. Future changes to measurement and manufacturing 
uncertainties located in the current TS will be evaluated in 
accordance with the requirements of 10 CFR 50.59. Since the 10 CFR 
50.59 process does not allow any reduction in the margin of safety, 
prior NRC approval is required prior to a reduction in the margin of 
safety. Additionally, the Salem TS require revisions of the plant 
COLR be submitted to the NRC upon issuance. Therefore, the 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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: June 12, 2000.
    Description of amendment request: This amendment would revise the 
Virgil C. Summer Nuclear Station (VCSNS) Technical Specifications (TS) 
to incorporate new temperature and level limits for the ultimate heat 
sink (UHS) during plant operation in Modes 1-4. These limits are 
contained in TS Section 3/4.7.5. The minimum required service water 
pond (SWP) level would be increased from the 415' elevation to 416.5' 
and the maximum allowed temperature at the discharge of the service 
water pumps would be decreased from 95 deg.F to 90.5 deg.F.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Implementation of the new temperature and level limits for the 
service water pond do not contribute to the initiation of any 
accident evaluated in the FSAR [Final Safety Analysis Report]. 
Supporting factors are as follows:
     The new limits maintain the Service Water System (SWS) 
design temperature of 95 deg.F during a normal shutdown and post 
accident and have been developed in accordance with the general 
requirements of Regulatory Guide 1.27, Revision 2.
     Overall plant performance and operation is not altered 
by the proposed changes.
     Fluid and auxiliary systems, which are important to 
safety, are not adversely impacted and will continue to perform 
their design function.
    Therefore, since the reactor coolant pressure boundary integrity 
and system functions are not impacted, the probability of occurrence 
of an accident evaluated in the VCSNS FSAR will be no greater than 
the original design basis of the plant.
    The SWP level and temperature limits relate to the plant's 
ability to reject heat to the ultimate heat sink during normal 
operation, a normal plant shutdown and hypothetical accident 
conditions. The new limits preserve the SWS design temperature of 
95 deg.F, even during worst case post accident conditions, thus 
assuring that equipment within the SWS and its interfacing systems 
remain qualified and that the heat transport capability of the SWS 
and its interfacing systems [remain] within design values. Since the 
SWS and its interfacing systems will continue to perform their 
design functions, it is concluded that the consequences of an 
accident previously evaluated in the FSAR are not increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes revise the UHS temperature and level limits 
within TS 3/4.7.5 to incorporate the results of a new thermal 
analysis performed in accordance with the requirements of Regulatory 
Guide 1.27, Revision 2. The new limits ensure that SW temperature, 
as measured at the discharge of the SW pump, [remains] less than the 
design value of 95 deg.F. No new accident initiator mechanisms are 
introduced as:
     Structural integrity of the RCS [reactor coolant 
system] is not challenged.
     No new failure modes or limiting single failures are 
created.
     Design requirements on all affected systems are met.
    Since the safety and design requirements continue to be met and 
the integrity of the reactor coolant system pressure boundary is not 
challenged, no new accident scenarios have been created. Therefore, 
the types of accidents defined in the FSAR continue to represent the 
credible spectrum of events to be analyzed which determine safe 
plant operation.
    3. Does this change involve a significant reduction in margin of 
safety?
    The proposed changes revise the UHS temperature and level limits 
[within] TS 3/4.7.5 to incorporate the results of a new thermal 
analysis performed in accordance with the requirements of Regulatory 
Guide 1.27, Revision 2. The new limits ensure that SW temperature, 
as measured at the discharge of the SW pump, [remains] less than the 
design value of 95 deg.F under both normal and post-accident 
conditions using the worst case combination of meteorology and 
operational parameters. Design margins associated with systems, 
structures and components that are cooled by the SWS are not 
affected. Since the SWS design temperature is maintained during both 
normal and worst case accident conditions, the results and 
conclusions for all design basis accidents remain applicable.
    The proposed changes impose more restrictive operating 
limitations, and their use provides increased assurance that the SWS 
design temperature will not be exceeded. Since the UHS will continue 
to provide a 30 day cooling water supply to safety related equipment 
without exceeding their design basis temperature, it is concluded 
that the changes do not involve a significant reduction in the 
margin of safety.


