Federal Register, Volume 65 Issue 202 (Wednesday, October 18, 2000)

[Federal Register Volume 65, Number 202 (Wednesday, October 18, 2000)]
[Notices]
[Pages 62380-62402]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-26645]

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NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from September 25, 2000, through October 6, 2000.
The last biweekly notice was published on October 4, 2000.

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.

[[Page 62381]]

Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC's Public Document Room, the Gelman Building, 2120 L
Street, NW, Washington, DC through September 22, 2000. The NRC is
relocating its Public Document Room to the NRC's headquarters building.
Effective September 26, 2000, documents may be examined at the NRC's
Public Document Room, located at One White Flint North, 11555 Rockville
Pike (first floor), Rockville, Maryland 20852. The filing of requests
for a hearing and petitions for leave to intervene is discussed below.
By November 17, 2000, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, 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: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, 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).

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

Date of amendment request: August 9, 2000.
Description of amendment request: The proposed amendment revises
Sections 6.5.3 and 6.5.4 of the Technical Specifications to eliminate
reference to the Independent Onsite Safety Review

[[Page 62382]]

Group (IOSRG) and to redefine the performance of the IOSRG function by
the nuclear quality assurance organization.
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 of occurrence or the consequences of an accident
previously evaluated. The proposed changes do not affect assumptions
contained in plant safety analyses, the physical design and/or
operation of the plant, nor do they affect Technical Specifications
that preserve safety analysis assumptions. None of the proposed
changes involve a physical modification to the plant, a new mode of
operation or a change to the UFSAR [Updated Final Safety Analysis
Report] transient analyses. No Technical Specification Limiting
Condition for Operation, Action Statement, or Surveillance
Requirement is affected by any of the proposed changes. The proposed
changes do not alter the design, function, or operation of any plant
component. Therefore, the proposed amendment does not affect the
probability of occurrence or consequences of an accident previously
evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any previously evaluated. The proposed changes
do not affect assumptions contained in the plant safety analyses,
the physical design and/or modes of plant operation defined in the
plant operating license, or Technical Specifications that preserve
safety analysis assumptions. The proposed changes do not introduce a
new mode of plant operation or surveillance requirement, nor involve
a physical modification to the plant. The proposed changes do not
alter the design, function, or operation of any plant components.
Therefore, the proposed amendment does not affect 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. None of the proposed changes involve a physical modification
to the plant, a new mode of operation or a change to the UFSAR
transient analyses. No Technical Specification Limiting Condition
for Operation, Action Statement, or Surveillance Requirement is
affected. Therefore, the proposed amendment does not reduce the
margin of safety.
Based upon the analysis provided herein [the licensee's August
9, 2000 application], the proposed changes will not increase the
probability or consequences of an accident previously evaluated,
create the possibility of a new or different kind of accident from
any accident previously evaluated, or involve a reduction in a
margin of safety. The performance of safety assessment and the IOSRG
functions by a single qualified organization will lead to
efficiencies in the performance of both functions. The training and
qualification of the personnel performing the IOSRG functions will
be unchanged from the current requirements. Therefore, the proposed
changes meet the requirements of 10 CFR 50.92(c) and involve no
significant hazards consideration.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Edward J. Cullen, Jr., Esq., PECO Energy
Company, 2301 Market Street, S23-1, Philadelphia, PA 19103.
NRC Section Chief: Marsha Gamberoni.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

