[Federal Register Volume 66, Number 7 (Wednesday, January 10, 2001)]
[Notices]
[Pages 2010-2031]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-596]


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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 December 18, 2000, through December 29, 
2000. The last biweekly notice was published on December 27, 2000 (65 
FR 81907).

[[Page 2011]]

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 and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC's Public Document 
Room, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. The filing of requests for a hearing 
and petitions for leave to intervene is discussed below.
    By February 9, 2001, 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, located at One 
White Flint North, 11555 Rockville Pike (first Floor), Rockville, 
Maryland 20852. Publicly available records will be accessible and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room). If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission,

[[Page 2012]]

Washington, DC 20555-0001, Attention: Docketing and Services Branch, or 
may be delivered to the Commission's Public Document Room, located at 
One White Flint North, 11555 Rockville Pike (first floor), Rockville, 
Maryland 20852, by the above date. A copy of the petition should also 
be sent to the Office of the General Counsel, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, and to the attorney for the 
licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 
Publicly available records will be accessible 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. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of amendment request: October 20, 2000.
    Description of amendment request: The proposed amendment revises 
the Technical Specification (TS) surveillance interval for emergency 
diesel generator (EDG) maintenance from annually to 2 years. This 
interval is in conformance with guidelines of the Fairbanks Morris 
Owner's Group and the EDG manufacturer.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The changes do not affect the ability of the Emergency Diesel 
Generators (EDGs) to mitigate the consequences of an accident, 
including the loss of coolant accident coupled with loss of offsite 
power accident, which would be considered the most demanding on EDG 
System and components. A reduction in the number of diesel outages 
will also reduce the possibility of introducing problems resulting 
from human error or foreign material intrusion. Extending the 
maintenance interval should reduce the two-year unavailability from 
about 2% to about 1.4%. This is an approximate 30% reduction in 
unavailability. An extension of the outage inspection frequency to 
24 months will result in increased EDG availability to mitigate the 
consequences of a potential accident. When this program is taken in 
its entirety, the extended maintenance intervals, coupled with the 
defined enhancements, is judged to result in an overall increase in 
Emergency Diesel Generator availability and reliability. The 
surveillance testing requirements of Technical Specifications 
Section 4.6.1a&b will continue to verify the operability and 
reliability of the EDG System. Therefore, operation of the facility 
in accordance with the proposed amendment would not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    The Emergency Diesel Generator System is not an accident 
initiator. The operation, testing, and design of the Emergency Power 
System (including the Emergency Diesel Generators) is not being 
changed. The maintenance inspection interval is being expanded from 
annual to two years and will improve availability and enhance 
reliability. Plant design requires the full load capability of one 
Emergency Diesel Generator to support accident loads and the 
respective emergency electrical busses. Performance of the 
maintenance inspection on the extended interval will not have an 
adverse affect on the ability of the Emergency Diesel Generators to 
meet the design response criteria or contribute to the occurrence or 
the consequences of an accident. The proposed changes do not involve 
any physical design or operation changes that could create a 
malfunction extending beyond an individual Emergency Diesel 
Generator, nor does it increase the potential for a common-mode 
Emergency Diesel Generator failure. Therefore, operation of the 
facility in accordance with the proposed amendment would not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in the margin of 
safety.
    The change of the maintenance inspection frequency and the 
detailed programmatic changes that implement the Fairbanks Morse 
Owners Group recommendations, will increase the availability and 
reliability of the Emergency Diesel Generators. Based on improving 
the availability and reliability, the margin of safety will actually 
be enhanced. The amount of time the Emergency Diesel Generators are 
out-of-service during on-line maintenance will decrease, thereby 
reducing the number of plant operating hours that the unit is 
exposed to a single mode failure. Therefore, operation of the 
facility in accordance with the proposed amendment would not involve 
a significant reduction in the margin of safety.
    Based on the analysis provided herein, the proposed change meets 
the requirements of 10 CFR 50.92(c) and involves no 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: Edward J. Cullen, Jr., Esq., PECO Energy 
Company, 2301 Market Street, S23-1, Philadelphia, PA 19103.
    NRC Section Chief: Marsha Gamberoni.

Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-318, Calvert 
Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

    Date of amendment request: September 14, 2000 as supplemented on 
December 21, 2000.
    Description of amendment request: The proposed amendment 
incorporates the changes described below into the Technical 
Specifications (TSs) for Calvert Cliffs Unit 2. Calvert Cliffs Nuclear 
Power Plant, Inc. (the licensee) also requested an exemption for 
Calvert Cliffs Unit 2 from the requirements of 10 CFR 50.46, 10 CFR 
50.44, and 10 CFR Part 50, Appendix K.
    The exemption and TS change will allow a lead fuel assembly (LFA) 
with a limited number of fuel rods clad with advanced zirconium-based 
alloys to be inserted into the core during the next Unit 2 refueling 
outage, scheduled to begin in March 2001. This LFA was approved to be 
inserted into Unit 1 Cycle 15. Because of concerns with corrosion 
performance, all of the Anikuloy, Alloy C, and Zr-2P clad rods were 
removed from this LFA and replaced with OPTIN and Alloy E rods from 
another LFA. Due to the length of time needed to perform this activity 
and the duration of the Unit 1 outage, it was not possible to reinsert 
this LFA into Unit 1 for Cycle 15 operation. Therefore, the licensee is 
requesting approval to insert this assembly into Unit 2 for Cycle 14 
operation.
    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:


[[Page 2013]]


    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    Supporting analyses indicate that since the LFA will be placed 
in a non-limiting location, the placement scheme and the similarity 
of the advanced alloys to zircaloy-4 will assure that the behavior 
of the fuel rods with these alloys are bounded by the fuel 
performance and safety analyses performed for the zircaloy-4 clad 
fuel rods currently in the Unit 2 Core. Therefore, the addition of 
these advanced claddings does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different [kind] 
of accident from any accident previously evaluated.
    The proposed change does not add any new equipment, modify any 
interfaces with existing equipment, change the equipment's function, 
or change the method of operating the equipment. The proposed change 
does not affect normal plant operations or configuration. Since the 
proposed change does not change the design, configuration, or 
operation, it could not become an accident initiator.
    Therefore, the proposed change does not create the possibility 
of a new or different [kind] of accident from any previously 
evaluated.
    3. Would not involve a significant reduction in the margin of 
safety.
    Supporting analyses indicate that since the LFA will be placed 
in a non-limiting location, the placement scheme and the similarity 
of the advanced alloys to zircaloy-4 will assure that the behavior 
of the fuel rods with these alloys are bounded by the fuel 
performance and safety analyses performed for the zircaloy-4 clad 
fuel rods currently in the Unit 2 Core. Therefore, the addition of 
these advanced claddings does not involve a significant reduction in 
the margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

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

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

    Date of amendment request: October 20, 2000.
    Description of amendment request: The amendments would revise the 
Technical Specification 3.7.10, ``Control Room Area Ventilation System 
(CRAVS).'' The primary purpose of the request is to eliminate the 
requirement for the CRAVS high chlorine protection function. Duke 
Energy Corporation (the licensee) indicated that a chlorine detection 
system with safety related detectors and automatic CRAVS intake 
isolation capability is no longer needed at Catawba. In addition, the 
licensee is also requesting NRC approval to allow the use of non-safety 
related detectors and to delete the automatic intake isolation 
capability. Finally, the amendments would also revise the Bases for the 
CRAVS to more clearly describe the system function and to make other 
clarifying changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

First Standard

    Implementation of this amendment would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. Neither the CRAVS, nor its automatic control 
room intake isolation function on a high chlorine condition is 
capable of initiating any accident. The CRAVS is responsible for 
maintaining an acceptable environment in the control room during 
normal operation and accident conditions. The CRAVS will continue to 
function as designed to provide this environment in accordance with 
all applicable TS. Following implementation of this amendment, 
Catawba plans to pursue elimination of the automatic intake 
isolation capability. This will not affect the system's ability to 
maintain an acceptable control room environment during and following 
an accident. No other design changes to the system are being made. 
It has been shown that the quantity of gaseous chlorine used at 
Catawba is less than the threshold stated in applicable Regulatory 
Guides. Hence, there is no control room habitability issue due to 
chlorine. Therefore, there will be no impact on any accident 
probabilities or consequences.

Second Standard

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

Third Standard

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

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

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

    Date of amendment request: December 11, 2000.
    Description of amendment request: The proposed amendment would 
modify the existing Minimum Critical

[[Page 2014]]

Power Ratio (MCPR) Safety Limit contained in Technical Specification 
2.1.1.2. Specifically, the change modifies the MCPR Safety Limit value, 
as calculated by Global Nuclear Fuel, by increasing the limit for two 
recirculation loop operation from 1.09 to 1.10.
    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.
    Per the Perry Nuclear Power Plant (PNPP) Updated Safety Analysis 
Report (USAR) Section 4.2.1, the fuel system design bases are 
provided in the General Electric Standard Application for Reactor 
Fuel (GESTAR II). The Minimum Critical Power Ratio (MCPR) Safety 
Limit is one of the limits used to protect the fuel in accordance 
with the design basis. The NRC-approved MCPR Safety Limit 
calculations establish margin to the onset of transition boiling. 
The basis of the MCPR Safety Limit calculation remains the same, 
ensuring that greater than 99.9% of all fuel rods in the core avoid 
transition boiling. These NRC-approved calculations were used to 
determine the proposed limit, therefore there is not an increase in 
the probability of transition boiling. Also, the change does not 
result in any physical plant modifications or physically affect any 
plant components. Therefore, no individual precursors of an accident 
are affected. As a result, there is no increase in the probability 
of occurrence of a previously analyzed accident.
    The fundamental sequences of accidents and transients have not 
been altered. The Safety Limit MCPR is established to avoid fuel 
damage in response to anticipated operational occurrences. 
Compliance with a MCPR safety limit greater than or equal to the 
calculated value will ensure that less than 0.1% of the fuel rods 
will experience boiling transition. This in turn ensures fuel damage 
does not occur following transients due to excessive thermal 
stresses on the fuel cladding. The MCPR Operating Limits are set 
higher (i.e., more conservative) than the Safety Limit such that 
potentially limiting plant transients prevent the MCPR from 
decreasing below the MCPR Safety Limit during the transient. 
Therefore, there is no impact on any of the limiting USAR Appendix 
15B transients. The radiological consequences remain the same as 
previously stated in the USAR. Therefore, the consequences of an 
accident do not increase over previous evaluations in the USAR.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The MCPR Safety Limit basis is preserved, which is to ensure 
that transition boiling does not occur in at least 99.9% of the fuel 
rods in the core as a result of the limiting postulated transient. 
The value is calculated in accordance with GESTAR II. The GESTAR II 
analyses have been accepted by the NRC as comprehensive for ensuring 
that fuel designs will perform within acceptable bounds. The MCPR 
Safety Limit is one of the limits established to ensure the fuel is 
protected in accordance with the design basis. The function, 
location, operation, and handling of the fuel remain unchanged. No 
changes in the design of the plant or the method of operating the 
plant are associated with this revised safety limit value. 
Therefore, no new accident precursors are created due to this 
change. As a result, no new or different kind of accident from any 
previously evaluated is created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    This change revised the PNPP MCPR Safety Limit value. The new 
MCPR Safety Limit value does not alter the design or function of any 
plant system, including the fuel. The new MCPR Safety Limit value 
was calculated using NRC-approved methods described in GESTAR II. 
The MCPR Safety Limit value is consistent with GESTAR II, the NRC 
Safety Evaluation of GESTAR II, and the Technical Specification 
Bases (Section 2.1.1.2) for the MCPR Safety Limit. Use of these 
methods satisfies the fuel design safety criteria that less than 
0.1% of the fuel rods are predicted to experience transition boiling 
if the safety limit is not violated. Therefore, enforcing the new 
value for the MCPR Safety Limit does not involve a reduction in the 
margin of safety.

