[Federal Register Volume 66, Number 104 (Wednesday, May 30, 2001)]
[Notices]
[Pages 29349-29367]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-13400]


-----------------------------------------------------------------------

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 May 7, 2001 through May 18, 2001. The last 
biweekly notice was published on May 16, 2001 (66 FR 27174).

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 Public

[[Page 29350]]

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 June 29, 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, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications 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 and electronically from 
the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room).

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of amendments request: May 1, 2001.
    Description of amendments request: The proposed amendments would 
revise the pressure-temperature limits curves contained in Technical 
Specification 3.4.9, ``RCS Pressure and Temperature (P/T) Limits.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

1. The Proposed License Amendments Do Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The changes to the calculation methodology for the pressure-
temperature limits are based on American Society of Mechanical 
Engineers (ASME) Code Case N-640, ``Alternative Reference Fracture 
Toughness for Development of P-T Limit Curves for ASME Section XI, 
Division 1,'' and provide adequate margin in the prevention of a 
non-ductile type fracture of

[[Page 29351]]

the reactor pressure vessel. The code case was developed based upon 
the knowledge gained through years of industry experience. The 
pressure-temperature limits developed using the allowances of ASME 
Code Case N-640 provide more operating margin. However, experience 
gained in the areas of fracture toughness of materials and pre-
existing undetected defects shows that some of the existing 
assumptions used for the calculation of pressure-temperature limits 
are unnecessarily conservative and unrealistic. Therefore, use of 
the allowances of ASME Code Case N-640 in developing the pressure-
temperature limits will provide adequate protection against 
nonductile-type fractures of the reactor pressure vessel.
    Development of the revised BSEP [Brunswick Steam Electric 
Plant], Unit 1 and 2 pressure-temperature limits was performed using 
the approved methodologies of 10 CFR 50, Appendix G, and using the 
allowances of ASME Code Case N-640. The pressure-temperature limits 
generated using these methods ensure the pressure-temperature limits 
will not be exceeded during any phase of reactor operation. 
Therefore, the probability of occurrence and the consequences of a 
previously analyzed event are not significantly increased. Finally, 
the proposed changes will not affect any other system or piece of 
equipment designed for the prevention or mitigation of previously 
analyzed events.
    Thus, the probability of occurrence and the consequences of any 
previously analyzed event are not significantly increased as the 
result of the proposed changes to the pressure-temperature limits.

2. The Proposed License Amendments Will Not Create the Possibility of a 
New or Different Kind of Accident From Any Accident Previously 
Evaluated

    The proposed changes provide more operating margin in the 
pressure-temperature limits for hydrostatic pressure testing, non-
nuclear heatup and cooldown, and criticality, with the benefits 
being primarily realizable during the pressure tests. The changes 
also extend the pressure-temperature limits for use up to 32 EFPY 
[effective full-power years] of operation. However, operation in the 
``new'' regions of the pressure-temperature limits has been analyzed 
and will provide adequate protection against a nonductile-type 
fracture of the reactor pressure vessel. Otherwise, the proposed 
pressure-temperature limits do not result in any new or unanalyzed 
operation of any system or piece of equipment important to safety 
and, as a result, the possibility of a new type event is not 
created.

3. The Proposed License Amendments Do Not Involve a Significant 
Reduction in a Margin of Safety

    The revised pressure-temperature limits provide more operating 
margin and operational flexibility than the existing pressure-
temperature limits. With the increased operational margin, a 
reduction in the safety margin results with respect to the existing 
limits. However, the industry experience since the inception of 
pressure-temperature limits confirms that some of the existing 
methodologies used to develop pressure-temperature limits are 
unrealistic and unnecessarily conservative. Accordingly, ASME Code 
Case N-640 takes advantage of this acquired knowledge by 
establishing more realistic methodologies for the development of 
pressure-temperature limits. Therefore, operational flexibility is 
gained and an acceptable margin of safety to reactor pressure vessel 
non-ductile type fracture is maintained. Evaluation of the revised 
pressure-temperature limits for use up to 32 EFPY was performed 
using 10 CFR 50 and ASME Code Case N-640; thus, the margin of safety 
is not significantly reduced as the result of the proposed changes.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Patrick M. Madden, Acting.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: April 26, 2001.
    Description of amendment request: The proposed amendment would add 
Section 6.22, ``Reactor Coolant Pump Flywheel Inspection Program'' to 
Section 6, ``Administrative Controls'' of the Technical Specifications 
(TSs) and relocate the requirements of TS 3/4.4.10, ``Reactor Coolant 
System, Structural Integrity'' to the Millstone Unit No. 2 Technical 
Requirements Manual (TRM). The Bases of the affected TSs would also be 
relocated to the TRM. The Index pages would also be updated to reflect 
these 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:

1. Involve a Significant Increase in the Probability or Consequences of 
An Accident Previously Evaluated

    Missile generation from a Reactor Coolant Pump (RCP) flywheel 
could damage the Reactor Coolant System, the Containment, or other 
equipment or systems important to safety. The fracture mechanics 
analyses conducted to support the change to Inservice Inspection 
(ISI) requirements in accordance with the proposed Section 6.22, 
``Reactor Coolant Pump Flywheel Inspection Program'' shows that a 
pre-existing crack sized just below the detection level will not 
grow to the flaw size necessary to create flywheel missiles within 
the life of the plant. This analysis conservatively assumes minimum 
material properties, maximum flywheel accident speed, location of 
the flaw in the highest stress area, and a number of startup/
shutdown cycles eight times greater than expected. Since an existing 
flaw in a Millstone Unit No. 2 flywheel will not grow to the 
allowable flaw size under Loss of Coolant Accident (LOCA) conditions 
over the life of the plant, reducing the ISI requirements for the 
detection of such cracks over the life of the plant will not 
significantly increase the probability or consequences of an 
accident previously evaluated
    The proposed Technical Specification changes to relocate the 
requirements for Technical Specification 3/4.4.10, ``Reactor Coolant 
System, Structural Integrity'' (with the exception of the RCP 
inspection requirements) to the TRM will have no adverse effect on 
plant operation or the availability or operation of any accident 
mitigation equipment. Therefore, the Reactor Coolant System 
structural integrity (with the exception of the RCP flywheel which 
is addressed above) will not adversely impact an accident initiator 
and can not cause an accident. Therefore these changes will not 
increase the probability or consequences of an accident previously 
evaluated.
    The Index pages will be updated to reflect the proposed changes. 
These changes are administrative in nature. These changes will not 
increase the probability or consequences of an accident previously 
evaluated.

2. Create the Possibility of a New or Different Kind of Accident From 
Any Accident Previously Evaluated

    The proposed changes will not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. They do not alter the way any 
structure, system, or component functions and do not alter the 
manner in which the plant is operated. These changes do not 
introduce any new failure modes. Therefore, the proposed changes 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

3. Involve a Significant Reduction in a Margin of Safety

    The fracture mechanics analyses conducted to support the change 
to ISI requirements in accordance with the proposed Section 6.22, 
``Reactor Coolant Pump Flywheel Inspection Program'' shows that 
significant conservatism has been used for calculating the allowable 
flaw size, critical flaw size, and crack growth rate in the RCP 
flywheels. These include minimum material properties, maximum 
flywheel accident speed, location of the flaw in the highest stress 
area and a number of startup/shutdown cycles eight times greater 
than expected. Since an existing flaw in a Millstone Unit No. 2 
flywheel will not grow to the allowable flaw size under normal 
operating conditions or to the critical flaw size under LOCA 
conditions over the life of the plant, reducing ISI requirements for 
the

[[Page 29352]]