[[Page 46016]]


    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: Thomas G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: L. Raghavan, 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: October 13, 1999, as supplemented by 
letter dated June 1, 2000.
    Description of amendment request: The proposed amendments would 
revise Vogtle's Technical Specification to permit relaxation of allowed 
bypass test time and completion times for Limiting Conditions for 
Operations (LCO) 3.3.1, Reactor Trip System Instrumentation and LCO 
3.3.2, Engineered Safety Feature Actuation System Instrumentations. 
These changes specifically revise the completions times from 6 hours to 
72 hours for inoperable analog instruments, increase bypass times from 
4 hours to 12 hours for surveillance testing of analog channels, and 
increase completion times from 6 hours to 24 hours for an inoperable 
logic cabinet or master and slave relays.
    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 license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The reactor trip and engineered safety features functions are 
not initiators of any design basis accident or event, and therefore 
the proposed changes do not increase the probability of any accident 
previously evaluated. The proposed changes to the allowed Completion 
Times and bypass test times do not change the response of the plant 
to any accidents and have an insignificant impact on the reliability 
of the reactor trip system and engineered safety feature actuation 
system (RTS and ESFAS) signals. The RTS and ESFAS will remain highly 
reliable and the proposed changes will not result in a significant 
increase in the risk of plant operation. This is demonstrated by 
showing that the impact on plant safety as measured by core damage 
frequency (CDF) is less than 1.0E-06 per year and the impact on 
large early release frequency (LERF) is less than 1.0E-07 per year. 
In addition, the incremental conditional core damage probabilities 
(ICCDP) and incremental conditional large early release 
probabilities (ICLERP) are less than 5.0E-08. These increases/values 
meet the acceptance criteria in Regulatory Guide 1.174 and 1.177. 
Therefore, since the RTS and ESFAS will continue to perform their 
functions with high reliability as originally assumed, and the 
increase in risk as measured by CDF, LERF, ICCDP, ICLERP is within 
the acceptance criteria of existing regulatory guidance, there will 
not be a significant increase in the consequences of any accidents.
    2. The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed changes do not result in a change in the manner in 
which the RTS and ESFAS provide plant protection. The RTS and ESFAS 
will continue to have the same setpoints after the proposed changes 
are implemented. There are no design changes associated with the 
license amendment. The changes to Completion Times or increased 
bypass test times do not change any existing accident scenarios nor 
create any new or different accident scenarios. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed license amendment does not involve a significant 
reduction in margin of safety.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. Safety analysis acceptance criteria are 
not impacted. Redundant RTS and ESFAS trains are maintained, and 
diversity with regard to signals to provide reactor trip and 
engineered safety features actuation will be maintained. All signals 
credited as primary or secondary, and all operator action credited 
in the accident analyses will remain the same. The proposed changes 
will not result in plant operation in a configuration outside the 
design basis. The calculated impact on risk is insignificant and 
meets the acceptance criteria in Regulatory Guide 1.174 and 1.177. 
Although there was no attempt to quantify any positive human factors 
benefit due to increased Completion Times and bypass test times, it 
is expected that there would be a net benefit due to a reduced 
potential for spurious reactor trips and actuations associated with 
testing. Therefore, the proposed license amendment does not involve 
a significant reduction in margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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.
    Acting Section Chief: L. Raghavan, Acting.

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 28, 1998, as revised on April 
22, 1999, and April 27, 2000. This application was originally noticed 
on November 18, 1998 (63 FR 64122).
    Description of amendment request: The proposed amendments would 
modify the requirements associated with the control room and fuel 
handling building heating, ventilation, and air conditioning systems by 
adding an allowed outage time of 12 hours for a condition where 
multiple trains of the control room and fuel handling building heating, 
ventilation, and air conditioning systems are inoperable. The proposed 
amendments also include changes to make the required action for the 
affected ventilation actuation instrumentation consistent with the 
action for inoperable ventilation trains. In addition, the proposed 
amendments include minor administrative changes to remove an expired 
dated action and to provide consistency of terminology used in the 
Technical Specifications (TSs).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes do not involve an [significant] increase in 
the probability or consequences of an accident previously evaluated. 
The proposed changes consist of:
    (a) Assuring that the Specifications define consistent allowed 
outage times when the same safety function is addressed in multiple 
Specifications,
    (b) Allowing a system to remain inoperable when appropriately 
restrictive administrative controls are placed on operations that 
could result in a challenge to the safety function of the system,
    (c) Providing an appropriately short Allowed Outage Time for 
inoperability needed to permit required maintenance and testing that 
affects all trains of a system,
    (d) Redefining system operability and associated actions in a 
manner consistent with the system design and function,
    (e) Aligning a system to the actuated condition on the loss of 
an actuation channel,
    (f) Using consistent terminology throughout the Specifications.