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

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

    Date of amendment request: October 23, 2000.
    Description of amendment request: The proposed license amendments 
would revise Table 3.3-5, Accident Monitoring Instrumentation, and 
Table 4.3-4, Accident Monitoring Instrumentation Surveillance 
Requirements. The revision would delete reference to the containment 
hydrogen monitors from the Accident Monitoring Instrumentation. 
Additionally, the proposed amendments would delete Technical 
Specification (TS) 3/4.6.5, Combustible Gas Control--Hydrogen Monitors, 
and TS 3/4.6.6, Post Accident Containment Vent System.
    In addition, the licensee requested an exemption from the 
requirements of 10 CFR 50.44, ``Standards for Combustible Gas Control 
Systems in Light-Water-Cooled Power Reactors and 10 CFR Part 50, 
Appendix E, Section VI, ``Emergency Response Data System.'' The purpose 
of the exemption is to remove the requirements for hydrogen-control 
systems from the Turkey Point (TP) Units 3 and 4 design basis. 
Moreover, the licensee's submittal requested a change to the 
Confirmatory Order dated March 14, 1983, and revised by NRC letter 
dated October 5, 2000, confirming TP Units 3 and 4 commitments related 
to NUREG-0737, post-TMI requirements. Specifically, the licensee 
requests deletion of the commitment to NUREG-0737, Item II.F.I, Item 6, 
Containment Hydrogen Monitor requirements. The exemption request and 
the revision to the Confirmatory Order will be evaluated separately 
from the proposed license amendments.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. The Containment Combustible Gas Control System is composed 
of two hydrogen monitors, the Post-Accident Containment Vent System, 
and a leased hydrogen recombiner. Hydrogen control components are 
not considered to be accident initiators. Therefore, this change 
does not increase the probability of an accident previously 
evaluated.
    The Containment Combustible Gas Control System is provided to 
ensure that the hydrogen concentration is maintained below 4.0% so 
that containment integrity is not challenged following a design 
basis Loss Of Coolant Accident (LOCA). Existing analysis show that 
the hydrogen concentration will not reach 4.0% for at least 12 days 
after a design basis LOCA. Containment failure due to hydrogen 
combustion without the Post-Accident Containment Vent System and 
backup hydrogen recombiner is not credible based on the results of 
the Turkey Point Units 3 and 4 Individual Plant Examination study. 
Therefore, this change does not increase the consequences of 
accidents previously evaluated.
    Removal of the existing requirements for hydrogen control will 
reduce the consequences of postulated accidents by eliminating Post-
Accident Containment Vent System releases, and by eliminating 
potential unfiltered release paths during operation of the hydrogen 
recombiner.

[[Page 2015]]

    Removal of the existing requirements for hydrogen control will 
also allow elimination of the Emergency Operating Procedure (EOP) 
steps for hydrogen control and hence simplify migration through the 
EOPs. This would have a positive impact on public health risk by 
reducing the probability of operator error during potential 
accidents and hence reduce the core damage frequency. In addition, 
approval of these amendment requests will minimize the potential for 
actuation of the Post-Accident Containment Vent System and/or the 
backup hydrogen recombiner during severe accidents. The changes 
described in this request result in an overall decrease in risk.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    No. This proposed change does not change the design or 
configuration of the plant beyond the containment Combustible Gas 
Control System. Hydrogen generation following a design basis LOCA 
has been evaluated in accordance with regulatory requirements. 
Deletion of the containment Combustible Gas Control System from the 
technical specifications does not alter the hydrogen generation 
processes post-LOCA. The consideration of hydrogen generation will 
no longer be included in the design basis of Turkey Point Units 3 
and 4. Therefore, this change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    No. The Containment Combustible Gas Control System is provided 
to ensure that the hydrogen concentration is maintained below 4.0% 
so that containment integrity is not challenged following a design 
basis Loss Of Coolant Accident (LOCA). Existing analysis show that 
the hydrogen concentration will not reach 4.0% for at least 12 days 
after a design basis LOCA. Containment failure due to hydrogen 
combustion without the Post-Accident Containment Vent System and 
backup hydrogen recombiner is not credible based on the results of 
the Turkey Point Units 3 and 4 Individual Plant Examination study. 
Therefore, this change does not result in a reduction in a margin of 
safety.
    The changes proposed in these amendment requests result in a 
reduction in risk. Removal of the existing requirement for a 
containment Combustible Gas Control System will, by eliminating the 
EOP steps for hydrogen control, result in lower operator error 
probabilities. In addition, approval of these amendment requests 
will minimize the potential for actuation of the Post-Accident 
Containment Vent System and/or the backup hydrogen recombiner during 
severe accidents. Therefore, this change involves an increase in 
safety, not 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: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

GPU Nuclear Corporation and Saxton Nuclear Experimental Corporation 
(SNEC), Docket No. 50-146, Saxton Nuclear Experimental Facility (SNEF), 
Bedford County, Pennsylvania

    Date of amendment request: November 30, 2000.
    Description of amendment request: The proposed amendment would 
change the name in the license of GPU Nuclear Corporation to GPU 
Nuclear, Inc.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:

    The NRC has previously determined that similar amendments 
reflecting this name change have involved no significant hazards 
consideration. See 62 Fed. Reg. 4341, 4350 (1997) and 62 Fed. Reg. 
59912, 59915 (1997).
    Consistent with these prior NRC determinations, GPU Nuclear has 
determined that the License Amendment involves no significant 
hazards considerations as defined in 10 CFR 50.92.
    1. The proposed changes to the Saxton License do not involve a 
significant increase in the probability of occurrence or 
consequences of an accident or malfunction of equipment important to 
safety previously analyzed in the safety analysis report. The 
changes have no impact on plant operations or the release of 
radioactive materials.
    2. The proposed changes to the Saxton License will not create 
the possibility for an accident or malfunction of a different type 
than any previously evaluated in the safety analysis report because 
no plant configuration or operation changes are involved.
    3. The changes will not involve a significant reduction in the 
margin of safety as defined in the basis of any technical 
specification for Saxton because no change to operational limits 
will be made.

    The NRC staff has reviewed the analysis of the licensees 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 the Licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts, and Trowbridge, 2300 N Street, N.W., Washington, D.C. 
20037.
    NRC Branch Chief: Ledyard B. Marsh.

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: December 5, 2000.
    Description of amendment request: The proposed amendment would 
reformat the Technical Specifications to be more consistent with the 
proposed Improved Standard Technical Specifications applicable to 
permanently shutdown and defueled facilities. The proposed changes also 
modify the specifications to better reflect the decommissioned status 
of Millstone Nuclear Power Station, Unit 1. Other changes relocate 
requirements out of the Technical Specifications to other controlled 
license basis documents, consistent with the Improved Standard 
Technical Specifications and Nuclear Regulatory Commission (NRC) 
guidance.
    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:

Administrative Changes (``A.x'' Labeled Comments/Discussions)

    In accordance with the criteria set forth in 10 CFR 50.92, 
Northeast Nuclear Energy Company (NNECO) has evaluated this proposed 
Technical Specifications change and determined it does not represent 
a significant hazards consideration. The following is provided in 
support of this conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change involves reformatting, renumbering, and 
rewording the existing Technical Specifications. The reformatting, 
renumbering, and rewording process involves no technical changes to 
the existing Technical Specifications. As such, this change is 
administrative in nature and does not impact initiators of analyzed 
events or assumed mitigation of accident or transient events. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed)

[[Page 2016]]

or changes in methods governing normal plant activities. The 
proposed change will not impose any new or eliminate any old 
requirements. Thus, this change 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 change will not reduce a margin of safety because 
it has no impact on any safety analyses assumptions. This change is 
administrative in nature. Therefore, the change does not involve a 
significant reduction in a margin of safety.

Technical Changes--More Restrictive (''M.x'' Labeled Comments/
Discussions)

    In accordance with the criteria set forth in 10 CFR 50.92, NNECO 
has evaluated this proposed Technical Specifications change and 
determined it does not represent a significant hazards 
consideration. The following is provided in support of this 
conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change provides more stringent requirements for 
operation of the facility. These more stringent requirements do not 
result in operation that will increase the probability of initiating 
an analyzed event and do not alter assumptions relative to 
mitigation of an accident or transient event. The more restrictive 
requirements continue to ensure process variables, structures, 
systems, and components are maintained consistent with the safety 
analyses and licensing basis. Therefore, this change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    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 the methods governing normal plant operation. The 
proposed change does impose different requirements. However, these 
changes are consistent with the assumptions in the safety analyses 
and licensing basis. Thus, this change 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 imposition of more restrictive requirements either has no 
impact on or increases the margin of plant safety. As provided in 
the discussion of the change, each change in this category is by 
definition, providing additional restrictions to enhance plant 
safety. The change maintains requirements within the safety analyses 
and licensing basis. Therefore, this change does not involve a 
significant reduction in a margin of safety.

``Generic'' Less Restrictive Changes: Relocating Details to Other Plant 
Controlled Documents (``LA.x'' Labeled Comments/Discussions)

    In accordance with the criteria set forth in 10 CFR 50.92, NNECO 
has evaluated this proposed Technical Specifications change and 
determined it does not represent a significant hazards 
consideration. The following is provided in support of this 
conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change relocates certain details from the Technical 
Specifications to the TRM [Technical Requirements Manual] or the 
Millstone [Nuclear Power Station (Millstone),] Unit 1 Northeast 
Utilities Quality Assurance Program (NUQAP). The TRM will be 
maintained in accordance with 10 CFR 50.59. The NUQAP is subject to 
the change control provisions 10 CFR 50.54(a). Since any changes to 
the TRM or NUQAP will be evaluated per the requirements of 10 CFR 
50.59 or 10 CFR 50.54(a) respectively, no increase (significant or 
insignificant) in the probability or consequences of an accident 
previously evaluated will be allowed. Therefore, this change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. The 
proposed change will not impose or eliminate any requirements, and 
adequate control of the information will be maintained. Thus, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change will not reduce a margin of safety because 
it has no impact on any safety analysis assumptions. In addition, 
the details to be transposed from the Technical Specifications to 
the TRM, or the NUQAP documents are the same as the existing 
Technical Specifications. Since any future changes to these details 
in the TRM or NUQAP will be evaluated per the requirements of 10 CFR 
50.59 or 10 CFR 50.54(a) respectively, no reduction (significant or 
insignificant) in a margin of safety will be allowed.