Date of amendment request: August 9, 2000.
Description of amendment request: The proposed amendment revises
the Three Mile Island Nuclear Station, Unit 1 (TMI-1), Updated Final
Safety Analysis Report (UFSAR), Section 14.1.2.10, ``Steam Generator
Tube Failure Analysis,'' to include the dose resulting from the
postulated post-accident steam release through the main steam safety
valves. The revised dose for the TMI-1 steam generator tube failure
analysis would be increased above the values previously reviewed and
approved by the NRC, but would continue to be below the limits in Title
10 of the Code of Federal Regulations (10 CFR) Part 100. The proposed
change to the UFSAR modifies the existing analysis to account for
release of radioactivity to the atmosphere for the postulated tube
rupture analysis. The existing dose calculations do not account for
this release. Editorial and grammatical corrections are also made.
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 of occurrence or the consequences of an accident
previously evaluated. This change has no effect on structures,
systems or components prior to the postulated steam generator tube
failure accident or any other accident. The proposed change corrects
the existing UFSAR Steam Generator Tube Failure accident analysis to
account for the release to atmosphere through the main steam safety
valves (MSSVs). The resulting revised radiological consequences for
the postulated Steam Generator Tube Failure accident remain well
below the 10 CFR 100 limits.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any previously evaluated. This change has no
impact on any plant structures systems or components. The only
impact is the revised radiological consequences of the Steam
Generator Tube Failure accident analysis to account for the release
to atmosphere through the MSSVs. This change only corrects the
existing TMI Unit 1 UFSAR.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety. No change to any plant structure, system or component is
being made or proposed by this change. This change does not involve
any change to safety system setpoints for operation. The revised
radiological consequences of the Steam Generator Tube Failure
accident analysis remain well below 10 CFR 100 limits.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland

Date of amendments request: September 14, 2000.
Description of amendments request: The proposed amendment revises
the Unit 1 and Unit 2 heatup curves (Technical Specification Figures
3.4.3-1) and Unit 1 and Unit 2 cooldown curves (Technical Specification
Figures 3.4.3-2) to increase the allowable heatup and cooldown rates.
Use of stress intensity factor KIC, permitted by American
Society of Mechanical Engineers (ASME) Code Case N-640,

[[Page 62383]]

made it possible to increase the heatup and cooldown rates without
changing existing pressure-temperature (P-T) limits. The existing P-T
limits were approved previously. Application of Code Case N-640 to
generate P-T curves is not currently permitted by the regulations.
Therefore, pursuant to 10 CFR 50.12, a separate request for an
exemption to use Code Case N-640 was submitted in a letter dated
September 14, 2000.
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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
In accordance with 10 CFR Part 50, Appendix G, the Calvert
Cliffs pressure/temperature (P-T) limits for material fracture
toughness requirements of the reactor coolant pressure boundary
materials were developed using the methods of linear elastic
fracture mechanics and the guidance found in the American Society of
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section
III, Appendix G. The Calvert Cliffs P-T limits are based on fluence
level. The fluence levels are determined in the same manner as the
pressurized thermal shock (PTS) screening criteria defined in 10 CFR
50.61 for the critical elements. Methods described in the Nuclear
Regulatory Commission Regulatory Guide 1.99, Revision 2, are used to
predict the embrittlement effect of neutron irradiation on reactor
vessel materials. Regulatory Guide 1.99 defines embrittlement effect
in terms of adjusted reference temperatures, which depends on the
material property of the PTS critical elements.
The proposed higher heatup and cooldown rates for the Technical
Specification P-T limits were made possible by the ASME Code Case N-
640 which permits use of reference stress intensity factor
KIC, in place of KIA. Use of KIC,
for the maximum stress intensity factor that will not lead to
failure, is the correct value to use. Although conservative in terms
of developing P-T limits, use of KIA results in a very
restrictive heatup and cooldown rate that challenges plant safety.
To bound the existing LTOP [low-temperature overpressure protection]
enable temperatures, while increasing the heatup and cooldown rates,
the criteria described in ASME Section XI Code Case 514 is used.
Code Case 514 is listed in Regulatory Guide 1.147 as acceptable to
the Nuclear Regulatory Commission (NRC) for this application. With
the new higher heatup and cooldown rates, the underlying intent of
the 10 CFR Part 50, Appendix G, requirement for adequate margin to
prevent brittle failure of the reactor coolant pressure boundary
materials is maintained. Additionally, since the cooldown rates are
not changed above 300 deg. F, the safety analyses and dose
consequences in the Updated Final Safety Analysis Report are not
affected.
Therefore the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different [kind]
of accident previously evaluated.
The implementation of the proposed revision has no significant
effect on either the configuration of the plant, or the manner in
which it is operated.
Therefore, this proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
As discussed above, although conservative in terms of developing
P-T limits, use of KIA results in a very restrictive
heatup and cooldown rate that challenges plant safety. The
insignificant margin reduction in P-T limits is more than
compensated by the safety benefits that are realized in terms of
plant component integrity as a result of the higher heatup and
cooldown rates. With the proposed change, the underlying intent of
the 10 CFR Part 50, Appendix G, requirement for adequate margin to
prevent brittle failure of the reactor coolant pressure boundary
materials is maintained, and there is a net gain in overall plant
safety margin.
Therefore, this proposed change does not significantly reduce
the margin of safety.