detection of such cracks over the life of the plant will not involve 
a significant reduction in the margin of safety. The proposed 
changes have no impact on plant equipment operation. Therefore, the 
proposed changes will not result in a reduction in a margin of 
safety.
    Relocation of Technical Specification 3/4.4.10 (whole 
specification except the portion specifying surveillance requirement 
for the RCP flywheel) to the TRM does not imply any reduction in its 
importance in ensuring that the structural integrity and operational 
readiness of ASME Code Class 1, 2, and 3 components will be 
maintained at an acceptable level throughout the life of the plant. 
The proposed change has no impact on plant equipment operation. 
Therefore, the proposed change will not result in a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: January 27, 2000, as supplemented by 
letter dated March 1, 2001.
    Description of amendment request: The proposed changes to Arkansas 
Nuclear One, Unit 1 (ANO-1) and Unit 2 (ANO-2), Technical 
Specifications (TSs) allow for the qualified condensate storage tank 
(QCST) to be used for both units as the preferred source of water for 
emergency feedwater (EFW). Currently, the QCST is aligned to the ANO-1 
EFW system while ANO-2 relies on non-safety related tanks and an 
automatic switchover to the Service Water System as the source of EFW 
coolant water.
    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 staff's analysis is 
presented below.
Criterion 1--Does Not Involve a Significant Increase in the Probability 
or Consequences of an Accident Previously Evaluated
    The condensate storage tanks provide a source of condensate grade 
water for the EFW System. The tanks, one for ANO-1 and two for ANO-2, 
are already included in the plant TSs. The proposed change allows for 
both units to operate while aligned to the QCST, but does not affect 
the physical design, construction, or operation of the condensate 
storage tanks. These tanks are not associated with the precursors of 
any accident. This change does not increase the probability of any 
accident previously evaluated.
    As a source of EFW, the tanks serve an accident mitigation 
function. The proposed change does not alter this function. In addition 
to the tanks, the Service Water System is also available as a long-term 
assured source of EFW. The proposed change allows the use of the QCST 
as the preferred source of EFW for both units. The combination of 
available sources of water for EFW assures that both units are able to 
respond to accidents previously evaluated. Because this function 
continues to be assured, the proposed changes do not increase the 
consequences of a previously evaluated accident.
    Therefore, this change does not involve a significant increase in 
the probability or consequences of any accident previously evaluated.
Criterion 2--Does Not Create the Possibility of a New or Different Kind 
of Accident From any Accident Previously Evaluated
    The condensate storage tanks serve an accident mitigation function 
as a temporary source of EFW. These tanks have not been identified as a 
precursor to any accident previously evaluated. The design and 
operation of these tanks have not changed. While the proposed change 
does permit the qualified tank to be used by ANO-2, the design has been 
evaluated and it has been demonstrated that the existing tank is 
capable of meeting the intended design function of both units.
    Therefore, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
Criterion 3--Does Not Involve a Significant Reduction in a Margin of 
Safety
    The existing sources of water for the ANO-1 EFW system will 
continue to ensure adequate EFW system performance after the proposed 
change. The QCST and, if necessary, the Service Water System will 
ensure that the EFW system performs to maintain margins of safety. The 
Service Water System is the assured long-term source of cooling water 
for both units. The safety function of decay heat removal and core 
cooling continues to be met. There is no reduction in the margin of 
safety for ANO-1.
    The proposed change to the ANO-2 specifications will provide a 
qualified alternative source of EFW. The required function of the tanks 
is the same as for ANO-1; that is, to provide a source of water until 
the unit can successfully transfer to decay heat cooling or until the 
Service Water System is aligned for long-term cooling. The addition of 
this QCST to the specification as an alternative to the existing tanks 
does not decrease the margin of safety.
    Therefore, this change does not involve a significant reduction in 
a margin of safety.
    Based on this analysis, it appears that the three standards of 10 
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: May 2, 2001.
    Description of amendment request: The proposed amendment would 
relocate the requirements for the containment recirculation system from 
the technical specifications to the Technical Requirements Manual 
(TRM).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

1. 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?

    The containment recirculation fans, along with the containment 
cooling units and containment spray systems, provide a means of 
circulating the containment atmosphere to ensure adequate mixing of 
the containment atmosphere. The containment cooling units and 
containment spray systems are safety-related systems and required by 
TS 3.6.2.3 and 3.6.2.1, respectively. Adequate air mixing is assured 
with the use of these two systems. The containment recirculation 
fans are not credited in the mitigation of any accidents.
    Based on an evaluation of the criteria listed in 10 CFR 
50.36(c)(2)(ii), the relocation of the

[[Page 29353]]

containment recirculation fans to the TRM is acceptable.
    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?

    The containment recirculation fans are not accident initiators. 
The function they fulfill will continue to be maintained by the 
containment cooling units and containment spray pumps. Because the 
proposed amendment will not change the design, configuration or 
method of plant operation, it will not create the possibility of a 
new or different kind of accident.
    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?

    Air mixing of the containment atmosphere can be accomplished 
following a LOCA [loss-of-coolant accident] by the containment 
recirculation fans, the containment cooling units, or the 
containment spray systems. Any one of these systems is capable of 
providing adequate air mixing. The proposed change does not change 
the design function of the containment recirculation fans. 
Additionally, the containment recirculation fans are not credited in 
any accident analysis. Since adequate mixing of the containment 
atmosphere is credited through the containment cooling units and 
spray systems, relocation of the containment recirculation fan 
requirements to the TRM does not result in any impact to the margin 
of safety.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

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

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

    Date of amendment request: May 3, 2001.
    Description of amendment request: The proposed amendment deletes 
requirements from the Technical Specifications (TSs) (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 Nuclear Station] 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 TSs 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 Nuclear Regulatory Commission (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 license amendment applications 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 May 3, 2001.
    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 [Three Mile Island Nuclear Station, Unit 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 (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

[[Page 29354]]

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

Exelon Generation Company, LLC, PSEG Nuclear LLC, and Atlantic City 
Electric Company, Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic 
Power Station, Units Nos. 2 and 3, York County, Pennsylvania

    Date of application for amendments: February 8, 2001.
    Description of amendment request: The proposed amendment would 
revise the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3, 
technical specifications (TSs) and the associated TS Bases, to reflect 
changes to support the activation of the trip outputs of the 
oscillation power range monitor (OPRM) portion of the power range 
neutron monitoring (PRNM) system and delete the interim corrective 
action requirements from the TSs. The OPRM trip function provides 
protection from exceeding the fuel minimum critical power ratio (MCPR) 
safety limit in the event of thermal-hydraulic power oscillations. 
PBAPS is currently operating under interim corrective actions that 
specify restrictions on plant operations and actions by operators in 
response to power oscillations. The OPRM system provides an automatic 
reactor trip which eases the burden on the operators if power 
oscillations were to occur.
    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 Amendment Does Not Involve a Significant Increase in 
the Probability or Consequences of an Accident Previously Evaluated
    This modification has no impact on any of the existing PRNM 
functions. It connects the OPRM trip function to the reactor protection 
system; connects the associated trip alarm to the annunciator 
circuitry; updates the TSs to add the OPRM-related functions and to 
delete Interim Corrective Actions (ICAs) related requirements; and 
revises affected procedures.
    The ICAs include a restricted region on the power-to-flow map where 
thermal-hydraulic instabilities were known to be more likely. Operation 
in the restricted region requires more frequent monitoring of the 
average power range monitors (APRMs) and local power range monitors 
(LPRMs), which are part of the PRNM system. This restricted region is 
less than 10 percent of the full power-to-flow map. Plant operation in 
portions of the former restricted region without the increased 
monitoring of APRMs and LPRMs previously required by the ICAs may cause 
a slight increase in the probability of occurrence of an instability. 
This potential increase in probability is not significant because 
operation in this region will still result in a low likelihood of core 
power oscillations. Because of the more reliable detection of an 
instability event, should it occur, the automatic scram if preset 
limits are exceeded, and the elimination of dependence on the operator, 
the consequences of an instability event are not increased with this 
modification.
    Based on the above discussion, the proposed amendment does not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
2. The Proposed Amendment Does Not Create the Possibility of a New or 
Different Kind of Accident From Any Accident Previously Evaluated
    Enabling the OPRM reactor scram function does not create any new 
system interactions except for the reactor scram function. The failure 
modes for the new OPRM circuits would be to initiate a reactor scram 
unnecessarily, or to fail to initiate a reactor scram when 
instabilities were present. These failures would not create the 
possibility of a new or different kind of accident. Since the present 
system has no automatic reactor scram for instabilities, the operators 
insert a manual scram if necessary, and the effect of core 
instabilities has been analyzed. The use of a manual scram is still 
available with the OPRM scram function enabled. Removing the ICAs from 
the TSs does not create the possibility of a new or different kind of 
accident, since the effect of core instabilities has been evaluated.
    Based on the above discussion, the proposed amendment will not 
create the possibility of a new or different kind of accident from any 
accident previously evaluated.
3. The Proposed Amendment Does Not Involve a Significant Reduction in a 
Margin of Safety.
    The current safety analyses assume that the existing ICA related TS 
requirements are adequate to prevent exceeding the MCPR safety limit 
due to an instability event. As a result, there is currently no 
quantitative or qualitative assessment of an instability event with 
respect to its impact on MCPR.
    The OPRM trip function is being implemented to automate the 
detection (via direct measurement of neutron flux) and subsequent 
suppression (via reactor scram) of an instability event prior to 
exceeding the MCPR safety limit. The OPRM trip provides a trip output 
of the same type as currently used for the APRMs. Its failure modes and 
types are identical to those for the present APRM output. Currently, 
the MCPR safety limit is not impacted by an instability event since the 
event is mitigated by manual means via the ICAs. In both methods of 
mitigation (manual and automated), the margin of safety associated with 
the MCPR safety limit is maintained.
    Therefore, based on the fact that the MCPR safety limit will not be 
exceeded as a result of an instability event following implementation 
of the OPRM trip function in place of the existing manual ICAs, it is 
concluded that the proposed amendment does not reduce a margin of 
safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for Licensee: Mr. Edward Cullen, Vice President and 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.
    NRC Section Chief: James W. Clifford.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, 
Beaver Valley Power Station, Unit No. 1 (BVPS-1), Shippingport, 
Pennsylvania

    Date of amendment request: March 28, 2001.