Relocated Specifications (``R.x'' Labeled Comments/Discussions

    In accordance with the criteria set forth in 10 CFR 50.92, NNECO 
has evaluated this proposed Technical Specifications change and 
determined it does not represent a significant hazards 
consideration. The following is provided in support of this 
conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change relocates requirements and surveillances for 
structures, systems, or components (SSCs) that do not meet the 
criteria for inclusion in Technical Specifications as defined in 10 
CFR 50.36. The affected SSCs are not assumed to be initiators of 
analyzed events and are not assumed to mitigate accident or 
transient events. The requirements and surveillances for these 
affected SSCs will be relocated from the Technical Specifications to 
an appropriate administratively controlled document which will be 
maintained pursuant to 10 CFR 50.59. In addition, the affected SSCs 
are addressed in existing surveillance procedures which are also 
controlled by 10 CFR 50.59 and subject to the change control 
provisions imposed by plant administrative procedures, which endorse 
applicable regulations and standards. Therefore, this change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant activities. The 
proposed change will not impose or eliminate any requirements and 
adequate control of existing requirements will be maintained. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change will not reduce a margin of safety because 
it has no impact on any safety analysis assumptions. In addition, 
the relocated requirements and surveillances for the affected SSCs 
remain the same as the existing Technical Specifications. Since any 
future changes to these requirements or the surveillance procedures 
will be evaluated per the requirements of 10 CFR 50.59, no reduction 
in a margin of safety will be permitted.

Specific Less Restrictive Changes (L.1 Labeled Comments/Discussions)

    In accordance with the criteria set forth in 10 CFR 50.92, 
Northeast Nuclear Energy Company (NNECO) has evaluated this proposed 
Technical Specifications change and determined it does not represent 
a significant hazards consideration. This change modifies the 
Applicability of LCO 3.1.1 from ``Whenever irradiated fuel is stored 
in the Fuel Storage Pool'' to ``During movement of irradiated fuel 
assemblies in the Fuel Storage Pool.'' This is consistent with the 
conditions addressed and assumed in the analysis of a fuel handling 
accident. Required Action A.2 is also deleted since, with the 
corresponding change to the Applicability, it is no longer required. 
The following is provided in support of this conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change involves modifying the Applicability of LCO 
3.1.1 to correspond directly with the conditions to which the LCO 
applies. LCO 3.1.1 provides assurance that adequate pool water level 
is maintained to ensure that the assumptions of the design basis 
fuel handling accident are met. The design basis accident assumes a 
non-mechanistic failure of the fuel pins in four assemblies. The 
analysis assumes that a water level below that required by LCO 
3.1.1.

[[Page 2017]]

If fuel handling is not occurring, the fuel pool water level does 
not satisfy the criteria for inclusion in the Technical 
Specifications as a parameter assumed as an initial condition of the 
safety analysis. Therefore this change merely aligns the LCO 
Applicability with the safety analysis assumptions.
    Aligning the Applicability directly with the conditions that 
must exist for a design basis accident to occur does not affect the 
probability or consequences of an accident previously evaluated. 
Rather, it ensures that the previously evaluated accident 
probability and consequences are unchanged. Therefore, this change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    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 activities. The 
proposed change will merely align the Applicability of an existing 
LCO with the conditions that exist when the limit of the LCO is 
credited in the safety analysis. Thus, this change 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 change will not reduce a margin of safety because 
the change merely aligns the Applicability of LCO 3.1.1 with the 
conditions that exist when the limit of the LCO is credited in the 
safety analysis. Therefore, the change does not involve a 
significant reduction in a margin of safety.

Specific Less Restrictive Changes (L.2 Labeled Comments/Discussions)

    In accordance with the criteria set forth in 10 CFR 50.92, 
Northeast Nuclear Energy Company (NNECO) has evaluated this proposed 
Technical Specifications change and determined it does not represent 
a significant hazards consideration. The proposed change removes a 
restriction from Section 4.1, Site Location, which restricts the 
sale or lease of portions of the site other than to the listed 
organizations. The following is provided in support of this 
conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change involves removing an administrative 
restriction on the ownership and ability to lease portions of the 
site to organizations other than those listed. Removing this 
restriction will not affect the probability of an accident 
previously evaluated, since these restrictions are not related to 
any precursor or contributor to the causes for any accident 
previously evaluated. Removing the restrictions will similarly not 
increase the consequences of an accident previously evaluated, since 
the proposed change does not result in a transfer of ownership or 
grant of lease of the described property. Any such activity would be 
subjected to a review in accordance with the requirements of 10 CFR 
50.59, since the ownership and physical description of the plant are 
described in the Defueled Safety Analysis Report. The evaluation 
performed at that time would ensure that no increase in the 
consequences of an accident previously evaluated. Therefore, this 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant activities. The 
proposed change merely removes an administrative requirement that 
limits the ability to sell or lease portions of the site. These 
controls are not associated with any onsite activity that could 
result in a new or different kind of accident. Thus, this change 
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 change will not reduce a margin of safety, because 
it does not result in any change to the plant or the way it is 
operated. The proposed change merely removes an administrative 
restriction on the ability to lease or sell portions of the site. 
Since the site description is provided in the Defueled Safety 
Analysis Report, any such activity would be subject to a review in 
accordance with the requirements of 10 CFR 50.59. This review would 
ensure that there is no reduction in margin of safety associated 
with any future proposed changes. Therefore, this change does not 
involve a significant reduction in a margin of safety.

Specific Less Restrictive Changes (L.3 Labeled Comments/Discussions)

    In accordance with the criteria set forth in 10 CFR 50.92, 
Northeast Nuclear Energy Company (NNECO) has evaluated this proposed 
Technical Specifications change and determined it does not represent 
a significant hazards consideration. The proposed change removes a 
limit associated with the storage of fuel in the new fuel storage 
facility. With the permanent shutdown and defueled condition of the 
plant, and the removal of all un-irradiated fuel from the site, the 
new fuel storage facility will no longer be used and this 
restriction is no longer required. The following is provided in 
support of this conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change involves removing restrictions on 
keff in the new fuel storage facility. Fuel can no longer 
be stored in the new fuel storage facility because all un-irradiated 
fuel has been removed from the site, and radiological considerations 
prevent the placement of irradiated fuel in the new fuel storage 
facility. The design basis accident for Millstone, Unit No. 1 is the 
postulated Fuel Handling Accident described in the Defueled Safety 
Analysis Report. The postulated accident involves irradiated fuel 
located in the spent fuel storage pool. Therefore, this requirement 
provides no useful information and does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    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 activities. The 
proposed change will not impose any new requirements. Thus, this 
change 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 change will not reduce a margin of safety, because 
the requirements that are proposed for elimination do not affect the 
design or operation of the facility since the plant was permanently 
shutdown, defueled, and all un-irradiated fuel has been removed from 
the unit. Since the proposed change has no affect on the facility 
and merely removes unnecessary information from the Technical 
Specifications, the change does not involve a significant reduction 
in a margin of safety.

Specific Less Restrictive Changes (L.4 Labeled Comments/Discussions)

    In accordance with the criteria set forth in 10 CFR 50.92, 
Northeast Nuclear Energy Company (NNECO) has evaluated this proposed 
Technical Specifications change and determined it does not represent 
a significant hazards consideration. The proposed change involves 
removing the requirement for a Shift Manager who is qualified as a 
Certified Fuel Handler and is responsible for the control room 
command function. The following is provided in support of this 
conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change involves removing the requirement for a 
Shift Manager who is qualified as a Certified Fuel Handler and who 
is responsible for the control room command function. Millstone, 
Unit No. 1 has been shutdown for over four years, and there are no 
remaining postulated or credible accidents that require a complex 
immediate response from operating personnel. The required response 
to postulated and credible accidents at the facility are a small 
subset of those that were required when the facility was in 
operation. Based on this, there is no longer a need for a specific 
position designation for the individual who will exercise the 
control room command function.
    In addition, the requirement for a Certified Fuel Handler to 
fulfill the Shift Manager responsibility is no longer appropriate 
because for extended periods no fuel handling operations will be 
conducted. Fuel Handling activities are deliberate pre-planned 
evolutions. There are no postulated or credible accidents that would 
result in the need to perform an unplanned fuel movement. Plant 
procedures and other administrative controls will continue to ensure 
that Certified Fuel Handler responsibilities are fulfilled by 
appropriately

[[Page 2018]]

qualified individuals when activities dictate the need.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    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 activities because 
qualified individuals will continue to be available to perform 
required functions. The proposed change will not impose any new or 
eliminate any old requirements associated with any structure, system 
or component. Thus, this change 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 change will not reduce a margin of safety, because 
qualified individuals will continue to be available to perform 
activities required to ensure the safe storage of irradiated fuel 
and control of radioactive materials. The proposed changes will 
eliminate unnecessarily burdensome requirements that were developed 
to address the requirements of an operating facility but which no 
longer apply at a permanently shutdown and defueled facility such as 
Millstone, Unit No. 1. Therefore, the change does not involve a 
significant reduction in a margin of safety.

Specific Less Restrictive Changes (L.5 Labeled Comments/Discussions)

    DOC L.5 is not used.

Specific Less Restrictive Changes (L.6 Labeled Comments/Discussions)

    In accordance with the criteria set forth in 10 CFR 50.92, 
Northeast Nuclear Energy Company (NNECO) has evaluated this proposed 
Technical Specifications change and determined it does not represent 
a significant hazards consideration. The proposed change removes an 
administrative requirement for notification to be made to the NRC 
prior to changes to acceptance criteria for chemistry control of the 
Fuel Storage Pool. The following is provided in support of this 
conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Removing the requirement for prior notification of the NRC 
cannot have any effect on the probability or consequence of an 
accident previously evaluated, since the requirement to perform this 
notification is not associated or related in any way to the 
probability or consequences of any accident.
    The consequence of an accident previously evaluated are not 
affected since no change to the way the fuel storage pool is 
monitored, is proposed. Notification of the NRC does not affect the 
consequences of any previously evaluated accident. The proposed 
change merely reduces the administrative burden associated with 
maintaining the program in compliance with the Technical 
Specifications.
    Therefore, these changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do 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 activities. The proposed 
changes will not impose any new or eliminate any old requirements. 
Thus, these changes do 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 will not reduce a margin of safety because 
they merely remove administrative burden associated with 
implementing the Fuel Storage Pool Program by eliminating a 
requirement for notification to the NRC of proposed changes to 
acceptance criteria to be used. Therefore, the change does not 
involve a significant reduction in a margin of safety.