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

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland

Date of amendments request: September 14, 2000.
Description of amendments request: The proposed amendment adds two
analytical methods to the list of approved core operating limits
analytical methods in the Technical Specifications (TSs) for Calvert
Cliffs, Unit Nos. 1 & 2. In a letter dated March 16, 2000, from Mr. S.
A. Richards, NRC to Mr. I. C. Rickard, ABB Combustion Engineering, the
Nuclear Regulatory Commission approved the Topical Report CENPD-387-P,
``ABB Critical Heat Flux Correlations for [pressurized-water reactor]
PWR Fuel'' for referencing in licensing applications for Asea Brown
Boveri, Inc. Combustion Engineering, Inc. (ABB-CE) plants.
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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed change allows the use of the ABB-NV and ABB-TV CHF
[critical heat flux] correlations in the thermal hydraulic analysis
for Calvert Cliffs Nuclear Power Plant. The ABB-NV is used for a
non-mixing vane fuel assembly and the ABB-TV correlations are used
for Turbo mixing vane fuel assembly. The CHF correlations determine
the departure from nucleate boiling ratio (DNBR). The specified
acceptable fuel design limit for DNBR will change for ABB-NV and
ABB-TV. The use of the ABB-NV and/or ABB-TV correlations with the
appropriate DNBR limit provides additional operating margin for
those analyses that presently use the CE-1 correlation.
The use of a different CHF correlation will not increase the
probability of an accident because the plant systems will not be
operated outside of design limits, the plant equipment will not be
operated in a different manner, and system interfaces will not
change.
As Turbo fuel is introduced to reactor, transistion cores will
exist in which Turbo mixing vane grid fuel assemblies are co-
residents with non-mixing vane grid fuel assemblies. The grid
hydraulic loss coefficient in the Turbo grids is greater than the
grid hydraulic loss coefficient for the non-mixing grids. The flow
diversion that will result does not increase the probability of an
accident previously evaluated because assembly flow has no impact on
accident initiators, and because plant systems will not be operated
outside of design limits, plant equipment will not be operated in a
different manner, and system interfaces will not change.
The change in the CHF correlation was the subject of Topical
Report CENPD-387-P-A, which was reviewed and approved by the NRC.
The use of a different CHF correlation will not increase the
consequences of an accident because Limiting Conditions [for]
Operation (LOCs) will continue to restrict operation to within the
regions that provide acceptable results, and Reactor Protection
System (RPS) trip setpoints will plant transients so that the
consequences of accidents will be acceptable.
The transition cores that will exist as Turbo fuel is introduced
to the reactor will not increase the consequences of an accident.
The TORC code accurately predicts the flow conditions in adjacent
fuel bundles that contain grids with different designs and
coefficients. The flow diversion will be compensated for by DNBR
margin gains. Operation within the LOCs and RPS setpoints will
continue to restrict plant

[[Page 62384]]

transients so that consequences of accidents will be acceptable.
Therefore, the proposed TS changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different [kind]
of accident from any accident previously evaluated.
The proposed change does not add any new equipment, modify any
interfaces with any existing equipment, alter the equipment's
function or change the method of operating the equipment. The
proposed change does not alter plant conditions in a manner that
could affect other plant components. The proposed change does not
cause any existing equipment to become an accident initiator. The
Turbo grid design does not introduce features that could initiate an
accident.
Therefore the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in the margin of
safety.
Safety Limits ensure that specified acceptable fuel design
limits are not exceeded during steady state operation, normal
operational transients, and anticipated operational occurrences. One
of the safety limits that accomplishes this is the DNBR limit. The
CHF correlations that have been approved for ABB-NV and ABB-TV
result in a DNBR limit that provides a 95% probability, at a 95%
confidence, that the hot fuel rod in the core will not experience
departure from nucleate boiling. The RPS in combination with the
LCOs, will continue to prevent any anticipated combination of
transient conditions for reactor coolant system temperature,
pressure and thermal power level that would result in a violation of
the Safety Limits.
Therefore the margin of safety is not significantly reduced by
this proposed change.