[[Page 29355]]

    Description of amendment request: The requested amendment proposes 
changes to the BVPS-1 Technical Specifications (TSs) associated with 
the reductions of the reactor coolant system and secondary coolant 
system specific activity limits. These TS changes support a revised 
main steam line break safety analysis with a higher assumed primary-to-
secondary leak rate in accordance with the methodology described in 
Nuclear Regulatory Commission (NRC) Generic Letter (GL) 95-05, 
``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes 
by Outside Diameter Stress Corrosion Cracking.'' TS Bases and other 
administrative changes are proposed for consistency.
    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. Does the Change Involve a Significant Increase in the Probability or 
Consequences of an Accident Previously Evaluated?
    The proposed change involves the reduction of the Technical 
Specification Dose Equivalent Iodine 131 (I-131) activity limits for 
the reactor coolant system (RCS) and the secondary system which 
facilitates an increase in the assumed accident-induced primary-to-
secondary leak rate in the event of a postulated main steam line break 
(MSLB) accident. There are no proposed changes to any facility 
structures, systems, or components. The proposed changes do not affect 
any initiators of accidents previously evaluated nor does the proposed 
change introduce any new failure mechanisms that may initiate a 
previously-evaluated accident. Furthermore, the proposed change would 
not affect the ability of any accident mitigation system to perform its 
design-basis function as defined in the Updated Final Safety Analysis 
Report (UFSAR). The dose consequence analysis for a postulated MSLB 
accident are being revised as part of this amendment and the resulting 
calculated dose consequences do not increase.
    Therefore, the proposed 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 is only associated with the reduction of the 
RCS and secondary system I-131 activity limits and does not involve 
changes to any facility structures, systems, or components. There are 
no proposed changes to the facility or its operation. Since there are 
no changes being made to any structures, systems, or components, no new 
failure mechanisms are introduced by the proposed changes that would 
result in the occurrence of a new or different kind of accident from 
any accident previously evaluated. The accident analyses contained in 
the UFSAR continue to remain bounding with regard to the spectrum of 
possible accidents.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
3. Does the Change Involve a Significant Reduction in a Margin of 
Safety?
    There are no proposed changes to any structure, systems, or 
components. Changes proposed to the dose consequence analysis for a 
postulated MSLB accident are included in the amendment request. The 
reduction in the RCS and secondary system activity limits are being 
made to offset the effects of an increased accident-induced primary-to-
secondary leak rate resulting from a postulated MSLB accident in 
accordance with GL 95-05. The margins to safety that could be affected 
are those associated with the resulting calculated doses to the public 
and facility personnel. However, the dose-decreasing effect of lowering 
the activity limits offsets the dose-increasing effect of raising the 
assumed accident-induced primary-to-secondary leak rate. Consequently, 
the resulting calculated doses do not increase.
    Therefore, the proposed change does not involve a significant 
reduction in a margin to 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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Richard P. Correia (Acting).

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

    Date of amendment request: March 30, 2001.
    Description of amendment request: The proposed amendment would 
involve changes to Technical Specification (TS) 3/4.5.2, Emergency Core 
Cooling--ECCS Subsystems--T<INF>avg</INF> <gr-thn-eq> 280 deg.F.
    Technical Specification Limiting Condition for Operation (LCO) 
3.5.2 requires two independent Emergency Core Cooling Systems (ECCS) 
Subsystems to be operable. Surveillance Requirement (SR) 4.5.2.f 
requires each ECCS subsystem to be demonstrated operable by performing 
a vacuum leakage rate test of the watertight enclosure for Decay Heat 
Removal System valves DH-11 and DH-12 that assures the motor operator 
on valves DH-11 and DH-12 will not be flooded for at least (7) days 
following a Loss-of-Coolant Accident (LOCA). The test is required to be 
performed: (1) At least once per 18 months, (2) After each opening of 
the watertight enclosure, and (3) After any maintenance on or 
modification to the watertight enclosure which could affect its 
integrity.
    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:

    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no such accidents are affected 
by the proposed changes. Initial conditions and assumptions remain 
as previously analyzed for accidents in the Davis-Besse Nuclear 
Power Station Updated Safety Analysis Report.
    The proposed changes would increase the surveillance test 
interval in Technical Specification 4.5.2.f.1 from 18 to 24 months 
for the vacuum leakage rate test of the watertight enclosure for 
Decay Heat Removal System valves DH-11 and DH-12. The surveillance 
data and maintenance records have been reviewed and support an 
increase in the surveillance test interval from 18 to 24 months 
based on the low potential for a significant increase in the failure 
rate of the watertight enclosure due to an increased surveillance 
interval, and based on the introduction of no new failure modes. The 
proposed change to the surveillance interval has been evaluated 
consistent with the NRC guidance on evaluating and proposing such 
revisions as provided in Generic Letter 91-04, ``Changes in 
Technical Specification Surveillance Intervals to Accommodate a 24-
Month Fuel Cycle,'' dated April 2, 1991. The watertight enclosure 
and its condition do not contribute to the initiation of any 
accident.

[[Page 29356]]

Therefore, the probability of any accident previously evaluated is 
not increased.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the integrity of the 
watertight enclosure sealing mechanisms has been evaluated, and it 
has been determined that the sealing mechanisms will remain intact 
for the proposed increased surveillance interval. Therefore, there 
is assurance that the backup boric acid precipitation control flow 
path will remain available, so that there will be no impact on the 
source term, containment isolation or radiological releases.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because the proposed 
changes do not alter the manner in which the watertight enclosure is 
sealed or tested, and the operability requirements of Decay Heat 
Removal System valves DH-11 and DH-12 will continue to be adequately 
addressed by Surveillance Requirement 4.5.2.f.1.
    No changes are being proposed to the type of testing currently 
being performed, only to the length of the surveillance test 
interval. An increase in the surveillance test interval from 18 to 
24 months is justified based on the low potential for a significant 
increase in the failure rate of the watertight enclosure due to an 
increased surveillance interval, and based on the introduction of no 
new failure modes.
    No different accident initiators or failure mechanisms are 
introduced by the proposed change. Thus, it does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Not involve a significant reduction in a margin of safety.
    An increase in the surveillance test interval from 18 to 24 
months is justified based on the low potential for a significant 
increase in the failure rate of the watertight enclosure due to an 
increased surveillance interval, and based on the introduction of no 
new failure modes.
    Since there are no new or significant changes to the initial 
conditions contributing to accident severity or consequences, there 
are no significant reductions in a margin of safety.
    On the basis of the above, the Davis-Besse Nuclear Power Station 
has determined that the License Amendment Request does not involve a 
significant hazards consideration. As this License Amendment Request 
concerns a proposed change to the Technical Specifications that must 
be reviewed by the Nuclear Regulatory Commission, this License 
Amendment Request does not constitute an unreviewed safety question.

    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.

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

    Date of amendment request: April 1, 2001.
    Description of amendment request: The proposed amendment would add 
new Technical Specification (TS) Administrative Controls Section 6.17, 
TS Bases Control Program, and make a related change to the TS Index. 
The proposed new TS Administrative Control would provide requirements 
for changing and updating the TS Bases. This proposed new TS is similar 
to the Specification 5.5.14 of NUREG-1430, ``Standard Technical 
Specifications--Babcock and Wilcox Plants,'' Revision 1, April 1995.
    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:

    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no such accidents are affected 
by the proposed changes. The amendment application proposes to add a 
new Technical Specification (TS) Administrative Controls Section 
6.17, ``Technical Specifications (TS) Bases Control Program,'' and 
to make a related change to the TS Index. The proposed changes do 
not involve a change to any structure, system, or component or to 
the assumptions of any accident analyses.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because no equipment, accident 
conditions, or assumptions are affected which could lead to a 
significant increase in radiological consequences.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new or 
different accident initiators are introduced by these proposed 
changes.
    3. Not involve a significant reduction in a margin of safety 
because there are no new or significant changes to the initial 
conditions contributing to accident severity or consequences. 
Consequently, there are no significant reductions in a margin of 
safety.
    On the basis of the above, the DBNPS has determined that the 
License Amendment Request does not involve a significant hazards 
consideration. As this License Amendment Request concerns a proposed 
change to the Technical Specifications that must be reviewed by the 
Nuclear Regulatory Commission, this License Amendment Request does 
not constitute an unreviewed safety question.

    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.