Specific Less Restrictive Changes (L.7 Labeled Comments/Discussions)

    In accordance with the criteria set forth in 10 CFR 50.92, 
Northeast Nuclear Energy Company (NNECO) has evaluated this proposed 
Technical Specifications change and determined it does not represent 
a significant hazards consideration. The proposed change merely adds 
the option to use electronic dosimetry. The following is provided in 
support of this conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change involves adding the explicit option to 
utilize electronic dosimetry as a means of monitoring occupational 
radiation exposure. The means of monitoring occupational dose are 
unrelated to the probability or consequences of any accident 
previously evaluated. The means of measuring occupational exposures 
is merely a limit on the technology that may be utilized to perform 
a measurement required by [F]ederal regulations. Therefore, this 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
plant systems, structures or components (no new or different type of 
equipment will be installed) or changes in methods governing normal 
plant activities. The proposed change will not impose any new or 
eliminate any old requirements related to the safe storage of 
irradiated nuclear fuel or the control of radioactive materials. 
Thus, this change 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 change will not reduce a margin of safety, because 
the means of measuring the occupational exposure of workers is 
unrelated to the margin of safety of the facility. The means of 
measuring occupational exposures is merely a limit on the technology 
that may be utilized to perform a measurement required by [F]ederal 
regulations.

Specific Less Restrictive Changes (L.8 Labeled Comments/Discussions)

    In accordance with the criteria set forth in 10 CFR 50.92, 
Northeast Nuclear Energy Company (NNECO) has evaluated this proposed 
Technical Specifications change and determined it does not represent 
a significant hazards consideration. This change will extend the 
surveillance Frequency from once every 24 hours to once every 
[seven] days. The proposed Frequency is consistent with the reduced 
decay heat load and the lack of available mechanistic failures that 
could lead to sudden or unanticipated reduction in spent fuel pool 
inventory. The associated Bases are modified to reflect the proposed 
interval. The following is provided in support of this conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change involves extending the Frequency interval of 
SR [Surveillance Requirement] 3.1.1 to correspond with the 
conditions of the facility. SR 3.1.1 provides assurance that 
adequate pool water level is maintained to ensure that the 
assumptions of the design basis fuel handling accident are met. 
There are no longer any credible mechanisms that could lead to an 
unanticipated or undetected reduction in spent fuel pool inventory. 
The proposed [seven] day Frequency is consistent with the decay heat 
load calculations, potential maximum evaporation rates, and the 
large volume of water available over the spent fuel in the storage 
pool.
    Aligning this SR directly with the conditions that exist in the 
facility does not affect the probability or consequences of an 
accident previously evaluated. Rather, it continues to ensure that 
the previously evaluated accident probability and consequences are 
unchanged. Therefore, this change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    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 activities. The 
proposed change will merely align the Frequency of an existing SR 
with the conditions in the facility. Thus, this change 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 change will not reduce a margin of safety because 
the change merely aligns the Frequency of performance of SR 3.1.1 
with the conditions that exist in the plant. Therefore, the change 
does not involve a significant reduction in a margin of safety.

[[Page 2019]]

Specific Less Restrictive Changes (L.9 Labeled Comments/Discussions)

    In accordance with the criteria set forth in 10 CFR 50.92, 
Northeast Nuclear Energy Company (NNECO) has evaluated this proposed 
Technical Specifications change and determined it does not represent 
a significant hazards consideration. This change modifies the spent 
fuel storage rack limit on Keff from less than or equal 
to 0.90 to less than or equal to 0.95. The following is provided in 
support of this conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change involves modifying the keff limit 
that the spent fuel storage racks are designed and maintained to. 
The current and proposed limit are established to provide a 
significant margin of assurance that the spent fuel cannot be made 
critical while stored in the racks and under design basis accident 
conditions.
    Changing the limit on keff from 0.90 to 0.95 does not 
significantly affect the assurance that the spent fuel racks will 
maintain the fuel in a sub-critical configuration. Both limits are 
substantially below the limit of 1.0, and provide adequate assurance 
of safety. The proposed change therefore does not affect the 
probability or consequences of an accident previously evaluated. 
Rather, it continues to ensure that the previously evaluated 
accident probability and consequences are unchanged. Therefore, this 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    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 activities. The 
proposed change will merely increase the limit on keff so 
that it is consistent with industry practice and established 
standards applicable to the storage of spent fuel. Criticality 
continues to be avoided by maintaining the storage racks such that 
keff is less than or equal to 0.95. Thus, this change 
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 margin of safety defined by the limit is that the spent fuel 
will remain sub-critical during anticipated circumstances and design 
basis accidents. Since the proposed limit continues to provide this 
assurance, the change does not involve a significant reduction in a 
margin of safety.

Specific Less Restrictive Changes (L.10 Labeled Comments/Discussions)

    In accordance with the criteria set forth in 10 CFR 50.92, 
Northeast Nuclear Energy Company (NNECO) has evaluated this proposed 
Technical Specifications change and determined it does not represent 
a significant hazards consideration. This change removes the 
redundant requirement to maintain an NRC approved training and 
retraining program for the Certified Fuel Handlers (CFHs). The 
following is provided in support of this conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change removes a TS [Technical Specification] 
administrative requirement that is redundant to existing 
requirements that derive from 10 CFR 50.2. Therefore the TS 
requirement is not needed and does not [a]ffect the probability or 
consequences of an accident previously evaluated. The change is 
purely administrative, albeit a specific reduction in the 
requirements of the TS. The requirement will continue to apply to 
the unit. Therefore, this change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    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 activities. The 
proposed change will merely remove an unneeded, redundant 
requirement. Therefore, this change 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 change will not reduce a margin of safety because 
the requirement for an NRC approved training program for CFHs will 
continue to exist as specified in 10 CFR 50.2. 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: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P. O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: Michael T. Masnik.

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

    Date of amendment request: August 3, 2000.
    Description of amendment request: The proposed amendment would 
delete Section 3.D, ``License Term'' from Facility Operating License 
No. DPR-40. The long-term load factor described in Section 3.D is used 
in the projection of reactor vessel fast neutron fluence and 
consequently for calculation of the pressurized thermal shock (PTS) 
reference temperature (RTPTS) value to ensure that the 10 
CFR 50.61 screening criteria for reactor vessel integrity are not 
exceeded. The previous fluence analysis was performed by Combustion 
Engineering (ABB/CE). Recently, Westinghouse Electric Company has 
completed an analysis (WCAP-15443, ``Fast Neutron Fluence Evaluations 
for the Fort Calhoun Unit 1 Reactor Pressure Vessel,'' dated July 2000) 
to update the ABB/CE calculation. In accordance with 10 CFR 50.61, this 
assessment must be updated whenever there is a significant change in 
projected values of RTPTS or upon request for a change in 
the expiration date of the facility. Thus, Section 3.D can be deleted 
from Facility Operating License No. DPR-40 based upon the recent 
Westinghouse analysis and the fact that Section 3.D is redundant to 10 
CFR 50.61 requirements.
    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 previously evaluated accidents affected by this change are 
limited to the pressurized thermal shock (PTS) events. Vessel 
embrittlement due to fast neutron associated damage to the limiting 
beltline region reactor vessel material (which for Fort Calhoun 
Station is included in the lower course axial welds) is a component 
in the PTS analysis. The fast neutron, thermal neutron and dpa 
[displacement per atom] values of the FCS reactor vessel were 
recalculated using actual power history values for Cycles 1 through 
14 rather than conservative estimates, along with the revised BUGLE-
93 cross sections from the ENDF/B-VI cross section library to 
appropriately account for the iron atoms in the thermal shield and a 
methodology that the NRC has previously approved for neutron fluence 
calculations performed by Westinghouse. The fluence evaluation 
included data from the three surveillance capsules (W-225, W-265, 
and W-275) previously removed and analyzed. The RTPTS 
evaluation applied Position 2.1 of Regulatory Guide 1.99, Revision 2 
in conjunction with surveillance data from other plants containing 
the limiting FCS weld materials. The evaluation results indicate 
that the FCS reactor vessel is able to reach more than 20 years 
beyond current licensed life without exceeding the 10 CFR 50.61 
screening criterion for RTPTS of 270 deg.F for axial 
welds.
    In accordance with 10 CFR 50.61, this assessment must be updated 
whenever there is a significant change in projected values of 
RTPTS or upon request for a change in the expiration date 
of the facility. Since these requirements are contained in 10 CFR 
50.61, Section 3.D can be deleted from Operating License No. DPR-40 
without resulting in a

[[Page 2020]]

significant increase in the probability or consequences of any 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not physically alter the configuration 
of the plant and no new or different mode of operation is proposed. 
Increasing the long term load factor from 0.77 to 0.85 more 
accurately projects RTPTS by accounting for improvement 
in FCS operating cycle efficiency. Requirements for assessing and 
reporting RTPTS are contained in 10 CFR 50.61 and 
therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously analyzed.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety is defined by both the screening criteria 
of 10 CFR 50.61 and draft regulatory guide DG-1053 for neutron 
fluence calculations, which requires the methodology to be capable 
of providing best estimate fluence evaluations within 20 percent 
(1). The analysis for FCS shows that when the applicable 
regulatory criteria are applied, the screening criteria of 10 CFR 
50.61 are not exceeded; therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