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

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland

Date of amendments request: September 14, 2000.
Description of amendments request: Calvert Cliffs Nuclear Power
Plant, Inc. (CCNPPI) proposed an amendment to incorporate changes
described below into the Technical Specifications (TSs) for Calvert
Cliffs Units 1 and 2.
On September 9, 1996, a final rule amending 10 CFR 50.55a was
issued requiring owners to implement, by September 9, 2001, the
requirements of the 1992 Edition through the 1992 Addenda of the
American Society of Mechanical Engineers Boiler and Pressure Vessel
Code (ASME Code) Section XI, Subsections IWE and IWL, as modified and
supplemented by 10 CFR 50.55a. CCNPPI has developed a program to effect
the implementation of Subsections IWE and IWL. This submittal requests
a license amendment in support of the program.
The TSs change replaces the reference to Regulatory Guide (RG) 1.35
with a reference to Section XI of the ASME Code, and deletes the
applicability of Surveillance Requirement 3.0.2. Compliance with RG
1.35 is not sufficient to comply with 10 CFR 50.55a, as amended, and
inspection frequencies will be in accordance with Subsection IWL of
Section XI; therefore, Surveillance Requirement 3.0.2 will no longer
apply.
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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The Containment Building is a passive safety structure that
prevents the release of radioactive materials to the environment in
post-accident conditions. The proposed Technical Specification
change updates requirements of the Technical Specifications that
have been made obsolete by the improvements of the Containment
[B]uilding inspections required by the changes in the regulations.
The improved inspections required by the American Society of
Mechanical Engineers [Boiler and Pressure Vessel] Code serve to
maintain containment response to accident conditions, by causing the
identification and repair of defects in Containment Buildings.
Relocating existing requirements, eliminating requirements that
duplicate regulations, and making administrative improvements
provide Technical Specifications that are easier to use. Because
existing requirements are controlled by regulation, there is no
reduction in commitment and adequate control is still maintained.
Therefore, the proposed change would not involve a significant
increase in probability or consequences of an accident previously
evaluated.
2. Would not create the possibility of a new or different [kind]
of accident from any accident previously evaluated.
The Containment Building is a passive safety structure designed
to contain radioactive materials released from the reactor coolant
system. The performance of the Containment Building is not evaluated
as the causal factor in any accident at Calvert Cliffs Nuclear Power
Plant. The proposed Technical Specification change updates
requirements of the Technical Specifications that were made obsolete
by the improvements of the Containment [B]uilding inspections
required by the changes in the regulations. Revising the Technical
Specifications, to comply with current regulations and to eliminate
duplication of requirements, does not create the possibility of a
new or different [kind] of accident from any previously evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The safety function of the Containment Building is to provide a
boundary to the release of radioactive material to the environment
during post-accident conditions. The change to the Technical
Specifications incorporate[s] improved inspection techniques and
criteria to ensure optimum containment integrity and, therefore,
optimum containment response in the event of an accident resulting
in a release of radioactive material from the reactor coolant
system. Optimizing containment integrity will result in maintaining
the margin of safety allowed by the Containment Buildings.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.

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

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland

Date of amendments request: September 15, 2000.
Description of amendments request: The proposed amendment revises
the Unit 1 and Unit 2 Technical Specification Surveillance Requirement
(SR) 3.1.7.2 which verifies that each control element assembly (CEA)
not fully inserted is capable of full insertion when tripped from at
least the 50 percent withdrawn position. Specifically, the proposed
amendment adds a note to SR 3.1.7.2, which allows the SR to not be
performed during initial power escalation following a refueling outage
if SR 3.1.4.6 (CEA drop time test) has been met. In addition, ``once''
was added to the SR frequency