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

    Date of amendment request: April 1, 2001.
    Description of amendment request: The proposed amendment would 
involve changes to Technical Specification (TS) 3/4.3.1, Reactor 
Protection System Instrumentation; 3/4.3.2.1, Safety Features Actuation 
System Instrumentation; TS 3/4.3.2.2, Steam and Feedwater Rupture 
Control System Instrumentation; and Bases 3/4.3.1 and 3/4.3.2, Reactor 
Protection System and Safety System Instrumentation.
    The proposed changes would revise TS Table 3.3-3, Safety Features 
Actuation System (SFAS) Instrumentation, TS Table 3.3-11, Steam and 
Feedwater Rupture Control System (SFRCS) Instrumentation, and 
associated Bases to add a provision to allow an eight-hour delay in 
entering an Action statement when an SFAS or SFRCS instrumentation 
channel is undergoing Channel Functional Testing. The proposed changes 
would provide a reasonable time to perform the required surveillance 
testing and relieve the control room staff of the burden of making 
multiple Action statement entries and exits in order to complete the 
testing. Additionally, Surveillance Requirements 4.3.1.1.2, 4.3.2.1.2, 
and 4.3.2.2.2 would be revised to clarify the term ``total bypass 
function.''
    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:

    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no such accidents are affected 
by the proposed changes. The amendment application proposes to add a 
provision to TS Table 3.3-3, Safety Features Actuation System (SFAS) 
Instrumentation, and TS Table 3.3-11, Steam and Feedwater

[[Page 29357]]

Rupture Control System (SFRCS) Instrumentation, to permit certain 
SFAS and SFRCS instrument channels to [be] placed in an inoperable 
condition for up to 8 hours during surveillance testing without 
declaring the channel inoperable and entering the Action statement. 
This proposed change would reduce burden placed on the control room 
operators and is essentially administrative in nature. The proposed 
change to the TS Bases 3/4.3.1 and 3/4.3.2, Reactor Protection 
System and Safety System instrumentation, is associated with the 
changes to TS Tables 3.3-3 and 3.3-11. These changes will not 
significantly change testing methodology, system unavailability, or 
system reliability. Initiating conditions and assumptions remain as 
previously analyzed for accidents in the DBNPS Updated Safety 
Analysis Report (USAR).
    The proposed changes to Limiting Condition for Operation 
3.3.1.1, Surveillance Requirement (SR) 4.3.1.1.2, SR 4.3.2.1.2, and 
SR 4.3.2.2.2 to clarify the nomenclature of the Reactor Protection 
System (RPS), SFAS, and SFRCS bypass functions being tested are 
administrative in nature. These changes will not effect any plant 
hardware or the performance of any test. Initiating conditions and 
assumptions remain as previously analyzed for accidents in the DBNPS 
USAR.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the source term, containment 
isolation, or radiological releases are not affected by the proposed 
changes. Existing system and component redundancy is not affected by 
the proposed changes. The existing system and component operation is 
not affected by the proposed changes, and the assumptions used in 
evaluating the radiological consequences in the DBNPS USAR are not 
invalidated. Therefore, for each postulated accident the 
consequences remain bounded by the consequences from the previously 
evaluated accidents.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because these 
proposed changes do not involve any physical changes to systems or 
components, nor do they alter the manner in which the systems or 
components are operated. No new or different accident initiators or 
equipment failure modes are introduced by the proposed changes.
    3. Not involve a significant reduction in a margin of safety 
because, for the proposed changes, there are no new or significant 
changes to the initial conditions contributing to accident severity 
or consequences. Accordingly, there are no significant reductions in 
a margin of safety.
    On the basis of the above, the DBNPS has determined that the 
License Amendment Request does not involve a significant hazards 
consideration. As this License Amendment Request concerns a proposed 
change to the Technical Specifications that must be reviewed by the 
Nuclear Regulatory Commission, this License Amendment Request does 
not constitute an unreviewed safety question.

    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.

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

    Date of amendment request: April 1, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Surveillance Requirement (SR) 
4.0.5, Applicability, TS Bases 4.0.5, and TS Bases 3/4.4.2 and 3/4.4.3, 
Reactor Coolant System--Safety Valves, Regarding Inservice Testing 
Requirements.
    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:

    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no such accidents are affected 
by the proposed changes. The amendment application proposes to 
revise DBNPS Technical Specification (TS) Surveillance Requirement 
4.0.5, Applicability, and its associated Bases and TS Bases 3/4.4.2 
and 3/4.4.3, Reactor Coolant System--Safety Valves. The proposed 
changes would modify the Technical Specifications to conform to the 
requirements of Section 50.55a(f) of Title 10 of the Code of Federal 
Regulations regarding the inservice testing of pumps and valves for 
the third and successive 120-month intervals. The current DBNPS TS 
reference the American Society of Mechanical Engineers Boiler and 
Pressure Vessel Code (ASME Code), Section XI requirements for the 
inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. 
The proposed changes would reference the ASME Code for Operation and 
Maintenance of Nuclear Power Plants (ASME OM Code) which is 
consistent with Section 50.55a(f).
    In addition, surveillance interval definitions for ``semi-
quarterly,'' ``every 9 months,'' and ``biennially or every 2 
years,'' as used in the ASME Code would be added to TS 4.0.5.b to 
ensure consistent interpretation of the terms. The proposed changes 
do not affect any plant hardware and do not affect the probability 
of any equipment malfunction or accident-initiating event.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because no equipment, accident 
conditions, or assumptions are affected which could lead to a 
significant increase in radiological consequences.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new or 
different accident initiators are introduced by these proposed 
changes.
    3. Not involve a significant reduction in a margin of safety 
because there are no changes to the initial conditions contributing 
to accident severity or consequences. Consequently, there are no 
significant reductions in a margin of safety.
    On the basis of the above, the DBNPS has determined that the 
License Amendment Request does not involve a significant hazards 
consideration. As this License Amendment Request concerns a proposed 
change to the Technical Specifications that must be reviewed by the 
Nuclear Regulatory Commission, this License Amendment Request does 
not constitute an unreviewed safety question.

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

    Date of amendment request: April 17, 2001.
    Description of amendment request: The proposed amendments would 
implement minor changes and corrections to the Technical Specifications 
(TS) to correct administrative errors (e.g., typographical, amendment 
tracking number, etc.), or to incorporate changes that have been 
justified by previously approved license amendments and should have 
been made as part of those submittals, or to correct logic errors 
(e.g., TS operating mode breakpoints based on pressurizer pressure and 
not temperature). Also, the proposed amendments would revise the Units 
1 and 2 TS to delete obsolete terminology and provide conforming 
changes to reflect the recently implemented change 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

[[Page 29358]]

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

    These proposed license amendments require no plant hardware or 
operational modifications. The proposed changes either correct 
various administrative errors (e.g., typographical errors, amendment 
tracking number errors), incorporate changes that have been 
justified by previously approved license amendments and should have 
been made as part of those submittals, correct logic errors, or are 
necessary to implement the 10 CFR 50.59 rule change.
    Therefore, operation of the facility in accordance with the 
proposed amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (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 Accident Previously Evaluated
    No modifications to either plant hardware or operational 
procedures are required to support these proposed license 
amendments; hence, no new failure modes are created. The proposed 
changes either correct various administrative errors (e.g., 
typographical errors, amendment tracking number errors), incorporate 
changes that have been justified by previously approved license 
amendments and should have been made as part of those submittals, 
correct logic errors, or are necessary to implement the 10 CFR 50.59 
rule change.
    Therefore, operation of the facility in accordance with the 
proposed amendments would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

(3) Operation of the Facility in Accordance With the Proposed Amendment 
Would Not Involve a Significant Reduction in a Margin of Safety

    The majority of TS corrections proposed by these license 
amendments are administrative in nature in that they either correct 
typographical errors (e.g., ODCM verses OCDM), are justified by 
previous license amendments (e.g., surveillance requirements for 
T<INF>hot</INF> wide versus narrow range instrumentation), or 
correct logic errors (e.g., ECCS subsystem TS headings based on 
operating mode, with Mode 3 breakpoints based on pressurizer 
pressure and not temperature). The overly restrictive emergency 
power requirements for non critical single train quality related 
radiation monitors are being removed, while critical radiation 
monitor emergency power requirements are unaffected by the change.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant reduction in a 
margin of safety.

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

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

    Date of amendment request: April 18, 2001.
    Description of amendment request: The proposed amendment would 
implement an improved heat flux correlation (designated ABB-NV) 
previously approved by the NRC for Westinghouse-Combustion Engineering, 
as documented in the topical reportCENPD-387-P-A, Rev 000. The proposed 
change updates Technical Specification (TS) 6.9.1.11, ``Core Operating 
Limits Report (COLR),'' to include the topical report in the list of 
analytical methods used. Additionally, the Bases for TS 2.1.1, 
``Reactor Core,'' would be modified to reflect use of the improved heat 
flux correlation.
    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 proposed amendment would allow the implementation of ABB-NV 
critical heat flux correlation to St. Lucie Unit 2 core. The 
proposed changes have no adverse impact on the operation of the 
plant and have no relevance to the accident initiators. There are no 
changes to the plant configuration, and thus the frequency of 
occurrence of previously analyzed accidents is not affected by the 
proposed changes. With the application of the added methodology (the 
approved ABB-NV DNB correlation), the safety analysis would continue 
to remain consistent with the design basis requirements. The 
proposed changes, including changes to the TS Bases, have no adverse 
effect on the safety analysis and thus would not involve a 
significant increase in the consequences of design basis accidents. 
Changes to the COLR limits will continue to be controlled per 
Generic Letter 88-16 under the provisions of 10 CFR 50.59 and the 
requirements of TS 6.9.1.11.c.
    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) Use of the Modified Specification Would Not Create the Possibility 
of a New or Different Kind of Accident From Any Previously Evaluated

    The proposed amendment updates the list of approved methodology 
in TS 6.9.1.11 and makes corresponding changes to the TS Bases for 
TS 2.1.1. These changes would not create the possibility of a new 
kind of accident since there is no change to plant configuration, 
systems, or components, which would create new failure modes. The 
modes of operation of the plant would remain unchanged.
    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 accident previously evaluated.