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

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

    Date of amendment request: December 14, 2000 (This supercedes the 
license amendment request dated July 28, 2000, that was published in 
the Federal Register on October 18, 2000 (65 FR 62388).
    Description of amendment request: The proposed amendment would 
permit Fort Calhoun Station to install leak tight sleeves as an 
alternative to plugging to repair defective steam generator tubes.
    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 CE Leak Tight Sleeves are designed using the applicable 
American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code and, therefore, meet the design objectives of the 
original steam generator tubing. The applicable design criteria for 
the sleeves conform to the stress limits and margins of safety of 
Section III of the ASME code. Mechanical testing has shown that the 
structural strength of repair sleeves under normal, upset, and 
faulted conditions provides margin to the acceptance limits. These 
acceptance limits bound the most limiting (three times normal 
operating pressure differential) burst margin recommended by 
Regulatory Guide 1.121. Burst testing of sleeved tubes has 
demonstrated that no unacceptable levels of primary-to-secondary 
leakage are expected during any plant condition.
    Evaluation of the repaired steam generator tubes indicates no 
detrimental effects on the sleeve or sleeve-tube assembly from 
reactor coolant system flow, primary or secondary coolant 
chemistries, thermal conditions or transients, or pressure 
conditions as may be experienced at Fort Calhoun Station. Corrosion 
testing of sleeve-tube assemblies indicates no evidence of sleeve or 
tube corrosion considered detrimental under anticipated service 
conditions.
    The installation of the proposed sleeves is controlled via the 
sleeving vendor's proprietary processes and equipment. The CE 
process has been in use since 1984 and has been implemented more 
than 24 times for the installation of over 4,200 sleeves. The FCS 
steam generator design was reviewed and found to be compatible with 
the installation processes and equipment.
    The implementation of the proposed amendment has no significant 
effect on either the configuration of the plant or the manner in 
which it is operated. The consequences of a hypothetical failure of 
the sleeved tube is bounded by the current steam generator tube 
rupture analysis described in Fort Calhoun Station's USAR [Updated 
Safety Analysis Report], Section 14.14. Due to the slight reduction 
in diameter caused by the sleeve wall thickness, primary coolant 
release rates would be slightly less than assumed for the steam 
generator tube rupture analysis, depending on the break location, 
and therefore, would result in lower total primary fluid mass 
release to the secondary system. A main steam line break or feed 
line break will not cause a SGTR [steam generator tube rupture] 
since the sleeves are analyzed for a maximum accident differential 
pressure greater than that predicted in the Fort Calhoun Station 
safety analysis. The proposed reduction of the steam generator 
primary to secondary operational leakage limit provides added 
assurance that leaking flaws will not propagate to burst prior to 
commencement of plant shutdown.
    In conclusion, based on the discussion above, these changes will 
not significantly increase the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    As discussed above, the CE Leak Tight Sleeves are designed using 
the applicable ASME Code as guidance; therefore, they meet the 
objectives of the original steam generator tubing. As a result, the 
functions of the steam generators will not be significantly affected 
by the installation of the proposed sleeves. The proposed repair 
sleeves do not interact with any other plant systems. Any accident 
as a result of potential tube or sleeve degradation in the repaired 
portion of the tube is bounded by the existing tube rupture accident 
analysis. The continued integrity of the installed sleeve is 
periodically verified by the Technical Specification requirements.
    The implementation of the proposed amendment has no significant 
effect on either the configuration of the plant or the manner in 
which it is operated. As discussed above, the reduced primary to 
secondary leakage limit is a conservative change in the plant 
limiting conditions for operation. Therefore, Omaha Public Power 
District concludes that this proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The repair of degraded steam generator tubes with CE Leak Tight 
Sleeves restores the integrity of the degraded tube under normal 
operating and postulated accident conditions. The design safety 
factors utilized for the repair sleeves are consistent with the 
safety factors in the ASME Code used in the original steam generator 
design. The portions of the installed sleeve assembly that 
represents the reactor coolant pressure boundary can be monitored 
for the initiation and progression of sleeve/tube wall degradation. 
Use of the previously identified design criteria and design 
verification testing assures that the margin of safety is not 
significantly different from the original steam generator tubes. The 
proposed sleeve inspection requirements are more stringent than 
existing requirements for inspection of the steam generator tubes, 
and the reduction in the operational limit for primary to secondary 
leakage through the steam generator tubes is more conservative than 
current requirements. Therefore, OPPD concludes that the proposed 
change does not involve a significant reduction in a margin of 
safety.

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

[[Page 2021]]

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

    Date of amendment request: September 5, 2000.
    Description of amendment request: The proposed change revises 
Surveillance Requirement 4.6.3.4 to require testing of a representative 
sample of Excess Flow Check Valves (EFCVs) such that each EFCV will be 
tested at least once every 120 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. 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 TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The current Surveillance Requirement (SR) frequency requires 
each reactor instrumentation line Excess Flow Check Valve (EFCV) to 
be tested every 24 months. The EFCVs at LGS (Limerick Generating 
Station), Units 1 and 2 are designed to not close accidentally 
during normal operation, but will close automatically in the event 
of a line break downstream of the valve. The EFCVs are provided with 
valve position indication in the reactor enclosure. A general alarm 
is provided in the control room to indicate that an EFCV position 
has changed state. As discussed in the LGS, Units 1 and 2 Updated 
Final Safety Analysis Report (UFSAR) (Section 6.2.4.3.1.5), 
instrument lines that penetrate the containment from the Reactor 
Coolant Pressure Boundary (RCPB) conform to Regulatory Guide 1.11 in 
that they are equipped with a restricting orifice located inside the 
drywell and an EFCV located outside the drywell as close as 
practical to the containment. The GE Nuclear Energy (GENE) Report 
demonstrates, through operating experience, a high degree of 
reliability with the EFCVs and the low consequences of an EFCV 
failure. A failure of an EFCV to isolate cannot initiate previously 
evaluated accidents. In addition, since the proposed changes will 
only change the surveillance frequency, there can be no increase in 
the probability of occurrence of an accident as a result of this 
proposed change.
    The postulated break of an instrument line attached to the RCPB 
is discussed and evaluated in the Updated Final Safety Analysis 
Report (UFSAR), Section 15.6.2. The integrity and functional 
performance of the secondary containment and standby gas treatment 
system are not impaired by this event, and the calculated potential 
offsite exposures are substantially below the guidelines of 10 CFR 
100. Therefore, a failure of an EFCV, though not expected as a 
result of the change in the surveillance frequency, is bounded by 
the previous evaluation of an instrument line break. The radiation 
dose consequences of such a break are not impacted by this proposed 
change. Therefore, the proposed TS changes do not involve a 
significant increase in the 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 proposed changes allow a reduced number of EFCVs to be 
tested each operating cycle. No other changes in requirements are 
being proposed. Industry operating experience as documented in the 
GENE report provides supporting evidence that the reduced testing 
frequency does not affect the kind of accident. The potential 
failure of an EFCV to isolate as a result of the proposed reduction 
in test frequency is not a physical alteration of the plant and will 
not alter the operation of the structures, systems and components as 
described in the UFSAR. Therefore, a new or different kind of 
accident will not be created.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The consequences of an unisolable rupture of an instrument line 
has been previously evaluated in the LGS, Units 1 and 2 UFSAR, 
Section 15.6.2. That evaluation assumed a continuous discharge of 
reactor water for the duration of the detection and cooldown 
sequence. The change in surveillance frequency only changes the 
potential for an undetected failure of an EFCV and does not change 
the event sequence upon which the current margin is based. 
Therefore, no change in the 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: 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.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: October 4, 2000.
    Description of amendment request: The amendment proposes changes to 
the Susquehanna Steam Electric Station (SSES), Units 1 and 2, Technical 
Specifications (TSs) to revise the surveillance requirement (SR) for 
certain isolation valves known as excess flow check valves (EFCV). The 
current TSs require that each EFCV be tested at least once every 24 
months. The proposed change would allow a representative sample to be 
tested every 24 months, such that each EFCV is tested at least once 
every 10 years.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This proposed changes do not involve an increase in the 
probability or consequences of an accident previously evaluated.
    The current SR frequency requires each reactor instrumentation 
line EFCV to be tested every 24 months. The EFCVs at SSES Unit 1 and 
Unit 2 are designed so that they will not close accidentally during 
normal operation, will close if a rupture of the instrument line is 
indicated downstream of the valve, can be reopened when appropriate, 
and have their status indicated in the control room. This proposed 
change allows a reduced number of EFCVs to be tested every 24 
months. There are no physical plant modifications associated with 
this change. Industry and SSES operating experience demonstrates a 
high reliability of these valves. Neither EFCVs nor their failures 
are capable of initiating previously evaluated accidents; therefore 
there can be no increase in the probability of occurrence of an 
accident regarding this proposed change.
    The SSES FSAR [Final Safety Analysis Report] Section 15.6.2 
demonstrates (consistent with the BWROG [Boiling Water Reactor 
Owners' Group] report) that the failure of an EFCV has very low 
consequence. SSES FSAR Section 15.6.2 evaluates a circumferential 
rupture of an instrument line that is connected to the primary 
coolant system. The evaluation assumes the EFCV fails to isolate the 
break. The dose consequences of the instrument line break are 
determined using the calculated mass of coolant released over 
approximately a 5 hour period. The reactor was assumed to be at full 
power prior to the break. The Standby Gas Treatment System (SGTS) 
and secondary contaimnent are not impaired by the event. The 
evaluation concludes that the

[[Page 2022]]

consequences of the event are well within 10 CFR 100 limits. Thus 
the failure of an EFCV, though not expected as a result of this 
proposed change, does not affect the dose consequences of an 
instrument line break.
    Based on the above, it is concluded that the proposed change to 
the EFCV surveillance requirement does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
This proposed change allows a reduced number of EFCVs to be tested 
each operating cycle. No other changes in requirements are being 
proposed. Industry and Susquehanna-specific operating experience 
demonstrates the high reliability of these valves. The potential 
failure of an EFCV to isolate by the proposed reduction in test 
frequency is bounded by the previous evaluation of an instrument 
line rupture. This change will not physically alter the plant (no 
new or different type of equipment will be installed). This change 
will not alter the operation of process variables, structures, 
systems, or components as described in the safety analysis. Thus, a 
new or different kind of accident will not be created from 
implementation of the proposed change.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    SSES FSAR Section 15.6.2 evaluates a circumferential rupture of 
an instrument line that is connected to the primary coolant system. 
The evaluation assumes the EFCV fails to isolate the break. The dose 
consequences of the instrument line break are determined using the 
calculated mass of coolant released over approximately a 5 hour 
period. The reactor was assumed to be at full power prior to the 
break. The Standby Gas Treatment System (SGTS) and secondary 
containment are not impaired by the event. The evaluation concludes 
that the consequences of the event are well within 10CFR100 limits. 
Thus the failure of an EFCV, though not expected as a result of this 
proposed change, does not affect the dose consequences of an 
instrument line break.
    Therefore, this proposed change does not represent a significant 
reduction in the margin of safety.

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

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: November 16, 2000.
    Description of amendment request: The request for amendment 
proposes changes to the Susquehanna Steam Electric Station (SSES), 
Units 1 and 2, Technical Specifications to eliminate response time 
testing requirements for certain reactor protection system and 
isolation actuation system instrumentation.
    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 proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. The 
proposed change eliminates certain response time testing [RTT] 
surveillance requirements in accordance with the NRC [Nuclear 
Regulatory Commission] approved methodology delineated in the BWROG 
[Boiling Water Reactor Owners' Group] Licensing Topical Report [LTR] 
NEDO 32291, ``System Analyses for Elimination of Selected Response 
Time Testing Requirements,'' dated October 1995, and its Supplement 
1, dated October, 1999.
    Implementation of the LTR and its supplement (i.e., elimination 
of response time testing for selected instrumentation in the Reactor 
Protection System [RPS] and Isolation Actuation System [IAS]) does 
not increase the probability or consequences of an accident or 
malfunction of equipment important to safety as previously evaluated 
in the FSAR [Final Safety Analysis Report]. All component models 
used in the affected trip channels at SSES were analyzed for a 
sluggish response, or a bounding response time. As documented in the 
LTR and supplement, the component's sluggish response can be 
detected by other Technical Specification required tests. The 
bounding response time of the relays discussed in the LTR Supplement 
1 can be used in place of actual measured response times to ensure 
that instrumentation systems will meet response time requirements of 
the accident analysis. Response Time Testing for the channel process 
sensors are also eliminated on a similar basis, or have previously 
been eliminated in license amendments (171 (Unit 1) and 144 (Unit 
2)).
    Based upon the analysis presented above, PPL concludes that the 
proposed action does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposal does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed change eliminates certain response time testing (RTT) 
surveillance requirements in accordance with the NRC approved 
methodology delineated in the BWROG Licensing Topical Report (LTR) 
NEDO 32291, ``System Analyses for Elimination of Selected Response 
Time Testing Requirements,'' dated October 1995, and its Supplement 
1, dated October, 1999.
    Implementation of the LTR methodology and the Supplement 
methodology does not create the probability of a new or different 
type of accident from any accident previously evaluated. A review of 
the failure modes of the affected sensors and relays indicates that 
a sluggish response of the instruments can be detected by other 
Technical Specification surveillances. A review of SSES RTT history 
(in support of the LTR) revealed one RTT failure. This failure would 
have been detectable by the logic system functional test for this 
channel. Redundancy and diversity of the affected channels provide 
additional assurance that all affected functions will operate within 
the acceptance limits of the safety evaluations.
    The sensors and relays in the affected RPS and IAS channels will 
be able to meet the bounding response times as defined and presented 
in the Supplement. It has been found acceptable to use component 
bounding response times in place of actual measured response times 
to ensure that instrumentation systems will meet response time 
requirements of the accident analysis.
    PPL's adherence to the conditions listed in the NRC SERs [Safety 
Evaluation Reports] for the LTR and Supplement provides additional 
assurance that the instrumentation systems will meet the response 
time requirements of the accident analyses.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The change does not involve a significant reduction in the 
margin of safety. The proposed change eliminates certain response 
time testing (RTT) surveillance requirements in accordance with the 
NRC approved methodology delineated in the BWROG Licensing Topical 
Report (LTR) NEDO 32291, ``System Analyses for Elimination of 
Selected Response Time Testing Requirements,'' dated October 1995, 
and its Supplement 1, dated October, 1999.
    Implementation of the LTR and Supplement methodologies for 
eliminating selected response time testing does not involve a 
significant reduction in the margin of safety. The current response 
time limits are based on the maximum allowable values assumed in the 
plant safety analyses. The analyses conservatively establish the 
margin