[[Page 62385]]

as an administrative change to clarify that the SR is only performed
once and not on a periodic basis. This proposed license amendment is
consistent with Technical Specification Task Force (TSTF)-134, Revision
1, which received Nuclear Regulatory Commission (NRC) approval on April
21, 1998.
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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
A risk assessment was performed to support a prior license
amendment request submitted to change Surveillance Requirement (SR)
3.1.7.2 frequency from 24 hours to 7 days. Results of a study
performed in support of the risk assessment indicated no change in
the geometry of those components utilized in control element
assembly (CEA) insertion over the 7-day period. The study also
evaluated electronic/electrical failures that could cause a CEA to
be stuck, concluding that the feature that controls the movement of
the CEAs is not time-related. Since there have been no modifications
performed on the components analyzed or changes in the manner in
which they are operated, it is reasonable to assume that the
conclusions remain valid.
The CEA drop time test SR 3.1.4.6 proves that any work done
during the refueling outage does not prevent the rods from tripping.
Revising SR 3.1.7.2, such that it could allow more than seven days
from successfully performing the CEA drop time test does not change
this. However, as with any component, there will eventually be some
time-related degradation that may impact the ability of the CEAs to
drop. Thus, when the seven days are exceeded, there is some
negligible increase in the probability that a rod would fail to
drop. This causes an insignificant increase in core damage frequency
because it requires multiple rod failures to cause core damage in
the event of an overcooling event (the most bounding accident for a
stuck CEA during rod worth testing). This additional risk is
believed to be small since the degradation is the result of core
changes, which occur slowly, and not the result of maintenance. Thus
the risk increase due to this Technical Specification change is
considered to be negligible. The probability of an overcooling event
is not changed by the proposed change.
Therefore the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different [kind]
of accident from any accident previously evaluated.
The proposed change to the surveillance requirement for CEA
trippability does not result in any change to the facility or the
manner in which it is operated.
Therefore, this proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
Operation of the facility in accordance with this proposed
amendment does not involve a significant reduction in a margin of
safety. Control element assembly trippability is still demonstrated
via performance of SR 3.1.4.6. The risk increase due to this change
is considered to be negligible. Thus, appropriate equipment
continues to be tested in a manner and at a frequency necessary to
provide reasonable assurance that the equipment can perform its
assumed safety function.
Furthermore, this change is consistent with Technical
Specification Task Force (TSTF)-134, Revision 1, which has been
approved by the Nuclear Regulatory Commission. Adopting testing
practices consistent with those specified in TSTF-134, Revision 1
are acceptable based on similar design, like-component testing for
the system application and the availability of other Technical
Specification requirements which provide regular checks to ensure
limits are met.
Therefore, this proposed modification does not significantly
reduce the margin of safety.

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

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: August 1, 2000.
Description of amendment request: The proposed amendments would
provide revised spent fuel pool configurations, revised spent fuel pool
storage criteria, and revised fuel enrichment and burnup requirements
which take credit for soluble boron in maintaining acceptable margins
of subcriticality in the spent fuel storage pools. Also, the proposed
amendments would provide additional criteria for ensuring acceptable
levels of subcriticality in the spent fuel storage pools.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Will the change involve a significant increase in the
probability or consequence of an accident previously evaluated?
No, based upon the following:

Dropped Fuel Assembly

There is no significant increase in the probability of a fuel
assembly drop accident in the spent fuel pools when considering the
degradation of the or Boraflex panels in the spent fuel pool racks
coupled with the presence of soluble boron in the spent fuel pool
water for criticality control. The handling of the fuel assemblies
in the spent fuel pool has always been performed in borated water,
and the quantity of Boraflex remaining in the racks has no affect on
the probability of such a drop accident.
The criticality analysis showed that the consequences of a fuel
assembly drop accident in the spent fuel pools are not affected when
considering the degradation of the Boraflex in the spent fuel pool
racks and the presence of soluble boron.