(3) Use of the Modified Specification Would Not Involve a Significant 
Reduction in a Margin of Safety

    The proposed changes have no significant adverse impact on the 
safety analysis. As such, these changes would continue to provide 
margin to the acceptance criteria for Specified Acceptable Fuel 
Design Limits (SAFDL), 10 CFR 50.46(b) requirements, primary and 
secondary overpressurization, peak containment pressure, potential 
radioactive releases, and existing limiting conditions for 
operation. The future use of updated approved methodology will 
follow all design basis requirements to ensure that a safety margin 
to the acceptance criteria would continue to remain available at all 
power levels for operation of St. Lucie Unit 2.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant reduction in a 
margin of safety.

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

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

    Date of amendment request: March 29, 2001.
    Description of amendment request: The licensee proposed to amend 
the Technical Specifications (TSs) in three areas, adopting three NRC-
approved Technical Specification Task Force (TSTF) issues. This notice 
is concerned

[[Page 29359]]

with changes covered by one of the three issues, identified as TSTF-51.
    The licensee proposed to adopt TSTF-51, reducing the operability 
requirements for certain engineered safeguard features (ESFs) such as 
secondary containment, standby gas treatment, control room envelop 
filtration. The current requirements specify that these ESFs be 
operable during movement of irradiated fuel in the secondary 
containment, and during core operation. The proposed changes would 
specify these ESFs be operable during movement of recently irradiated 
fuel in the secondary containment, and would eliminate the 
applicability during core alteration. The associated licensee-
controlled TS Basis document would also be changed to reflect the 
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. The NRC staff has reviewed the licensee's analysis 
against the three standards of 10 CFR 50.92(c). The NRC staff's review 
is presented below:
    The first standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated. TSTF-51 involves no hardware design change, thus there will 
be no adverse effect on the functional performance of the ESFs to 
mitigate accident consequences. The ESFs are not initiators of any 
previously analyzed accidents, thus the proposed changes cannot 
increase the probability of any previously analyzed accidents. 
Therefore, the proposed changes will not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The second standard requires that operation of the unit in 
accordance with the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. TSTF-51 involves no hardware design change or procedural 
change; hence all components, systems, and structures will continue to 
perform as originally designed by the licensee and previously accepted 
by the NRC staff. Therefore, the proposed changes covered by TSTF-51 
will not create the possibility of a new or different kind of accident 
from any previously evaluated.
    The third standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
reduction in a margin of safety. Since TSTF-51 involves no change to 
the design, operational procedure, or analysis methodology, TSTF-51 
will not affect in any way the performance characteristics and original 
intended functions of any system, structure or component. Therefore, 
the proposed changes do not involve a significant reduction in a margin 
of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the part of the amendment request identified as TSTF-51 involves 
no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard P. Correia, Acting.

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

    Date of amendment request: March 29, 2001.
    Description of amendment request: The licensee proposed to amend 
the Technical Specifications (TSs) in three areas, adopting three NRC-
approved Technical Specification Task Force (TSTF) issues. This notice 
is concerned with one of the three changes, identified as TSTF-204.
    The licensee proposed to adopt TSTF-204, revising Limiting 
Condition for Operation (LCO) 3.8.5. Currently, LCO 3.8.5 requires that 
direct current (DC) power subsystems shall be OPERABLE to support the 
electrical power distribution subsystems required by LCO 3.8.9 
(pertaining to shutdown conditions). Adoption of TSTF-204 would change 
this to require either the Division 1 or Division 2 DC electrical power 
subsystems, in addition to the Division 3 DC electrical power 
subsystem, shall be OPERABLE. This change would restore the TS to what 
it was before the TS was converted to the Improved TS format by 
Amendment No. 91 (February 15, 2000). The associated licensee-
controlled TS Basis document would also be changed to reflect the 
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. The NRC staff has reviewed the licensee's analysis 
against the three standards of 10 CFR 50.92(c). The NRC staff's review 
is presented below:
    The first standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated. The proposed changes to adopt TSTF-204 involve no hardware 
design change or operational procedure change, thus there will be no 
adverse effect on the functional performance of any plant SSC; the 
decreased operability requirement pertains to times when there is less 
demand on the electrical subsystems (i.e., during shutdown conditions). 
All structures, systems and components (SSCs) will continue to perform 
their design functions with no decrease in their capabilities to 
mitigate the consequences of postulated accidents. Accordingly, the 
proposed operability requirements will lead to no significant increase 
in the consequences of an accident previously evaluated, and no 
increase of the probability of an accident previously evaluated.
    The second standard requires that operation of the unit in 
accordance with the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. TSTF-204 involves no hardware design change or procedural 
change; hence all SSCs will continue to perform as originally designed 
by the licensee and previously accepted by the NRC staff. Therefore, 
the proposed changes covered by TSTF-204 will not create the 
possibility of a new or different kind of accident from any previously 
evaluated.
    The third standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
reduction in a margin of safety. Since TSTF-204 involves no change to 
the design, operational procedure, or analysis methodology, TSTF-204 
will not affect in any way the performance characteristics and intended 
functions of any SSC. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the part of the amendment request identified as TSTF-204 involves 
no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard Correia, Acting.

[[Page 29360]]

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

    Date of amendment request: March 29, 2001.
    Description of amendment request: The licensee proposed to amend 
the Technical Specifications (TSs) in three areas, adopting three NRC-
approved Technical Specification Task Force (TSTF) issues. This notice 
is concerned with one of the three changes, identified as TSTF-287.
    The licensee proposed to adopt TSTF-287, adding to Section 3.7.2, 
Control Room Envelope Filtration System (CREFS), a note to permit the 
control room envelope be opened intermittently under administrative 
control, and a new Condition B allowing 24 hours to restore operability 
of the two CREFS subsystems if their operability is lost due to 
inoperable control room envelope boundary. These proposed provisions 
would allow time to diagnose, plan and possibly repair, and test most 
problems with the control room envelope boundary. The associated 
licensee-controlled TS Basis document would also be changed to reflect 
the TS 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. The NRC staff has reviewed the licensee's analysis 
against the three standards of 10 CFR 50.92(c). The NRC staff's review 
is presented below:
    The first standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated. The proposed change to adopt TSTF-287 involves no hardware 
design change or operational procedure change, thus there will be no 
adverse effect on the functional performance of any plant structures, 
systems or components (SSCs). The allowance to open the control room 
envelope intermittently does not increase accident consequences on 
control personnel since the administrative controls would rapidly 
restore integrity. Allowing 24 hours to restore the integrity of the 
control room envelope could result in an increase in consequences of a 
design-basis accident occurring during this time to control room 
personnel, but the administrative controls in place would easily and 
quickly reverse the condition, re-establishing control room envelope 
integrity, and thus limiting increases in consequences. Thus, all SSCs 
will continue to perform their design functions with no decrease in 
their capabilities to mitigate the consequences of postulated 
accidents. Accordingly, the proposed operability requirements will lead 
to no significant increase in the consequences of an accident 
previously evaluated, and no increase of the probability of an accident 
previously evaluated.
    The second standard requires that operation of the unit in 
accordance with the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. TSTF-287 involves no hardware design change or procedural 
change; hence it does not negatively affect the design or performance 
of any SSC, and all SSCs will continue to perform as originally 
designed by the licensee and previously accepted by the NRC staff. 
Therefore, the proposed changes will not create the possibility of a 
new or different kind of accident from any previously evaluated.
    The third standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
reduction in a margin of safety. Since TSTF-287 involves no change to 
the design, operational procedure, or analysis methodology, TSTF-287 
will not affect in any way the performance characteristics and intended 
functions of any SSC. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the part of the amendment request identified as TSTF-287 involves 
no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard Correia, Acting.

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

    Date of amendment request: May 2, 2001.
    Description of amendment request: The proposed amendment would (1) 
relocate requirements of the American Society of Mechanical Engineers 
(ASME) Boiler and Pressure Vessel Code (the Code), Section XI, 
inservice testing (IST) program currently contained in technical 
specification surveillance requirement (TSSR) 4.15.B to the TS 
Administrative Control Section 6.8, Programs and Manuals, (2) make 
conforming changes to several surveillance requirements to reflect the 
change in reference from TSSR 4.15.B to the licensee-controlled IST 
Program, (3) reword TSSRs 4.5.A.3 and 4.5.D.1 to be consistent with 
NUREG-1433, (4) incorporate TS Task Force (TSTF) initiative TSTF-279 
into TS Administrative Control Section 6.8, and (5) revise TSSRs 
4.6.H.1, 4.6.H.3, and Table 4.6.1 to change the inspection and 
functional testing interval extensions reference from plus-or-minus 25 
percent to plus 25 percent.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

1. The Proposed Amendment Will Not Involve a Significant Increase in 
the Probability or Consequences of an Accident Previously Evaluated

    The requested changes are administrative in nature in that they 
relocate IST [inservice testing] requirements from the Monticello TS 
[Technical Specifications] to a licensee controlled IST program, 
rewrite TS Surveillance Requirements 4.5.A.3 and 4.5.D.1 for 
clarification using the wording from NUREG-1433 and revise TS 
surveillance requirements for inspection and functional testing 
interval extensions. The requested changes will not revise previous 
commitments to 10 CFR 50.55a of [sic] ASME [American Society of 
Mechanical Engineers] Code [ASME Boiler and Pressure Vessel Code], 
Section XI, IST requirements.
    The proposed changes do not involve any change to the 
configuration or method of operation of any plant equipment that is 
used to mitigate the consequences of an accident, nor do they affect 
any assumptions or conditions in any of the accident analyses. Since 
the accident analyses remain bounding, their radiological 
consequences are not adversely affected.
    Therefore, the probability or consequences of an accident 
previously evaluated are not affected.