[[Page 2023]]

of safety. The elimination of the selected response time testing 
does not affect the capability of the associated systems to perform 
their intended function within the allowed response time used as 
tile basis for plant safety analyses. Plant and system response to 
an initiating event will remain in compliance within the assumptions 
of the safety analyses, and therefore, the margin of safety is not 
affected. This is based upon the ability to detect a sluggish 
response of an instrument or relay by the other required Technical 
Specification tests, component reliability, and redundancy and 
diversity of the affected functions, as justified in the reviewed 
and approved Topical Report and Supplement.
    PPL's adherence to the conditions listed in the NRC SERs for the 
LTR and Supplement provides additional assurance that the 
instrumentation systems will meet the response time requirements of 
the accident analyses.
    Thus, PPL concludes that the proposed change does not involve a 
significant reduction in the margin of safety.

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

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: November 28, 2000.
    Description of amendment request: The proposed change would modify 
Technical Specification (TS) Surveillance Requirement (SR) 3.6.1.6.3 to 
expand the allowable vacuum breaker open differential pressure setpoint 
range to  .25 pounds-per-square-inch differential (psid) and 
 .75 psid. The SR in the current TSs requires testing to a 
range of  .25 psid and  .525 psid.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This proposed change does not involve an increase in the 
probability or consequences of an accident previously evaluated.
    The SR is required to verify that the vacuum breakers open when 
required by the containment safety analysis. The vacuum breakers 
setpoint prevent the creation of a vacuum in the drywell or an 
unacceptable differential pressure across the containment diaphragm 
slab. When the drywell pressure falls below the airspace pressure by 
an amount equal to the open set pressure of the vacuum breakers, the 
vacuum breakers open to allow the suppression chamber atmosphere 
from the wetwell airspace to flow into the drywell.
    The ability to maintain containment integrity is not affected by 
the proposed change. Containment analyses are not affected by the 
proposed change.
    Containment analyses assume the vacuum breakers open at .9 psid. 
Thus, the vacuum breakers at the new setpoint range are bounded by 
the setpoint assumed in the analysis. Sensitivity analyses show that 
the containment pressure response is insignificantly affected by the 
proposed change.
    The setpoint expansion does not adversely affect the vacuum 
breakers ability to perform their design basis functions.
    Based on the above, it is concluded that the proposed change to 
the vacuum breakers setpoint surveillance requirement does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
This proposed change allows an expanded setpoint range. No other 
changes in requirements are being proposed. This change will not 
physically alter the plant (no new or different type of equipment 
will be installed). This change will not alter the operation of 
process variables, structures, systems, or components as described 
in the safety analysis. The new range is bounded by the safety 
analysis assumptions. Thus, a new or different kind of accident will 
not be created from implementation of the proposed change.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    This proposal does not involve a significant reduction in the 
margin of safety.
    The containment pressures are insignificantly affected by the 
proposed change. Safety analyses assume a bounding setpoint. The 
operation of the vacuum breakers are not affected.
    Therefore, this proposed change does not represent a significant 
reduction in the margin of safety.

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

Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364, 
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

    Date of amendment request: August 17, 2000.
    Description of amendment request: The proposed amendments would 
eliminate the need for the licensee to perform periodic response time 
testing of selected reactor trip system and engineered safety feature 
actuation system equipment as defined in Westinghouse report WCAP-
14036-P-A Revision 1, ``Elimination of Periodic Protection Channel 
Response Time Tests.''
    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:

    Conformance of the proposed amendment to the standards for a 
determination of no significant hazard as defined in 10 CFR 50.92 is 
shown in the following.
    1. The proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This change to the Technical Specifications does not result in a 
condition where the design, material, and constructionstandards that 
were applicable prior to the change are altered. The same RTS 
[Reactor Trip System] and ESFAS [Engineered Safety Features 
Actuation] instrumentation is being used. The time response 
allocations and modeling assumptions used in the Chapter 6 and 
Chapter 15 safety analyses of the Final Safety Analysis Report 
(FSAR) are not changed; only the method of verifying response time 
is changed. The proposed change will not modify any system interface 
or equipment design specification. The proposed change can not 
increase the likelihood of an accident since such postulated events 
are independent of this change. The proposed activity will not 
change, degrade or prevent actions or alter any assumptions 
previously made in evaluating the radiological consequences of an 
accident described in the FSAR. Therefore, the proposed amendment 
does not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    This change does not alter the performance of the protection 
channel and actuation logic

[[Page 2024]]

equipment used in the RTS and ESFAS. These protection systems will 
still have response time verified by test before being placed in 
operational service. Changing the method of periodically verifying 
instrument response for these systems (assuring equipment 
operability) from time response testing to calibration and 
functional testing will not create any new accident initiators or 
scenarios. Periodic surveillance of these systems will continue and 
may be used to detect degradation that could cause the response time 
characteristic to exceed the total allowance. The total time 
response allowance for each function and the response time allowance 
for individual components (e.g., circuit boards and relays) bound 
all degradation that cannot be detected by periodic surveillance. 
Therefore, implementation of the proposed amendment does 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.
    This change does not affect the total RTS and ESFAS response 
times assumed in the safety analyses. The periodic response time 
verification method for the 7300 Process Protection racks, NIS 
[Nuclear Instrumentation System] racks, and SSPS [Solid-State 
Protection System] actuation logic is modified to allow use of 
actual test data or engineering data. The method of verification 
still provides assurance that the total system response is within 
that defined in the safety analysis. Periodic calibrations and 
functional tests will continue to be performed and may be used to 
detect degradation which might cause the response time to exceed the 
total allowance. The time response allowance for each component and 
function bounds all degradation that cannot be detected by periodic 
surveillance. Based on the above, it is concluded that the proposed 
license amendment request does not result in a significant reduction 
in margin with respect to plant safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Section Chief: Richard L. Emch, Jr..

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

    Date of amendment request: December 6, 2000.
    Brief description of amendment request: The proposed license 
amendment request will revise Administrative Controls in Technical 
Specification (TS) 5.5.14, entitled ``Technical Specifications (TS) 
Bases Control Program'' and TS 5.5.17, entitled ``Technical 
Requirements Manual (TRM)'' to incorporate the changes made to 10 CFR 
50.59 as published in the October 4, 1999, Federal Register, Volume 64, 
Number 191, ``Changes, Tests, and Experiments,'' pages 53582 through 
53617.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes replace the word ``involve(s)'' with 
``require(s)'' and deletes reference to the term ``unreviewed safety 
question.'' The above changes are consistent with the revision to 10 
CFR 50.59. Consequently, the probability of an accident previously 
evaluated is not increased. Changes to the Technical Specification 
(TS) Bases and the Technical Requirements Manual (TRM) are still 
evaluated in accordance with 10 CFR 50.59. As a result, the 
consequences of any accident previously evaluated are not affected.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or a 
change in the methods governing plant operation. These changes are 
considered administrative changes and do not modify, add, delete, or 
relocate any technical requirements in the TS.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes will not reduce the margin of safety 
because they have no direct effect on any safety analyses 
assumptions. Changes to the TS Bases and the TRM that result in 
meeting the criteria in paragraph (c)(2) of 10 CFR 50.59 will still 
require NRC approval. The proposed changes to TS 5.5.14 and TS 
5.5.17 are considered administrative in nature based on the 
revisions to 10 CFR 50.59.
    Therefore the proposed change does 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: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: November 27, 2000.
    Description of amendment request: This proposed change eliminates 
the specifications associated with the 24 Vdc Emergency Core Cooling 
System (ECCS) instrumentation batteries and chargers. The 24 Vdc ECCS 
instrumentation loads will be transferred to the 125 Vdc main station 
batteries.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The operation of the Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The loads previously supplied by the ECCS battery systems will 
be added to the main station battery systems. Redundancy and 
reliability are maintained within the main station battery systems 
and the equipment will operate, essentially the same. No change in 
accident assumptions or pre[]cursors are involved with this change 
and system operation and response to analyzed events is likewise 
unchanged.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The methods by which the DC system supplied equipment performs 
their safety functions are unchanged and remain consistent with 
current safety analysis assumptions. The redundancy and reliability 
of the equipment will be maintained. There is no change in system or 
plant operation that involves failure modes other than those 
previously evaluated.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.