[[Page 46017]]

    The proposed changes do not represent significant increases in 
the probability or consequences of an accident because:
    (a) The alignment of the action times between actuating system 
and actuated system operability requirements do not affect 
probability or consequences since inoperability of the actuated 
system has the same effect as inoperability of the actuating system. 
Since the changes proposed to the actuating system action times will 
reflect those of the actuated system action times, no change to the 
allowed outage time applicable to the safety function addressed and 
fulfilled by both, will occur.
    (b) Administrative controls to prevent the conduct of operations 
that could lead to a challenge to the safety function of the system 
when the actuation system is inoperable, assures that the design 
bases functions of the system will not be challenged. Therefore, the 
probability or consequences of an event previously identified have 
not been significantly changed.
    (c) Allowing up to 12 hours to recover from the inoperability of 
all 3 trains of Control Room Envelope HVAC [heating, ventilation, 
and air conditioning] or 2 or more trains of Fuel Handling Building 
HVAC does not represent a significant change to the probability of 
an accident. The inoperability of the Fuel Handling Building HVAC 
systems is not identified as a precursor to a design basis event. 
The inoperability of the Control Room Envelope HVAC is not a 
percursor to any event previously evaluated in the UFSAR [Updated 
Final Safety Analysis Report]. With respect to the PRA 
[probabilistic risk assessment] analysis for Control Room Envelope 
HVAC, the allowed outage time provides sufficient time to restore 
Control Room Envelope HVAC to the rooms serving the Reactor 
Protection System before any detrimental effects would occur or to 
place the plant in MODE 3 if Control Room Envelope HVAC could not be 
restored. The low likelihood of a design basis accident during the 
limited period of allowed inoperability of these systems does not 
involve a significant increase in the consequences of an accident. 
The proposed required actions to suspend all operations involving 
movement of spent fuel, and crane operations with loads over the 
spent fuel pool reduce the potential for accident initiation during 
the allowed outage time.
    (d) The redefinition of plant operability requirements into 
functional trains rather than individual components does not affect 
the required system functional operability. Therefore, this change 
does not involve an increase in the probability or consequences of 
an accident previously identified.
    (e) The alignment of the Control Room Envelope HVAC System to 
the same configuration it would be placed in from an actuation of 
the inoperable radiation monitoring channel places the system in the 
design condition. This alignment would result in maintaining the 
control room envelope pressurized and increases the protection 
afforded to the operators.
    (f) The change in terminology does not change any requirements 
or actions in the Specification. Therefore this change does not 
represent an increase in the probability or consequences of any 
accident previously evaluated.
    (g) Revising the applicability of Technical Specification ACTION 
b. in MODES 5 and 6 will add clarity to the specification and make 
it better reflect STP's three train design. The clarification 
provides some additional assurance that the system will perform as 
assumed in the analyses.
    Based on the above discussion, the individual changes do not 
represent an [significant] increase in the probability or 
consequences of any accident previously evaluated.
    In addition to the changes proposed to controls over Control 
Room Envelope HVAC, Fuel Handling Building HVAC, and associated 
actuation logic, an administrative change is proposed to remove the 
footnotes at the bottom of pages 3/4 3-28, 3/4 7-19, and 3/4 7-20. 
Since the footnotes no longer have meaning or relevance to the 
operation of the facility, their removal does not increase the 
probability or consequences of any accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes make the existing Specifications internally 
consistent, manually align a system to the actuated position, 
provide an alternative measure that assures [that] a safety function 
which is unavailable is not required to [be] perform[ed], provide an 
extended period of allowance for all trains of a system to be 
inoperable, and redefines system operability to reflect its 
functional design. The proposed changes do not introduce any new 
equipment into the plant or significantly alter the manner in which 
existing equipment will be operated. The limited allowed outage time 
of three inoperable Control Room Envelope HVAC systems has no 
detrimental effect on the operation of the Reactor Protection 
System. The systems affected by the proposed changes are not 
identified as contributing causal factors in design basis accidents; 
their function is to assist in mitigation of accidents postulated to 
occur. Since the proposed changes do not allow activities that are 
significantly different from those presently allowed, no possibility 
exists for a new or different kind of accident from those previously 
evaluated.
    In addition to the changes proposed to controls over Control 
Room Envelope HVAC, Fuel Handling Building HVAC, and associated 
actuation logic, an administrative change is proposed to remove the 
footnotes at the bottom of pages 3/4 3-28, 3/4 7-19, and 3/4 7-20. 
Since the footnotes no longer have meaning or relevance to the 
operation of the facility, their removal does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed changes do not involve a significant reduction in a 
margin of safety because the ability of the Fuel Handling Building 
HVAC and Control Room Envelope HVAC Systems to perform their 
function will be maintained. The margin of safety is defined by the 
ability of the systems to limit the release of radioactive materials 
and limit exposures to operators following a postulated design basis 
accident. The only aspect of the proposed change that can be 
postulated to have any effect on a margin of safety is the proposed 
allowance for all trains of Control Room Envelope HVAC or Fuel 
Handling Building HVAC to be inoperable for a limited period. The 
low probability of a design basis event that would require the 
system to perform its safety function during the limited period 
allowed by the proposed action assures that the change does not 
involve a significant change in a margin of safety. Therefore, the 
proposed changes do not significantly affect these operating 
restrictions and the margin of safety which support the ability to 
make and maintain the reactor in a safe shutdown and limit the 
release of radioactive material is not affected.
    Sufficient time is allowed to restore Control Room Envelope HVAC 
to the rooms serving the Reactor Protection System before any 
detrimental effects would occur or to place the plant in MODE 3 if 
Control Room Envelope HVAC could not be restored.
    Revising the applicability of Technical Specification 3.7.7 
ACTION b. in MODES 5 and 6 will add clarity to the specification, 
make it better reflect STP's three train design and provide greater 
assurance that desired margins are maintained.
    Suspending fuel movement and crane operations with loads over 
the spent fuel pool when all Fuel Handling Building or Control Room 
Envelope HVAC systems are inoperable prevents a Fuel Handling 
Accident from occurring, which maintains the margin of safety for 
this design event.
    In addition to the changes described above, an administrative 
change is proposed to remove the footnotes at the bottom of pages 3/
4 3-28, 3/4 7-19, and 3/4 7-20. Since these footnotes are no longer 
applicable to the facility, their removal cannot 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 standards of 10 CFR 50.92(c) are 
satisfied. The staff also reviewed the proposed change to provide 
consistency of terminology in the TSs for no significant hazards 
consideration. This proposed administrative change does not affect the 
design or operation of the facility and satisfies the three standards 
of 10 CFR 50.92(c). Therefore, the NRC staff proposes to determine that 
the request for amendments involves no significant hazards 
consideration.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: June 22, 2000.