2. The Proposed Amendment Will Not Create the Possibility of a New or 
Different Kind of Accident From Any Accident Previously Analyzed

    The requested changes are administrative in nature in that they 
relocate IST requirements from the Monticello TS to the licensee 
controlled IST program, rewrite TS Surveillance Requirements 4.5.A.3 
and 4.5.D.1 for clarification using the wording from NUREG-1433 and 
revise TS surveillance requirements for inspection and functional 
testing interval extensions. The requested changes will not revise 
previous commitments to 10 CFR 50.55a or ASME Code, Section XI, IST 
requirements.
    The proposed changes do not involve changes to the configuration 
or method of

[[Page 29361]]

operation of any plant equipment that is used to mitigate the 
consequences of an accident, nor do they affect any assumptions or 
conditions in any of the accident analyses. Accordingly, no new 
failure modes have been defined for any plant system or component 
important to safety nor has any new limiting single failure been 
identified as a result of the proposed changes.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously analyzed is not created.

3. The Proposed Amendment Will Not Involve a Significant Reduction in 
the Margin of Safety

    The requested changes are administrative in nature in that they 
relocate IST requirements from the Monticello TS to the licensee 
controlled IST program, rewrite TS Surveillance Requirements 4.5.A.3 
and 4.5.D.1 for clarification using the wording from NUREG-1433 and 
revise TS surveillance requirements for inspection and functional 
testing interval extensions. The requested changes will not revise 
previous commitments to 10 CFR 50.55a or ASME Code, Section XI, IST 
requirements. Program requirements will remain to ensure that Code 
requirements are met.

    Therefore, a significant reduction in the margin of safety is not 
involved.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: April 11, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to relax the frequency for 
testing of excess flow check valves (EFCVs). Specifically, TS 
surveillance requirement 4.6.3.4 would be changed to revise required 
testing of EFCVs from once per 18 months for all valves to a test of a 
representative sample each 18 months such that all valves are tested 
once in 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. The NRC staff's review is presented below:
    (1) The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes affect the surveillance interval for the 
EFCV's, allowing a reduced number of valves to be tested at each 
interval. There are no physical plant modifications associated with 
this change. The EFCV's, which are installed on instrument lines 
penetrating containment, are designed to close in order to isolate 
containment upon a failure of the instrument line downstream of the 
valve. Since the EFCV's are designed to provide an accident mitigation 
function (i.e., minimize radiological effects due to an instrument line 
break), their postulated failure to close as a result of the proposed 
reduced testing frequency is not considered an initiator to any 
previously evaluated accidents. Therefore, there is no increase in the 
probability of occurrence of an accident as a result of the proposed 
changes.
    The design basis analyses for an instrument line break is evaluated 
in Section 15.6.2 of the Hope Creek Updated Final Safety Analysis 
Report (UFSAR). These analyses do not take credit for the closure of 
the EFCV's. The postulated failure of an EFCV to close as a result of 
the proposed reduced testing frequency is bounded by the existing UFSAR 
analyses. Therefore, the proposed changes do not involve a significant 
increase in the consequences of an accident previously evaluated.
    (2) The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes in TS surveillance requirements allow a 
reduced number of EFCV's to be tested each operating cycle. No other 
changes are being requested. The proposed changes do not introduce any 
new modes of plant operation and do not involve physical modifications 
to the plant. These changes will not alter any process variables, 
structures, systems, or components as described in the safety analyses. 
Therefore, the proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    (3) The proposed changes do not involve a significant reduction in 
a margin of safety.
    The radiological consequences of an unisolable break of an 
instrument line have previously been evaluated in Section 15.6.2 of the 
Hope Creek UFSAR. The accident analyses assume that the line break 
results in the release of reactor coolant into the Reactor Building 
until the reactor pressure vessel is depressurized. The analyses do not 
take credit for the closure of the EFCV's. The proposed reduced testing 
frequency only changes the potential for an undetected failure of an 
EFCV and does not change the event sequence upon which the current 
safety margin related to radiological consequences is based. Therefore, 
the proposed changes do not involve a significant reduction in a margin 
of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: April 18, 2001.
    Description of amendment request: South Carolina Electric & Gas 
Company (SCE&G) proposes a change to the Virgil C. Summer Nuclear 
Station (VCSNS) Technical Specifications (TS) Surveillance Requirements 
to revise the volumetric flow units for TS 4.7.6.c.1, c.3, e.1, e.3, 
and f to identify standard flow units expressed as standard cubic feet 
per minute. Volumetric flow units for TS 4.6.3.b.1, b.2, c.1, d, and g, 
and TS 4.9.11.b.1, b.3, d.1, e, and f are being revised to identify 
actual air flow units and are expressed as actual cubic feet per 
minute.
    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:

    South Carolina Electric & Gas Company (SCE&G) has evaluated the 
proposed changes to the VCSNS TS described above against the 
significant Hazards Criteria of 10 CFR 50.92 and has determined that 
the changes do not involve any significant hazard. The following is 
provided in support of this conclusion.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated? 
Changes associated with the identification of proper flow units are 
editorial and have no impact.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated? 
Changes associated with the identification of proper flow units are 
editorial and have no impact.
    3. Does this change involve a significant reduction in margin of 
safety? The margin of

[[Page 29362]]

safety for any of the ventilation systems associated with the 
proposed change is not compromised. Changes associated with the 
identification of proper flow units are editorial and have no 
impact.
    There are no significant safety hazards created by the change. 
There is no new or different accident postulated since the change is 
considered editorial. The design requirements of Regulatory Guide 
1.52 remain satisfied. Therefore, there is no significant decrease 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Thomas G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: Richard L. Emch, Jr.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant, 
Unit 1, Hamilton County, Tennessee

    Date of application for amendments: May 14, 2001 (TS 01-02).
    Brief description of amendments: The proposed amendment would 
change the Sequoyah (SQN) Unit 1 Operating License Condition 2.C.(9)(d) 
to clarify the lower voltage threshold for eddy current inspections.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:

A. The Proposed Amendment Does Not Involve a Significant Increase in 
the Probability or Consequences of an Accident Previously Evaluated

    The proposed change does not alter plant equipment, system 
design, or operating practices. The clarification of SQN's Unit 1 
[steam generator] SG inspection commitment provides a conservative 
inspection strategy that defines 1 volt as the lower threshold. The 
1-volt threshold is based on the subjectivity uncertainties 
associated with interpreting bobbin coil probe data to distinguish a 
dent below 1 volt. Given the current capability of eddy current 
technology, TVA's proposed change will define a reasonable criteria 
for tube inspection.
    TVA's proposed change continues to ensure that structural and 
leakage integrity of SQN's Unit 1 SG tubes is maintained. 
Accordingly, the proposed amendment does not result in any increase 
in the probability or consequences of an accident previously 
evaluated within the SQN Final Safety Analysis Report.

B. The Proposed Amendment Does Not Create the Possibility of a New or 
Different Kind of Accident From Any Accident Previously Evaluated

    SQN limits SG tube leakage between the primary coolant system 
and the secondary coolant system to 150 gallons per day per SG. This 
leakage limit ensures that tube cracks have an adequate margin of 
safety to withstand the loads imposed during normal operation and by 
postulated accidents. In addition, inservice inspections are 
performed in accordance with Regulatory Guide 1.83, Revision 1, 
``Inservice Inspection of Pressurized Water Reactor Steam Generator 
Tubes,'' to ensure that structural integrity of SG tubes is 
maintained during the plant operation cycle.
    The proposed change does not modify plant equipment, system 
design, or operating practices. The clarification of SQN's Unit 1 SG 
inspection commitment provides an inspection strategy that defines a 
minimum ``calling'' threshold for dent inspection. The 1-volt 
threshold is an inspection strategy based on the subjectivity 
associated with interpreting bobbin coil probe data below 1 volt for 
dented intersections. TVA's proposed change will continue to provide 
conservative inspection criteria that maintains structural and 
leakage integrity of SQN's Unit 1 SG tubes.
    Based on the above, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

C. The Proposed Amendment Does Not Involve a Significant Reduction in a 
Margin of Safety

    TVA's proposed clarification of the 1-volt threshold will 
continue to provide a conservative inspection criteria that will 
ensure that SG tube structural and leakage integrity is maintained. 
Accordingly, the margin of safety is not reduced.
    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Patrick M. Madden (Acting).