[[Page 2025]]

    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    No adverse effect on equipment operation, capability or 
reliability will result from this change. The equipment supplied by 
the DC systems involved in this change will continue to be provided 
with adequate, redundant, reliable, safety class DC power. Safety 
related loads will continue to function in accordance with analysis 
assumptions.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: November 30, 2000.
    Description of amendment request: The proposed change would revise 
the operability requirements for the refueling interlocks contained 
within Technical Specification (TS) 3.12.A as well as the surveillance 
requirements specified within TS 4.12.A. In addition, TS 3.12.F will be 
clarified to articulate that there must be a minimum of 24 hours 
fission product decay prior to fuel handling.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. The only accident described within 
the [Final Safety Analysis Report] FSAR while the plant is in Cold 
Shutdown or Refueling is a fuel handling (dropped bundle) accident. 
The proposed change involves equipment that is not involved in the 
mitigation or prevention of a fuel handling accident as described in 
the FSAR. Accordingly, the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change will not effect the ability of the refueling 
interlocks to satisfy the safety function which is to prevent 
reactor criticality during refueling operations. The change only 
effects those interlocks which are not instrumental in satisfying 
the safety function of the interlocks.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change does not involve any physical alteration of 
plant equipment or to the status of the reactor core during 
refueling. The specifications will ensure either through the 
interlocks or the proposed alternative, that control rods are not 
withdrawn and cannot be inappropriately withdrawn. This will ensure 
that fuel is not loaded into the core when a control rod is 
withdrawn.
    Therefore, no new failure modes are introduced and the proposed 
change will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    The proposed change does not involve a significant reduction in 
a margin of safety since the refueling interlocks will continue to 
ensure against an inadvertent criticality. This is achieved by 
physical interlocks or Technical Specification restrictions on 
refueling operations which will prevent fuel from being loaded into 
a core cell void of a control rod. This is accomplished by blocking 
control rod withdrawal whenever fuel is being loaded into the 
reactor vessel or by preventing fuel from being loaded into the 
vessel when a control rod is withdrawn.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: December 19, 2000. This supersedes the 
March 17, 2000, submittal which was noticed on May 3, 2000 (65 FR 
25769).
    Description of amendment request: The proposed changes would modify 
the voltage setting limits specified in Technical Specification (TS) 
Table 3.7-4, page 3.7-26, item 7 for the emergency bus degraded 
voltage, and revise the loss of voltage setpoints from a percentage of 
nominal bus voltage to an actual bus voltage value. The degraded 
voltage setting limit is being changed to increase the minimum 
allowable bus voltage to improve long-term motor performance in the 
event of operation with bus voltage less than nominal. The emergency 
bus loss of voltage setting limit is being revised to better address 
expected relay performance over time (i.e., setting drift). Section 
3.6.B, page 3.6-1, of the TS would be changed to revise the required 
reactor coolant system conditions from the existing wording of ``350 
degrees F or 450 psig'' to ``350 degrees F and 450 psig.''
    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:

    We have reviewed the proposed change against the criteria of 10 
CFR 50.92 and have concluded that the change does not pose a 
significant safety hazards consideration as defined therein. 
Specifically, operation of Surry Power Station with the proposed 
change will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    No increase in the probability of occurrence or consequences of 
an accident previously evaluated will result from the proposed 
change in the setting limits for the emergency bus degraded voltage 
and loss of voltage relay setpoints. The proposed change only 
affects actuation limits and therefore has no bearing on the 
probability of an accident. Neither the logic nor the function of 
the undervoltage protection circuits is being changed, nor is 
circuit or equipment reliability being reduced. Further, the 
performance characteristics of the electrical distribution system 
and components supplied (motors, etc.) are not being altered, and 
compliance with GDC-17 [General Design Criterion] is being 
maintained. The electrical distribution system remains capable of 
performing its safety function without spurious separation of the 
emergency buses from offsite power. If offsite power is lost, the 
capability of the EDG's [emergency diesel generators] to perform 
their safety function is not altered. Therefore, the probability of 
an accident previously evaluated is not increased.
    The consequences of an accident would not increase since the 
proposed change implements setting limits that will continue to 
ensure that adequate voltages will be available for the continuous 
operation of safety-related equipment required to function to 
mitigate a design basis accident. The proposed setting limits for 
the emergency bus degraded voltage and loss of voltage bound

[[Page 2026]]

the setpoints and initial conditions assumed in the accident 
analyses and ensure that appropriate protection is maintained.
    The editorial change is administrative in nature and 
consequently does not affect the probability or consequences of an 
accident in any way.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Implementing the proposed Technical Specifications emergency bus 
degraded voltage and loss of voltage relay setting limits cannot 
create the possibility of a new or different kind of accident than 
any accident previously evaluated. Revising the setpoint setting 
limits does not introduce any new accident precursors, and operation 
of the electrical distribution system and the undervoltage relaying 
scheme is unchanged. The relays will continue to detect undervoltage 
conditions and transfer safety loads to the emergency diesel 
generators at a voltage level adequate to ensure proper safety 
equipment performance and to prevent long-term equipment degradation 
due to undervoltage conditions. The proposed setting limits include 
adequate tolerances to calibrate the undervoltage relays while 
ensuring that emergency bus voltages remain above analytical limits. 
As noted above, the performance characteristics of the electrical 
distribution system and the components being supplied are not being 
altered, and compliance with GDC-17 is being maintained.
    The editorial change is administrative in nature and 
consequently does 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 change continues to ensure that adequate voltage is 
available for safety-related equipment relied upon to respond to a 
design basis accident. The proposed setting limit for degraded bus 
voltage is conservative with respect to the existing Technical 
Specifications and ensures an adequate safety margin is being 
maintained. Further, the setting limit is maintained low enough to 
prevent spurious actuations given expected offsite grid voltages. 
While the loss of bus voltage setting limit is being expanded, 
sustained bus voltage in this range is not credible. Furthermore, 
there is no safety limit associated with the loss of voltage setting 
limit. The proposed change continues to ensure that the setting 
limits for the emergency bus degraded voltage and loss of voltage 
relays bound the setpoints and initial conditions assumed in the 
accident analyses and ensures that appropriate electrical protection 
is maintained. The editorial change is administrative in nature and 
consequently does not affect the safety analysis in any way. 
Consequently, the margin of safety is not being reduced by the 
proposed Technical Specifications change.

    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: Donald P. Irwin, Esq., Hunton and Williams, 
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
23219.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: December 7, 2000 (ET 00-0044).
    Description of amendment request: The proposed amendment adds text 
to Section 5.5.2, ``Primary Coolant Sources Outside Containment,'' and 
deletes Section 5.5.3, ``Post Accident sampling,'' from the 
administrative controls section of the Technical Specifications (TS). 
The proposed amendment deletes requirements from the TS (and, as 
applicable, other elements of the licensing bases) to maintain a Post 
Accident Sampling System (PASS). Licensees were generally required to 
implement PASS upgrades as described in NUREG-0737, ``Clarification of 
TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory 
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Assess Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades was an outcome of the 
lessons learned from the accident that occurred at TMI Unit 2. 
Requirements related to PASS were imposed by Order for many facilities 
and were added to or included in the TS for nuclear power reactors 
currently licensed to operate. Lessons learned and improvements 
implemented over the last 20 years have shown that the information 
obtained from PASS can be readily obtained through other means or is of 
little use in the assessment and mitigation of accident conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in a license amendment application in the Federal Register 
on October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated December 7, 2000.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications

[[Page 2027]]

(TS) (and other elements of the licensing bases) does not involve a 
significant increase in the consequences of any accident previously 
evaluated.

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

    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radionuclides 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.
    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: December 8, 2000.
    Description of amendment request: The proposed amendment would 
revise Administrative Controls Technical Specifications (TS) 5.5.14b 
and 5.5.14b.2 to incorporate the changes made to 10 CFR 50.59. The 
proposed changes would replace the word ``involve'' with ``require'' in 
TS 5.5.14b and revise TS 5.5.14b.2 to state: ``a change to the USAR 
[Updated Safety Analysis Report] or Bases that requires NRC approval 
pursuant to 10 CFR 50.59.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes replace the word ``involve'' with 
``require'' and deletes reference to the term ``unreviewed safety 
question'' consistent with 10 CFR 50.59. The above changes are 
consistent with the revision to 10 CFR 50.59. Consequently, the 
probability of an accident previously evaluated is not increased. 
Changes to the Technical Specification (TS) Bases are still 
evaluated in accordance with 10 CFR 50.59. As a result, the 
consequences of any accident previously evaluated are not affected.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or a 
change in the methods governing plant operation. These changes are 
considered administrative changes and do not modify, add, delete, or 
relocate any technical requirements in the TS.
    Therefore, the proposed changes do not create 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 will not reduce the margin of safety 
because they have no effect on any safety analyses assumptions. 
Changes to the TS Bases that result in meeting the criteria in 
paragraph (c)(2) of 10 CFR 50.59 will still require NRC approval. 
The proposed changes to TS 5.5.14 are considered administrative in 
nature based on the revision to 10 CFR 50.59.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

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

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 
Publicly available records will be accessible 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. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of application for amendment: October 29, 1999, as 
supplemented June 21, and September 8, 2000
    Brief description of amendment: The amendment revised the Technical

[[Page 2028]]

Specifications (TSs) to include: (1) the addition of operating limits 
for make-up tank (MUT) level and pressure; (2) the addition of 
surveillance requirements for the MUT pressure instrument channel; and 
(3) the revision of the calibration frequency for the MUT level 
instrument channel, the high- and low-pressure injection flow 
instrument channels, and the borated water storage tank instrument 
channel from ``Not to exceed 24 months'' to ``Refueling interval.'' 
Minor editorial changes and associated Bases changes were also made.
    Date of issuance: December 26, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 227.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Dates of initial notices in Federal Register: December 15, 1999 (64 
FR 70090) and July 12, 2000 (65 FR 43042). The September 8, 2000, 
letter provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination or expand 
the amendment beyond the scope of the original notices.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 26, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 26, 1999, as supplemented 
on January 20, 2000.
    Brief description of amendment: The proposed amendment would revise 
Technical Specifications (TSs) associated with the degraded voltage 
trip and the under-frequency reactor trip surveillance tests. For the 
degraded voltage trip, the proposed amendment would revise the TS to 
specify detailed operator actions to be taken if the minimum conditions 
could not be met rather than simply stating ``Cold Shutdown.'' The 6.9 
kV under-frequency and reactor trip surveillance tests currently 
combine voltage and frequency testing under one item. The proposed TS 
amendment would separate the 6.9 kV voltage testing from the frequency 
testing and specify separate test requirements. In addition, the 
proposed TS amendment would require more frequent testing of the 480 
volt emergency bus undervoltage reactor trip.
    Date of issuance: December 28, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 214.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 28, 2000 (65 
FR 10565) The January 20, 2000, submittal contained supplemental 
information that did not change the original no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 28, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 27, 2000.
    Brief description of amendment: This amendment changes the 
expiration date of the Operating License to 40 years from the date of 
issuance of the license rather than the date of the construction 
permit. Specifically, the amendment changes the expiration date of the 
Operating License from ``midnight on March 14, 2007'' to ``midnight on 
March 24, 2011.''
    Date of issuance: December 14, 2000.
    Effective date: As of the date of issuance.
    Amendment No.: 192.
    Facility Operating License No. DPR-20. Amendment revised the 
Operating License.
    Date of initial notice in Federal Register: May 17, 2000 (65 FR 
31352).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 14, 2000.
    No significant hazards consideration comments received: No. Other 
comments are addressed in the Commission's related Safety Evaluation.

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

    Date of application of amendments: September 12, 2000; supplemented 
October 4, October 26, November 6, and December 8, 2000.
    Brief description of amendments: The amendments revise the 
Technical Specification requirements related to the reroll repair 
process used to repair steam generator tubes. They also institute new 
license conditions.
    Date of Issuance: December 15, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 318/318/318.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 4, 2000 (65 FR 
59222) The supplements dated October 4, October 26, November 10, and 
December 8, 2000, provided clarifying information that did not change 
the scope of the September 12, 2000, application nor the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 15, 2000.
    No significant hazards consideration comments received: No

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

    Date of amendment request: November 23, 1999, as supplemented by 
letter dated October 12, 2000.
    Brief description of amendment: The amendment revises Technical 
Specifications to incorporate the use of American Society for Testing 
and Materials (ASTM) D3803-1989, ``Standard Test Method for Nuclear-
Grade Activated Carbon,'' into the River Bend Station, Unit 1, 
Technical Specifications.
    Date of issuance: December 20, 2000.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 115.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 8, 2000 (65 FR 
12291).
    The October 12, 2000, supplemental letter provided additional 
information to support staff review of the original application, and 
did not affect the initial finding of no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 20, 2000.
    No significant hazards consideration comments received: No.