[[Page 46018]]

    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to remove the applicability of 
core alteration requirements from those TS that are designed to 
mitigate the consequences of a fuel handling accident. The applicable 
TS bases would also be revised.
    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 revision eliminates requirements associated with 
core alterations for specifications that are intended to mitigate 
the consequences of a fuel handling accident (FHA). These functions 
will not impact accident generation because their function is to 
support mitigation of accidents and they are not considered to be 
the source of a postulated accident. The removal of these actions 
and surveillance requirements affects functions that are not 
necessary during core alterations because postulated events during 
these activities do not have the potential to result in major fuel 
cladding damage like that assumed for an FHA. Therefore, there is no 
adverse impact to nuclear safety by eliminating core alteration 
requirements for specifications that provide for the mitigation of 
an FHA.
    The proposed revision also clarifies the use of equivalent 
methods for isolation of containment penetrations. Equivalent 
isolation methods will maintain acceptable isolation capability for 
postulated conditions that could occur during the movement of 
irradiated fuel. This change does not alter the current intent or 
expectations for containment closure requirements during the 
movement of irradiated fuel and only serves to delineate other 
methods that provide an acceptable level of isolation. The status of 
penetration isolation methods during fuel movement does not impact 
the generation of an accident. This is based on these functions only 
providing a radiation barrier in the event of an FHA and not as a 
potential initiator for postulated accidents.
    Based on the previous discussions, the proposed revision does 
not alter any plant equipment or operating practices; therefore, the 
probability of an accident is not significantly increased. In 
addition, the consequences of an accident is not significantly 
increased by eliminating core alteration requirements for 
specifications that only support the mitigation of FHAs or by using 
equivalent isolation methods for containment penetrations. This is 
based on sufficient safety function capabilities being available for 
the mitigation of an FHA or other potential events that could occur 
during core alteration activities.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed allowance to eliminate core alteration requirements 
for FHA related specifications and utilize equivalent isolation 
methods for containment penetrations will not alter plant functions 
or equipment operating practices. The proposed elimination of core 
alteration requirements will not impact accident generation because 
these functions provide for FHA mitigation and are not postulated to 
be an initiator of postulated accidents. Containment penetration 
isolation methods are not considered to be the source of a 
postulated accident. Therefore, since plant functions and equipment 
are not altered and the availability of FHA mitigation functions and 
isolation of containment penetrations do not contribute to the 
initiation of postulated accidents, the proposed revision will not 
create a new or different kind of accident.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The elimination of core alteration requirements for 
specifications that provide mitigation functions for FHAs will not 
affect the ability of these functions to perform as necessary. This 
is based on postulated events during core alteration not having the 
potential to result in fuel cladding damage that is assumed for the 
FHA and therefore, not requiring functions necessary to mitigate the 
FHA event. The proposed revision will continue to provide acceptable 
provisions for activities that could result in an FHA or events 
postulated during core alterations to maintain the necessary margin 
of safety.
    The equivalent methods for containment penetration isolation 
provide the same level of isolation for conditions that may occur 
during fuel movement. Therefore, the equivalent isolation methods 
provide an acceptable barrier to the release of radiation as do the 
other listed methods and maintains the required margin of safety.
    Therefore, the margin of safety provided by the containment 
building penetration requirements and other specifications for the 
mitigation of FHAs is not significantly reduced by the proposed 
allowance to eliminate affected core alteration requirements or to 
use equivalent methods for containment penetration isolation.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H Knoxville, Tennessee 37902 .
    NRC Section Chief: Richard P. Correia.
    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: June 22, 2000.
    Description of amendment request: The proposed changes will modify 
the Technical Specifications in Sections 3.1.2.7, 3.1.2.8, 3.5.1, 
3.5.5, 3.6.2.2, 3.9.1, and associated Bases Sections to allow for an 
increase of boron in the refueling water storage tank (RWST), casing 
cooling tank (CCT), spent fuel pool, and safety injection accumulators 
(SIAs).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Increased boron concentration limits for the RWST, CCT, SIAs, 
and Spent Fuel Pool (SFP) will not increase the consequences of an 
accident previously evaluated. The increased boron concentration 
limits reduce the time to switchover from cold to hot leg 
recirculation, which will prevent boron precipitation in the reactor 
vessel following a loss of coolant accident (LOCA). The post-LOCA 
sump boron concentration limit is revised to ensure adequate post-
LOCA shutdown margin. The post-LOCA containment sump and quench 
spray (QS) pH remain within the limits specified in the Standard 
Review Plan. All other transients either were not impacted or were 
made less severe as a result of the increased boron concentrations.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed increase in boron concentration does not add new or 
different equipment to the facility, nor does this change the manner 
that plant equipment is being operated. Although the increased boron 
concentration requires procedure changes to ensure that cold to hot 
leg (reactor coolant loops) recirculating after an accident occurs 
earlier in the event, there are no changes to the methods utilized 
to respond to plant transients. The proposed Technical Specification 
changes do not alter instrumentation setpoints that initiate 
protective or mitigative actions. As a result, no new failure modes 
are being introduced. Therefore, the possibility for an accident of 
a different type than was previously evaluated in the Safety 
Analysis Report is not created.
    3. Does the change involve a significant reduction in a margin 
of safety.
    The LOCA considerations, including the recirculation switchover 
time, the post-LOCA sump boron concentration limit, and the quench 
spray and post-LOCA sump pH have been evaluated and found to be 
acceptable. The acceptance criteria of all non-LOCA transients 
continue to be met. Therefore, there is no significant reduction in 
the margin of safety in the accident analyses