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: April 25, 2001.
    Brief description of amendments: The proposed changes would revise 
Technical Specification (TS) 3.8.1, ``AC [alternating current] 
Sources--Operating,'' to extend the allowable Completion Times for the 
Required Actions associated with restoration of an inoperable Emergency 
Diesel Generator (EDG) and an inoperable offsite circuit (i.e., startup 
transformer). In addition, the TS Surveillance Requirement (SR) 
corresponding to the 24-hour EDG endurance run in SR 3.8.1.14 would be 
revised to allow the SR to be performed during Modes 1 and 2. The 
proposed changes would also revise TS 3.8.9, ``Distribution Systems--
Operating,'' to extend the allowable Completion Times for the Required 
Actions associated with restoration of an inoperable AC electrical 
power distribution subsystem (i.e., 6.9 kilovolt (kV) AC safety bus).
    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 Technical Specification changes do not 
significantly increase the probability of occurrence of a previously 
evaluated accident because the 6.9 kV AC components (i.e., Emergency 
Diesel Generators (EDGs), startup transformers (STs), and safety-
related (Class 1E) busses) are not initiators of previously 
evaluated accidents involving a loss of offsite power. The proposed 
changes to the Technical Specification Action Completion Times do 
not affect any of the assumptions used in the deterministic or the 
Probabilistic Safety Assessment (PSA) analysis.
    The proposed Technical Specification changes will continue to 
ensure the 6.9 kV AC components perform their function when called 
upon. Extending the Technical Specification Completion Times to 14 
days and allowing the performance of the EDG 24-hour run test in 
either MODES 1 or 2 does not affect the design of the EDGs, the 
operational characteristics of the EDGs, the interfaces between the 
EDGs and other plant systems, the function, or the reliability of 
the EDGs. Thus, the EDGs will be capable of performing either 
accident mitigation function and there is no impact to the 
radiological consequences of any accident analysis.
    To fully evaluate the effect of the changes to the 6.9 kV AC 
components, Probabilistic Safety Analysis (PSA) methods and 
deterministic analysis were utilized. The results of this analysis 
show no significant increase in the Core Damage Frequency.
    The Configuration Risk Management Program (CRMP) in Technical 
Specification 5.5.18 is an administrative program that assesses risk 
based on plant status. Adding the requirement to implement the CRMP 
for Technical Specification 3.8.1 and 3.8.9 requires the 
consideration of other measures to mitigate consequences of an 
accident occurring while a 6.9 kV AC component is inoperable.

[[Page 29363]]

    The proposed changes do not alter the operation of any plant 
equipment assumed to function in response to an analyzed event or 
otherwise increase its failure probability. Therefore, these changes 
do not involve a significant increase in the probability or 
consequences of any 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.
    These proposed changes do not change the design, configuration, 
or method of operation of the plant. The proposed activities 
involves [sic] a change to the allowed plant mode for the 
performance of specific Technical Specification surveillance 
requirements. No physical or operational change to the 6.9 kV AC 
components or supporting systems are made by this activity. Since 
the proposed changes do not involve a change to the plant design or 
operation, no new system interactions are created by this change. 
The proposed Technical Specification changes do not produce any 
parameters or conditions that could contribute to the initiation of 
accidents different from those already evaluated in the Final Safety 
Analysis Report.
    The proposed changes only address the time allowed to restore 
the operability of the 6.9 kV AC components. Thus the proposed 
Technical Specification changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.

3. Do the Proposed Changes Involve a Significant Reduction in a Margin 
of Safety?

    Response: No.
    The proposed changes do not affect the Limiting Conditions for 
Operation or their Bases that are used in the deterministic analysis 
to establish any margin of safety. PSA evaluations were used to 
evaluate these changes, and these evaluations determined that the 
net changes are either risk neutral or risk beneficial. The proposed 
activities involves [sic] changes to certain Completion Times and to 
the allowed plant mode for the performance of specific Technical 
Specification Requirements. The proposed changes remain bounded by 
the existing Surveillance Requirement Completion Times and therefore 
have no impact to the margins of safety.
    The proposed change does [sic] not involve a change to the plant 
design or operation and thus does not affect the design of the 6.9 
kV AC components, the operation characteristics of the 6.9 kV AC 
components, the interfaces between the 6.9 kV AC components and 
other plant systems, or the function or reliability of the 6.9 kV AC 
components. Because 6.9 kV AC components performance and reliability 
will continue to be ensured by the proposed Technical Specification 
changes, the proposed changes do not result in a reduction in the 
margin of safety.
    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: April 23, 2001.
    Description of amendment request: The amendment would update the 
facility operating license (FOL) by deleting obsolete information, 
correcting errors, and making administrative changes to enhance the 
context and provide consistency.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

1. The Operation of Vermont Yankee Nuclear Power Station in Accordance 
With the Proposed Amendment Will Not Involve a Significant Increase in 
the Probability or Consequences of an Accident Previously Evaluated

    The proposed change makes editorial changes and brings the FOL 
up to date with the expectations of Massachusetts regulatory 
agencies. Since reactor operation under the proposed amendment is 
unchanged, no design or analytical acceptance criteria will be 
exceeded. As such, this change does not impact initiators of 
analyzed events or assumed mitigation of accident or transient 
events. The structural and functional integrity of plant systems is 
unaffected. Thus, there is no significant increase in the 
probability or consequences of accidents 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 affect any parameters or conditions 
that could contribute to the initiation of any accident. No new 
accident modes are created. No safety-related equipment or safety 
functions are altered as a result of these changes. Because it does 
not involve any change to the plant or the manner in which it is 
operated, the proposed change does 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 affect design margins or 
assumptions used in accident analyses, and has no effect on any 
assumed analysis initial condition. The capability of safety systems 
to function and limiting safety system settings are similarly 
unaffected as a result of this change. Thus, the proposed change 
will not involve a significant reduction in a margin of safety.

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

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)

[[Page 29364]]

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 Public 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: December 20, 2000, as 
supplemented March 14, 2001.
    The March 14, 2001, letter provided additional clarifying 
information which did not change the initial proposed no significant 
hazards consideration determination or expand the amendment beyond the 
scope of the original notice. A March 23, 2001, letter provided a 
camera-ready copy of the revised technical specification pages.
    Brief description of amendment: The amendment allows the expanded 
use of the Framatome Cogema Fuels M5 alloy for fuel rod cladding and 
fuel assembly spacer grids. A related Bases change is included with the 
licensee's application.
    Date of issuance: May 10, 2001.
    Effective date: As of the date of issuance and shall be implemented 
no later than the startup of Cycle 14 operation, approximately October 
1, 2001.
    Amendment No.: 233.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.Date of initial notice in Federal Register: 
February 6, 2001 (66 FR 9379).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 10, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 13, 2000.
    Brief description of amendments: The amendment revised the 
technical specifications to delete the ``Power Range Neutron Flux High 
Negative Rate,'' Trip Function from Reactor Trip System 
Instrumentation. The changes allow elimination of this unnecessary 
function and thereby reduces the potential for a transient. The changes 
are consistent with the Westinghouse Topical report previously accepted 
by the NRC.
    Date of issuance: May 17, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 114, 114, 120, 120.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 21, 2001 (66 
FR 11054).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 17, 2001.
    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: November 21, 2000, as 
supplemented April 25, April 26, May 3 (two letters), and May 8, 2001.
    Brief description of amendment: Amendment conforms the license to 
reflect the transfer of operating authority under Operating License No. 
DPR-20 to Nuclear Management Company, LLC, as approved by order of the 
Commission dated April 19, 2001 (66 FR 21021 dated April 26, 2001).
    Date of issuance: May 15, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 201.
    Facility Operating License No. DPR-20. Amendment revised the 
Operating License.
    Date of initial notice in Federal Register: December 19, 2000 (65 
FR 79431).
    The supplemental letters dated April 25, April 26, May 3 (two 
letters), and May 8, 2001, were within the scope of the initial 
application as originally noticed. The Commission's related evaluation 
of the amendment is contained in a Safety Evaluation dated April 19, 
2001.
    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: December 7, 2000.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TSs) in accordance with changes to the ``Standard 
Technical Specifications, Combustion Engineering Plants,'' NUREG 1432, 
Revision 1, made by the Nuclear Energy Institute Technical 
Specifications Task Force Change Number 258, Revision 4, addressing 
changes to various administrative controls in the TSs.
    Date of issuance: May 3, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 196.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7678).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 3, 2001.
    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: December 7, 2000.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TSs) in accordance with changes to the ``Standard 
Technical Specifications, Combustion Engineering Plants,'' NUREG 1432, 
Revision 1, made by the Nuclear Energy Institute Technical 
Specifications Task Force Change Number 287, Revision 5, addressing 
allowances for breach of the control room envelope. Also, the action 
table for TS Limiting Condition for Operation 3.7.10 is corrected by 
restoring Required Action D.2 (now renumbered to E.2), which was 
inadvertently omitted in Amendment No. 189, issued on November 30, 
1999.
    Date of issuance: May 3, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 197.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7678).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 3, 2001.
    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: December 7, 2000.