[[Page 2029]]

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

    Date of application for amendment: November 23, 1999, as 
supplemented by letter dated October 19, 2000.
    Brief description of amendment: The amendment incorporated the use 
of American Society for Testing and Materials (ASTM) D3803-1989, 
``Standard Test Method for Nuclear-Grade Activated Carbon,'' into the 
Arkansas Nuclear One, Unit No. 2, Technical Specifications.
    Date of issuance: December 18, 2000.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 228.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 8, 2000 (65 FR 
12291).
    The October 19, 2000, supplemental letter provided clarifying 
information that was within the scope of the original Federal Register 
notice and did not change the staff's initial no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 18, 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: November 23, 1999, as supplemented by 
letter dated October 12, 2000.
    Brief description of amendment: The amendment incorporated the use 
of American Society of Testing and Materials D3803-1989, ``Standard 
Test Method for Nuclear-Grade Activated Carbon,'' into the facility's 
TS.
    Date of issuance: December 27, 2000.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 170.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 8, 2000, (65 FR 
12291).
    The October 12, 2000, supplement provided clarifying information 
that did not expand the scope of the original Federal Register notice, 
or change the scope of the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 27, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 23, 1999.
    Brief description of amendment: The amendment incorporated the use 
of American Society for Testing and Materials (ASTM) D3803-1989, 
``Standard Test Method for Nuclear-Grade Activated Carbon,'' into the 
Grand Gulf Nuclear Station Technical Specifications.
    Date of issuance: December 18, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 144.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 8, 2000 (65 FR 
12291)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 18, 2000.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant, 
Unit No. 1, St. Lucie County, Florida

    Date of application for amendment: April 23, 2000.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) Surveillance Requirement 4.6.4.2.b.4 by deleting the 
word ``immediately,'' in order to remove a timing restriction for the 
hydrogen recombiner post-operation resistance testing. As a result, the 
amendment allows the recombiner units to cool after an operational test 
run, and provides a more-reliable measurement of the resistance-to-
ground of the electrical insulation.
    Date of Issuance: December 27, 2000.
    Effective Date: December 27, 2000.
    Amendment No.: 169.
    Facility Operating License No. NPF-16: Amendment revised the TS.
    Date of initial notice in Federal Register: May 31, 2000 (65 FR 
34746)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 27, 2000.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No. 2, St. Lucie County, Florida

    Date of application for amendment: July 19, 2000.
    Brief description of amendment: Revised the Technical 
Specifications (TS) to extend the applicability of the current reactor 
coolant system pressure/temperature limits and allowed heatup and 
cooldown rates to 21.7 effective full power years of operation.
    Date of Issuance: December 28, 2000.
    Effective Date: December 28, 2000.
    Amendment No.: 112.
    Facility Operating License No. NPF-16: Amendment revised the TS.
    Date of initial notice in Federal Register: August 23, 2000 (65 FR 
51354). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 28, 2000.
    No significant hazards consideration comments received: No.
    Florida Power and Light Company, Docket Nos. 50-250 and 50-251, 
Turkey Point Plant, Units 3 and 4, Dade County, Florida
    Date of application for amendments: August 18, 2000.
    Brief description of amendments: The amendments consist of changes 
to the ACTION Statement 18 to allow operation of the units with both 
channels of undervoltage protection bypassed for up to 8 hours to allow 
performance of the monthly surveillance without placing the units in a 
condition not permitted by the Technical Specifications (TSs). In 
addition, the amendments authorize an administrative change to Item 
7.b. of TS Tables 3.3-2, 3.3-3, and 4.3-2 modifying ``Degraded Voltage 
`` to ``Undervoltage'' to make it consistent with the Updated Final 
Safety Analysis Report description.
    Date of issuance: December 20, 2000.
    Effective date: December 20, 2000.
    Amendment Nos. 209 and 203.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: September 20, 2000.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 20, 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: August 25, 2000, as supplemented 
November 20, 2000.

[[Page 2030]]

    Brief description of amendment: This amendment modifies Technical 
Specification (TS) 3.8.1.1, ``Electrical Power System--A.C. Sources--
Operating,'' by extending the allowed outage time (AOT) for Action a.2 
of TS 3.8.1.1 from 72 hours to 14 days, provided the Millstone Unit 3 
(MP3) station blackout diesel generator is available to supply 
Millstone Unit 2 (MP2) power, otherwise the AOT is only allowed to be 
extended for 7 days. This one-time change is needed to support the 
replacement of the MP2 4160-volt electrical cross-tie line from 
Millstone Unit 1 (MP1) with a cross-tie from MP3. The modification is 
being made due to the decommissioning of MP1.
    Date of issuance: December 21, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 251.
    Facility Operating License No. DPR-65: Amendment revised Technical 
Specifications.
    Date of individual notice in Federal Register: November 1, 2000 (65 
FR 65344).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 21, 2000.
    No significant hazards consideration comments received: No.

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 application for amendment: August 25, 2000.
    Brief description of amendment: The amendments authorize changes to 
the Millstone Nuclear Power Station, Unit Nos. 2 and 3 (MP2 and MP3) 
Final Safety Analysis Report (FSAR). Millstone Unit No. 1 (MP1) is 
being decommissioned. To support this activity, several modifications 
are required to modify/eliminate MP1 systems that support the operation 
of structures, systems, and components that are shared or common to MP2 
and MP3. One of the separation projects entails the replacement of the 
existing MP1 to MP2 4160-volt cross-tie with a new MP3 to MP2 4160-volt 
cross-tie. Northeast Nuclear Energy Company has evaluated this proposed 
new cross-tie utilizing the criteria of 10 CFR 50.59 and determined 
that the modification involved four unreviewed safety questions (USQs). 
One USQ pertains to MP2 and three USQs pertain to MP3.
    Date of issuance: December 21, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 252 and 190.
    Facility Operating License Nos. DPR-65 and NPF-49: Amendments 
authorize changes to the FSAR.
    Date of initial notice in Federal Register: October 20, 2000 (65 FR 
65345).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 21, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 26, 2000.
    Brief description of amendment: The amendments revise technical 
specifications (TSs) \3/4\.1.3.1, ``Reactivity Control Systems, Movable 
Control Assemblies, Full Length CEA Position'' and \3/4\.1.3.1, 
``Reactivity Control Systems, Movable Control Assemblies, Group 
Heights.'' Specifically, the changes revise the frequency for 
determining the operability of each rod not inserted fully in the core 
for Units 2 and 3 and the Deviation Circuit for Unit 2 from once every 
31 days to once every 92 days.
    Date of issuance: December 27, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 253 and 191.
    Facility Operating License Nos. DPR-65 and NPF-49: Amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 26, 2000 (65 FR 
46011)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 27, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 4, 2000, as supplemented 
August 31, October 5, and November 16, 2000.
    Brief Description of amendment: The amendment (1) adds new sections 
to the Technical Specifications (TSs) addressing missed surveillance 
test requirements and establishing a TS Bases control program, (2) 
revises TS Chapter 6 to allow use of generic personnel titles in lieu 
of plant-specific titles, (3) allows an alternative when the radiation 
protection manager does not meet the qualifications of Regulatory Guide 
1.8, (4) relocates sections of TS Chapter 6 pertaining to onsite and 
offsite review and special inspections to the Operational Quality 
Assurance Plan, and (5) corrects typographical errors.
    Date of issuance: December 21, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 115.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 31, 2000 (65 FR 
34749).
    The August 31, 2000, supplement provided updated TS pages to 
reflect incorporation of Amendment No. 110, which was issued subsequent 
to the May 4, 2000, application. In addition, a minor change in the 
proposed TS wording was proposed for consistency with the current TS. 
The October 5, 2000, supplement provided clarifying information to the 
May 4, 2000, application. The November 16, 2000, supplement proposed a 
minor wording change to be consistent with the latest revision of 
Standard TSs, NUREG-1433. The supplements were within the scope of the 
original Federal Register notice and did not change the staff's initial 
proposed no significant hazards considerations determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 21, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 15, 2000
    Brief description of amendments: The amendments revise station 
technical specification TS.3.7.B.6 to explicitly allow de-energizing 
motor control center (MCC) 1T1 or MCC 1T2 for up to 72 hours to 
accommodate installation of transfer switches for the MCCs.
    Date of issuance: December 15, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 155 and 146.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 12, 2000 (65 FR 
43049 )

[[Page 2031]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 15, 2000.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, 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: December 29, 1999, as 
supplemented on November 21, 2000
    Brief description of amendments: The amendments modify the Salem 
Unit Nos. 1 and 2 Technical Specifications (TS), and revise 
requirements stated in Notes 1 and 2 to Table 2.2-1, ``Reactor Trip 
System Instrumentation Setpoints,'' in order to add a tolerance 
associated with the setpoint values for the derivative module time 
constants (the Tau values) of the Over-Power, and the Over-Temperature 
delta temperature units.
    Date of issuance: December 19, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days of issuance.
    Amendment Nos.: 239 and 220.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4289).
    The November 21, 2000, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 20, 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: September 22, 2000 (PCN-520).
    Brief description of amendments: The amendments revise Technical 
Specifications (TSs) 3.1.10, 3.3.9, 3.3.13, 3.4.5, 3.4.6, 3.4.7, 3.4.8, 
3.8.2, 3.8.5, 3.8.8, 3.8.10, 3.9.2, 3.9.4 and 3.9.5 to allow small, 
controlled, safe insertions of positive reactivity while in shutdown 
modes.
    Date of issuance: December 20, 2000.
    Effective date: December 20, 2000, to be implemented within 30 days 
of issuance.
    Amendment Nos.: Unit 2--175; Unit 3--166.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the TSs.
    Date of initial notice in Federal Register: October 13, 2000 (65 FR 
60984).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 20, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 31, 2000 (TS 00-05).
    Brief description of amendments: These amendments revised the 
Technical Specifications (TSs) by relocating various reactivity control 
system requirements from the TSs to the Sequoyah Technical Requirements 
Manual.
    Date of issuance: December 18, 2000.
    Effective date: December 18, 2000.
    Amendment Nos.: 264 and 255.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: October 4, 2000 (65 FR 
59226).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 18, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 31, 2000 (TS 99-17).
    Brief description of amendments: These amendments revised the 
Technical Specifications (TSs) by adding new requirements for 
maintaining soluble boron in the spent fuel pool.
    Date of issuance: December 19, 2000.
    Effective date: December 19, 2000.
    Amendment Nos.: 265 and 256.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: October 18, 2000 (65 FR 
62392).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 19, 2000.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 3rd day of January 2001.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-596 Filed 1-9-01; 8:45 am]
BILLING CODE 7590-01-U