[[Page 46019]]

impacted by boron concentration increases in the RWST, CCT, SIAs, 
and SFP.

    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: L. Raghavan, Acting.

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: June 22, 2000.
    Description of amendment request: The proposed changes modify the 
limiting conditions for operation, surveillance requirements, and the 
Bases for the North Anna Power Station (NAPS) Units 1 and 2 Technical 
Specifications 3.4.1.4, 3.4.1.6, 4.4.1.4, 4.4.1.6.1, and add 4.4.1.6.4 
to extend the drained reactor coolant loop verification time from 2 
hours to 4 hours prior to backfilling when returning the drained loop 
to service. This amendment request supersedes the August 4, 1999, 
request in its entirety (64 FR 48868, September 8, 1999).
    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:

    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 [revision to 
the] Technical Specification[s] establishes limiting conditions for 
operation and surveillance requirements for isolated loops backfill. 
Specifically, Technical Specifications requirements are being 
established to control the source of borated water for seal 
injection to the reactor coolant pumps (RCP) and to address 
reactivity control of an isolated and filled loop. The proposed 
controls ensure that the boron concentration of any source of water 
used for reactor coolant pump seal injection is greater than or 
equal to the boron concentration corresponding to the shutdown 
margin requirements for the applicable Mode. The proposed changes 
will establish consistent reactivity controls for isolated Reactor 
Coolant Systems (RCS) loops. The Bases [have] been revised to 
further discuss the additional controls for the loop backfill 
evolution. Adequate Technical Specifications controls have been 
established to ensure that an uncontrolled positive reactivity 
addition does not occur during a loop backfill evolution. The 
proposed changes will ensure that an inadvertent/undetected positive 
reactivity addition does not occur. The following is provided to 
support this conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed Technical Specification limiting conditions for 
operation and surveillance requirements ensure that the initiation 
of seal injection in order to allow a partial vacuum to be 
established in an isolated and drained loop will not create the 
potential for an inadvertent/undetected introduction of under-
borated water into an isolated loop prior to returning the isolated 
loop to service. The proposed Technical Specification controls 
prevent any additions of makeup or seal injection that would violate 
the existing shutdown margin requirements for the active portion of 
the RCS. Thus, adequate Technical Specification controls are 
established to preclude an inadvertent/undetected boron dilution 
event. Therefore, there is no increase in the probability or 
consequences of any accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    There are no modifications to the plant as a result of the 
changes. The proposed Technical Specification Limiting Conditions 
for Operation and Surveillance Requirements ensure that the 
initiation of seal injection will not create an undetected positive 
reactivity addition. No new accident or event initiators are created 
by the initiation of seal injection for the RCP in the isolated loop 
in order to establish a partial vacuum in that isolated and drained 
loop. Therefore, the proposed changes do not create the possibility 
of any accident or malfunction of a different type previously 
evaluated.
    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 safety analyses 
assumptions. Rather, the proposed changes acknowledge the 
establishment of seal injection for the RCP in the isolated and 
drained loop as a prerequisite for the vacuum-assisted backfill 
technique. The proposed Technical Specification Limiting Conditions 
for Operation and Surveillance Requirements ensure that the 
initiation of seal injection in order to allow a partial vacuum to 
be established in an isolated and drained loop will not create the 
potential for an inadvertent/undetected introduction of under-
borated water into an isolated loop prior to returning the isolated 
loop to service. Adequate Technical Specifications controls are 
established to preclude an inadvertent/undetected boron dilution 