[[Page 29365]]

    Brief description of amendment: The amendment changes Technical 
Specification (TS) 3.5.2 in accordance with changes to the ``Standard 
Technical Specifications, Combustion Engineering Plants,'' NUREG 1432, 
Revision 1, made by the Nuclear Energy Institute Technical 
Specifications Task Force change number 325, Revision 0, addressing 
changes to the structure of the TS Limiting Condition for Operation 
(LCO) for the Emergency Core Cooling System.
    Date of issuance: May 3, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 198.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7675).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 3, 2001.
    No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of application for amendment: April 6, 2001, as supplemented 
by letter dated May 3, 2001.
    Brief description of amendment: The amendment revises the 
surveillance requirements pertaining to testing of the emergency diesel 
generators. The change removes the restrictions in plant technical 
specifications that prohibit performing the required testing during 
plant operation (Modes 1, 2, and 3). Additionally, the amendment 
modifies plant technical specifications to allow the endurance test to 
be performed in lieu of the load-run test provided the requirements of 
the load-run test, except the upper limit, are met.
    Date of issuance: May 18, 2001.
    Effective date: May 18, 2001, to be implemented within 30 days of 
the date of issuance.
    Amendment No.: 173.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 17, 2001 (66 FR 
19801).
    The May 3, 2001, supplemental letter provided clarifying 
information, did not expand the scope of the application as originally 
noticed and did not change the staff's original proposed no significant 
hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 18, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 5, 2001, as 
supplemented on April 13, 2001.
    Brief description of amendment: This amendment changes the Safety 
Limit Minimum Critical Power Ratio in Technical Specification (TS) 
2.1.2 from 1.08 to 1.06. The amendment makes administrative changes to 
TS 5.6.5, ``Core Operating Limits Report,'' section a and b. The 
amendment makes administrative changes to Bases section 2.1 to reflect 
this TS change and to Bases section 3.11 to reflect an earlier TS 
change.
    Date of issuance: May 8, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented prior to startup from Refueling Outage 13.
    Amendment No.: 191.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 7, 2001 (66 FR 
13802).
    The April 13, 2001, letter provided clarfying information that was 
within the scope of the amendment request and did not change the 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 8, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 9, 2000.
    Brief description of amendment: This amendment revised the 
technical specifications by approving thirteen of the simpler, generic 
administrative/editorial/consistency improvements agreed upon between 
the Nuclear Energy Institute (NEI) Technical Specification Task Force 
(TSTF) and the Nuclear Regulatory Commission (NRC), subsequent to the 
conversion of the PNPP Technical Specifications to the improved 
Standard Technical Specifications. The improvements include TSTF-5, 
TSTF-32, TSTF-38, TSTF-52, TSTF-65, TSTF-104, TSTF-106, TSTF-118, TSTF-
152, TSTF-166, TSTF-258, TSTF-278, and TSTF-279.
    Date of issuance: May 15, 2001
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 120
    Facility Operating License No.NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 13, 2000 (65 
FR 77920).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 15, 2001.
    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: March 12, 2001.
    Brief description of amendments: The amendments reduced the 
requirement for average reactor coolant temperature during the rod 
cluster control assembly drop test from greater than or equal to 
541 deg.F to greater than or equal to 500 deg.F.
    Date of issuance: May 7, 2001.
    Effective date: May 7, 2001.
    Amendment Nos: 214 and 208.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 4, 2001 (66 FR 
17967).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 7, 2001.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of application for amendment: June 9, 2000.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TS) 3.7.7 for the Main Turbine Bypass Valve 
surveillance test frequency, TS surveillance requirement SR 3.7.7.1 
frequency from 31 days to 92 days.
    Date of issuance: May 16, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 239.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 26, 2000 (65 FR 
46009).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 16, 2001.
    No significant hazards consideration comments received: No.

[[Page 29366]]

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

    Date of application for amendments: November 20, 2000, as 
supplemented February 6 and May 3, 2001.
    Brief description of amendments: These amendments incorporate 
changes to the Technical Specifications to increase the allowable 
deviation in individual rod position indication. By the February 6, 
2001, supplemental letter, the licensee withdrew portions of the 
original application that dealt with operation at greater than 85-
percent power. The licensee plans to submit those portions that deal 
with operation at greater than 85-percent power as a separate amendment 
request at a later time.
    Date of issuance: May 8, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 200 and 205.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 7, 2001 (66 FR 
9386).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 8, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 8, 2000, as supplemented 
by letter dated January 4, 2001.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) Section 3.5.5, ``Emergency Core Cooling Systems--
Seal Injection Flow,'' to replace the description of the seal injection 
flow with a description consistent with the method used to establish 
and verify reactor coolant pump seal injection flow limits and the 
method used to calculate the seal injection flow in the safety analyses 
for the Diablo Canyon Nuclear Power Plant.
    Date of issuance: May 7, 2001.
    Effective date: May 7, 2001, and shall be implemented within 30 
days from the date of issuance.
    Amendment Nos.: Unit 1--148; Unit 2--148.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 4, 2001 (66 FR 
17968).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 7, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 28, 2000.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) Section 5.5.2.b, ``Primary Coolant 
Sources Outside Containment'' by changing the system leak test 
frequency from ``at refueling cycle intervals or less'' to ``at least 
once every 18 months.'' The proposed change will also allow the 
provisions of Surveillance Requirement (SR) 3.0.2 to apply to TS 
Section 5.5.2.b.
    Date of issuance: May 11, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 119 and 97.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 20, 2000 (65 
FR 56955).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 11, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 5, 2000.
    Brief description of amendments: The amendments revise the licenses 
to reflect changes to the Updated Final Safety Analysis Report due to 
revisions to the dose equivalent iodine analysis.
    Date of issuance: May 11, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 120 and 98.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the license.
    Date of initial notice in Federal Register: December 13, 2000 (65 
FR 77925).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 11, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 6, 2000, as 
supplemented by letter dated February 9, 2001.
    Brief description of amendments: The amendments revised Technical 
Specifications (TS) 3.7.10, ``Control Room Emergency Filtration System 
(CREFS)--Both Units Operating,'' TS 3.7.11, ``Control Room Emergency 
Filtration System (CREFS)--One Unit Operating,'' and TS 3.7.13, 
``Piping Penetration Area Filtration and Exhaust System (PPAFES),'' to 
establish actions to be taken for inoperable ventilation systems due to 
a degraded control room pressure boundary or piping penetration area 
pressure boundary, respectively. Specifically, the changes allow the 
pressure boundaries of ventilation systems such as CREFS and PPAEFS to 
be opened intermittently under administrative control. A new condition 
is also added that allows 24 hours to restore inoperable CREFS and 
PPAFES pressure boundaries before requiring the units to perform an 
orderly shutdown. The applicable TS Bases have been revised to document 
these TS changes and to provide supporting information. These changes 
are based on Technical Specifications Task Force (TSTF) -287, Revision 
5, to the Standard Technical Specifications.
    Date of issuance: May 14, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 121 and 99.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 13, 2000 (65 
FR 77926).
    The supplemental letter dated February 9, 2001, provided clarifying 
information that did not change the scope of the November 6, 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 May 14, 2001.
    No significant hazards consideration comments received: No.

[[Page 29367]]

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant, 
Unit 1, Hamilton County, Tennessee

    Date of application for amendment: March 9, 2001.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) by allowing insertion of up to four lead test 
assemblies containing downblended uranium, in accordance with Framatome 
Cogema Fuels Topical Report BAW 2328, into the Sequoyah Unit 1 core for 
up to two fuel cycles.
    Date of issuance: May 9, 2001.
    Effective date: May 9, 2001.
    Amendment No.: 268.
    Facility Operating License No. DPR-77: Amendment revised the TSs.
    Date of initial notice in Federal Register: April 4, 2001 (66 FR 
17970).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 9, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 7, 2000.
    Brief Description of amendments: These amendments revise the 
Technical Specifications (TS) in Section 3.23 for the Main Control Room 
and Emergency Switchgear Room Ventilation and Air Conditioning Systems; 
TS Surveillance Requirement (SR) Section 4.20 for the Control Room Air 
Filtration System; and TS SR Section 4.12 for the Auxiliary Ventilation 
Exhaust Filter Trains. The proposed changes will revise the above SRs 
for the laboratory testing of the carbon samples for methyl iodide 
removal efficiency to be consistent with American Society for Testing 
and Materials Standard D3803-1989, ``Standard Test Method for Nuclear-
Grade Activated Carbon,'' with qualification, as the laboratory testing 
standard for both new and used charcoal adsorbent used in the 
ventilation system.
    Date of issuance: May 14, 2001.
    Effective date: May 14, 2001.
    Amendment Nos.: 225 and 225.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: March 21, 2001, (66 FR 
15931), supersedes March 20, 2000 (65 FR 15388).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 14, 2001.
    No significant hazards consideration comments received: No.

Yankee Atomic Electric Co., Docket No. 50-29, Yankee Nuclear Power 
Station (YNPS) Franklin County, Massachusetts

    Date of application for amendment: November 22, 2000.
    Brief description of amendment: The amendment relocated certain 
administrative requirements from the Yankee Nuclear Power Station 
(YNPS) Defueled Technical Specifications to the YNPS Decommissioning 
Quality Assurance Program. Additional editorial changes to titles and 
designations were also made.
    Date of issuance: May 15, 2001.
    Effective date: May 15, 2001.
    Amendment No.: 155.
    Facility Operating License No. DPR-3. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 4, 2001 (66 FR 
17972).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 15, 2001.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland this 22nd day of May 2001.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-13400 Filed 5-29-01; 8:45 am]
BILLING CODE 7590-01-P