event. In addition, the proposed controls prevent any additions of 
makeup or seal injection that would violate the existing shutdown 
margin requirements for the active portion of the Reactor Coolant 
System. 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 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: L. Raghavan, Acting.

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

    Date of amendment request: June 27, 2000 (WM 00-0026).
    Description of amendment request: The proposed amendment would 
revise Appendix C, ``Antitrust Conditions for Kansas Gas and Electric 
Company [KGE],'' for the Wolf Creek Generating Station (WCGS) operating 
license. The revisions would (1) state that the specific conditions 
applicable to Kansas Electric Power Cooperative, Inc. (KEPCo) do not 
restrict its rights, or the duties of KGE, under other license 
conditions, (2) define ``KGE members in licensee's service area'' in 
the appendix to include all KEPCo members with facilities in Western 
Resources' and KGE's combined service area, (3) delete license 
conditions restricting KEPCo's use of the power from WCGS, (4) remove 
out-of-date conditions, and (5) update conditions to be consistent with 
the terms and conditions of Western Resources' Federal Energy 
Regulatory Commission (FERC) open access transmission tariff. Western 
Resources is the parent company of KGE.
    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 change merely revises the KGE Antitrust Conditions 
in the Wolf Creek Generating Station Facility Operating License. The 
proposed change is considered an administrative change and does not 
modify, add, delete, or relocate any technical requirements of the 
Technical Specifications. As such, the administrative changes do not 
affect initiators of analyzed events or assumed mitigation of 
accident or transient events. Therefore, this change does not

[[Page 46020]]

involve a significant increase in the probability or consequences of 
an accident previously analyzed.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The proposed 
change will not impose any new [requirements] or eliminate any old 
requirements. Thus, the change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change will not reduce a margin of safety because 
there is no effect on any safety analyses assumptions. The changes 
are administrative in nature. Therefore, the 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: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Section Chief: Stephen Dembek.

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

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

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

    Date of application for amendment: June 7, 2000.
    Brief description of amendment: The proposed amendment would revise 
Section 3.10.8, ``SHUTDOWN MARGIN (SDM) Test--Refueling,'' of the 
Technical Specifications (TS), correcting an administrative error 
introduced when Amendment No. 91 (converting the TS to the Improved TS 
format) was processed.
    Date of publication of individual notice in Federal Register: June 
16, 2000 (65 FR 37807).
    Expiration date of individual notice: July 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).

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

    Date of application for amendments: May 26, 1999, as supplemented 
March 31, 2000.
    Brief description of amendments: The amendments revise Technical 
Specification 3.3.1, ``Reactor Protective System (RPS) 
Instrumentation--Operating,'' to change the allowable values for two of 
the trip setpoints. The change will reduce spurious reactor trip 
hazards associated with these setpoints while maintaining plant 
protection.
    Date of issuance: July 6, 2000.
    Effective date: July 6, 2000, to be implemented within 60 days. For 
surveillance requirements associated with the revised allowable values 
for functions 12 and 13 in Technical Specification Table 3.3.1-1, the 
first performance is due at the end of the first surveillance interval 
that began on the date the surveillance was last performed prior to the 
date of implementation of these amendments.
    Amendment Nos.: Unit 1-126, Unit 2-126, Unit 3-126.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 17, 2000 (65 FR 
31355).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 6, 2000.
    No significant hazards consideration comments received: No.

Baltimore Gas and Electric Company, Docket Nos. 50-317, 50-318, and 72-
8, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, and 
Independent Spent Fuel Storage Installation, Calvert County, Maryland

    Date of application for amendment: February 29, 2000, as 
supplemented April 7, April 27, May 2, May 19, and June 20, 2000.
    Brief description of amendment: These amendments conform the 
licenses to reflect the transfer of Operating Licenses Nos. DPR-53 and 
DPR-69 for the Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and 
Materials License No. SNM-2505 for the Calvert Cliffs Independent Spent 
Fuel Storage Installation held by Baltimore Gas and Electric Company to 
Calvert Cliffs Nuclear Power Plant, Inc.
    Date of Issuance: June 30, 2000.
    Effective date: As of the date of issuance to be implemented within 
45 days.

[[Page 46021]]

    Amendment No.: 237 and 211.
    Facility Operating License No. DPR-53, DPR-69: Amendments revised 
the Operating Licenses, and Materials License No. SNM-2505 and the 
Materials License Technical Specifications.
    Date of initial notice in Federal Register: May 4, 2000 (65 FR 
25963)
    The April 7, April 27, May 2, May 19, and June 20, 2000, 
supplements did not expand the scope of the initial application as 
originally noticed.
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated June 30, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 29, 1999.
    Brief description of amendment: The amendment revises items 1.a, 
2.a, 4.a, and 5.a of Technical Specification Table 3.3.5.1-1, 
``Emergency Core Cooling System Instrumentation,'' to change the 
reactor vessel water level--level 1 allowable value.
    Date of issuance: July 13, 2000.
    Effective date: July 13, 2000, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 166.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46431) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 13, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 20, 1999.
    Brief description of amendment: Incorporates 16 improvements 
(identified by Technical Specifications Task Force (TSTF) numbers) to 
the Improved Standard Technical Specifications, NUREG-1434 (for General 
Electric model Boiling Water Reactor/6 (BWR/6) plants such as Grand 
Gulf Nuclear Station (GGNS)), that was part of the basis for the 
current improved Technical Specifications for GGNS that were issued in 
Amendment 120 dated February 21, 1995. The 17 improvements are the 
following TSFTs: 2, 5, 17, 18, 32, 33, 38, 45, 60, 104, 118, 153, 163, 
166, 278, and 279. The licensee withdrew its request to incorporate 
TSTF-9 into the TSs.
    Date of issuance: June 30, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 142.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73089).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 30, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 18, 1999, as supplemented by 
letters dated May 16, 2000, and June 1, 2000.
    Brief description of amendment: The amendment 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 incorporated.
    Date of issuance: July 6, 2000.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 165.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 9, 2000 (65 FR 
6407). The May 16, 2000, and June 1, 2000, supplements did not expand 
the scope of the application as noticed or change the proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 6, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 15, 1999.
    Brief description of amendments: The amendments change the 
Technical Specifications (TSs) surveillance frequency for the quench 
and recirculation spray system nozzle air flow test. The amendments 
also change terminology in the TS action statement for the TS axial 
flux difference, and make other miscellaneous editorial and format 
changes.
    Date of issuance: July 11, 2000.
    Effective date: As of the date of its issuance and shall be 
implemented within 60 days.
    Amendment Nos.: 231 and 111.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 17, 1999 (64 
FR 62708) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 11, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 27, 2000, as 
supplemented May 30, 2000.
    Brief description of amendment: The amendment will modify the 
action statement for Technical Specification (TS) 3/4.7.11, ``Ultimate 
Heat Sink,'' to permit Unit 2 to remain in operation with the ultimate 
heat sink water temperature greater than 75 deg. F and less than 
77 deg. F, for a period of up to 12 hours provided the water 
temperature is verified below 77 deg. F at least once per hour. This is 
a one-time change during the summer period and will expire after 
October 15, 2000, and revert back to the original TS action statement.
    Date of issuance: July 10, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 247.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 22, 2000 (65 FR 
15382) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 10, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 19, 1999.
    Brief description of amendments: The amendments revise paragraph 
2.C.(4) of the Operating Licenses related to the fire

[[Page 46022]]

protection program at Prairie Island, Units 1 and 2. Specifically, the 
proposed amendments would (1) remove reference to two NRC safety 
evaluation reports (SEs) that are no longer applicable to the fire 
protection program at Prairie Island and (2) correct the date of one 
SE.
    Date of issuance: July 11, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 150 and 141.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 28, 2000 (65 FR 
25001).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 11, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 12, 1999, as supplemented 
July 7, 2000.
    Brief description of amendments: The amendments revise several 
Technical Specification (TS) sections in order to relocate shutdown 
margin requirements to the Core Operating Limits Report and to ensure 
that the TS requirements are consistent with the dilution analysis in 
the Updated Safety Analysis Report.
    Date of issuance: July 11, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 151 and 142.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 28, 2000 (65 FR 
24999). The July 7, 2000, supplemental letter provided clarifying 
information that was within the scope of the original application 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 July 11, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 17, 1999, as 
supplemented April 6, 2000.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3.1.A.1, ``Reactor Coolant Loops and Coolant 
Circulation,'' to (1) establish required actions and a 72 hour time 
limit for operation with the reactor coolant system (RCS) average 
temperature above 350  deg.F and no reactor coolant pumps (RCPs) 
running, (2) extend from 6 hours to 12 hours the time within which the 
RCS average temperature must be reduced to below 350  deg.F if 72 hours 
are exceeded and no RCPs are restored to operability and operation, and 
(3) extend the time limit for operations with no RCPs running from 1 
hour to 12 hours for situations where the RCPs are stopped as a result 
of preplanned work activities.
    Date of issuance: July 14, 2000
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 152 and 143
    Facility Operating License Nos. DPR-42 and DPR-60:  Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 29, 1999 (64 
FR 66670).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 14, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 15, 1999, as 
supplemented February 7, March 24, April 28, May 4, and May 30, 2000.
    Brief description of amendments: The amendments conform the 
operating licenses for each of the units to reflect the transfer of the 
operating licenses, to the extent held by PP&L, Inc., to PPL 
Susquehanna, LLC.
    Date of issuance: July 1, 2000
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment Nos.: 188 and 162.
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the license.
    Date of initial notice in Federal Register: March 3, 2000 (65 FR 
11611). The March 24, April 28, May 4, and May 30, 2000, letters 
provided clarifying information. The Commission's related evaluation of 
the amendments is contained in a Safety Evaluation dated June 6, 2000.

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

    Date of application for amendments: April 14, 1999, as supplemented 
on March 2, 2000.
    Brief description of amendments: The license amendment revises 
Technical Specification (TS) Section 3/4.9.12, ``Fuel Handling Area 
Ventilation System,'' and provides greater consistency between the two 
Salem units, removes inappropriate and invalid surveillance 
requirements (SR), and clarifies the Bases. The revised TS Section 3/
4.9.12 will require that the high efficiency particulate air (HEPA) and 
charcoal filters to be in service prior to moving irradiated fuel in 
the Fuel Handling Building. This will be accomplished by the addition 
of a new SR 4.9.12.b. The new SR allows the licensee to eliminate an 
automatic actuation feature from the Fuel Handling Area Ventilation 
system control circuit, as well as the requirement to test that 
feature. The new surveillance will also require verification of system 
line up every 24 hours during fuel movement or crane operation to 
ensure system flow through the HEPA-charcoal filter train.
    Date of issuance: June 14, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 231 & 211.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 2, 1999 (64 FR 
29715). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 14, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 17, 1999.
    Brief description of amendment: The proposed changes would revise 
the required minimum contained volume of the condensate storage tank 
from 172,700 gallons of water to 179,850 gallons of water.
    Date of issuance: July 7, 2000.
    Effective date: July 7, 2000.
    Amendment No.: 145.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.

[[Page 46023]]

    Date of initial notice in Federal Register: June 16, 1999 (64 FR 
32290).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 7, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 2, 1999, as 
supplemented May 16 and June 16, 2000 (PCN-506).
    Brief description of amendments: These amendments approve changes 
to Technical Specifications, Section 5.0, ``Administrative Controls,'' 
and the Environmental Protection Plan.
    Date of issuance: July 7, 2000.
    Effective date: July 7, 2000, to be implemented within 30 days of 
issuance.
    Amendment Nos.: Unit 2-168; Unit 3-159.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications and the Environmental Protection 
Plan.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73096). The May 16 and June 16, 2000, letters provided additional 
information and clarifications that were within the scope of the 
original Federal Register notice and did not change the staff's initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 7, 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 18, 1999, as 
supplemented May 11, 2000.
    Brief description of amendment: The amendment revises the Technical 
Specifications to require a revised activated charcoal testing 
methodology in accordance with the guidance provided by Generic Letter 
99-02, ``Laboratory Testing of Nuclear Grade Activated Charcoal.''
    Date of Issuance: July 11, 2000.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 189.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 17, 1999 (64 
FR 62716).
    The May 11, 2000, supplement did not expand the scope of the 
application as initially noticed, or change the proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of this amendment is contained in a Safety Evaluation dated 
July 11, 2000.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 19th day of July 2000.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-18771 Filed 7-25-00; 8:45 am]
BILLING CODE 7590-01-P