Fuel Misloading

There is no significant increase in the probability of the
accidental misloading of spent fuel assemblies into the spent fuel
pool racks when considering the degradation of the Boraflex in the
spent fuel pool racks and the presence of soluble boron in the pool
water for criticality control. Fuel assembly placement and storage
will continue to be controlled pursuant to approved fuel handling
procedures to ensure compliance with the Technical Specification
requirements. These procedures will be revised as needed to comply
with the revised requirements which would be imposed by the proposed
Technical Specification changes. Note that the proposed amendment
would increase the number of different storage limits in Technical
Specification 3.7.15. However, these revised storage limits were
developed with input from station personnel. Their awareness, in
conjunction with any procedure changes as described above, will
provide additional assurance that an accidental misloading of a
spent fuel assembly will not occur.
There is no increase in the consequences of the accidental
misloading of spent fuel assemblies into the spent fuel pool racks
because criticality analyses demonstrate that the pool will remain
subcritical following an accidental misloading if the pool contains
an adequate soluble boron concentration. Current Technical
Specification 3.7.14 will ensure that an adequate spent fuel pool
boron concentration is maintained in the McGuire spent fuel storage
pools. A McGuire Station UFSAR change will revise Chapter 16,
``Selected Licensee Commitments'', to provide for adequate
monitoring of the remaining Boraflex in the spent fuel pool racks.
If that monitoring identifies further reductions in the Boraflex
panels which would not support the conclusions of the McGuire
Criticality Analysis, then the McGuire TS's and design bases would
be revised as needed to ensure that acceptable subcriticality are
maintained in the McGuire spent fuel storage pools.

[[Page 62386]]

Significant Change in Spent Fuel Pool Temperature

There is no significant increase in the probability of either
the loss of normal cooling to the spent fuel pool water or a
decrease in pool water temperature from a large emergency makeup
when considering the degradation of the Boraflex in the spent fuel
pool racks and the presence of soluble boron in the pool water for
subcriticality control since a high concentration of soluble boron
has always been maintained in the spent fuel pool water. Current
Technical Specification 3.7.14 will ensure that an adequate spent
fuel pool boron concentration is maintained in the McGuire spent
fuel storage pools.
A loss of normal cooling to the spent fuel pool water causes an
increase in the temperature of the water passing through the stored
fuel assemblies. This causes a decrease in water density that would
result in a decrease in reactivity when Boraflex neutron absorber
panels are present in the racks. However, since a reduction in the
amount of Boraflex present in the racks is considered, and the spent
fuel pool water has a high concentration of boron, a density
decrease causes a positive reactivity addition. However, the
additional negative reactivity provided by the current boron
concentration limit, above that provided by the concentration
required to maintain keff less than or equal to 0.95
(1470 ppm), will compensate for the increased reactivity which could
result from a loss of spent fuel pool cooling event. Because
adequate soluble boron will be maintained in the spent fuel pool
water, the consequences of a loss of normal cooling to the spent
fuel pool will not be increased. Current Technical Specification
3.7.14 will ensure that an adequate spent fuel pool boron
concentration is maintained in the McGuire spent fuel storage pools.
A decrease in pool water temperature from a large emergency
makeup causes an increase in water density that would result in an
increase in reactivity when Boraflex neutron absorber panels are
present in the racks. However, the additional negative reactivity
provided by the current boron concentration limit, above that
provided by the concentration required to maintain keff
less than or equal to 0.95 (1470 ppm), will compensate for the
increased reactivity which could result from a decrease in spent
fuel pool water temperature. Because adequate soluble boron will be
maintained in the spent fuel pool water, the consequences of a
decrease in pool water temperature will not be increased. Current
Technical Specification 3.7.14 will ensure that an adequate spent
fuel pool boron concentration is maintained in the McGuire spent
fuel storage pools.
2. Will the change create the possibility of a new or different
kind of accident from any previously evaluated?
No. Criticality accidents in the spent fuel pool are not new or
different types of accidents. They have been analyzed in Section
9.1.2.3 of the Updated Final Safety Analysis Report and in
Criticality Analysis reports associated with specific licensing
amendments for fuel enrichments up to 4.75 weight percent U-235.
Specific accidents considered and evaluated include fuel assembly
drop, accidental misloading of spent fuel assemblies into the spent
fuel pool racks, and significant changes in spent fuel pool water
temperature. The accident analysis in the Updated Final Safety
Analysis Report remains bounding.
The possibility for creating a new or different kind of accident
is not credible. The amendment proposes to take credit for the
soluble boron in the spent fuel pool water for reactivity control in
the spent fuel pool while maintaining the necessary margin of
safety. Because soluble boron has always been present in the spent
fuel pool, a dilution of the spent fuel pool soluble boron has
always been a possibility, however, a criticality accident resulting
from a dilution accident was not considered credible. For the
proposed amendment, the spent fuel pool dilution evaluation
(Attachment 7) demonstrates that a dilution of the boron
concentration in the spent fuel pool water which could increase the
rack keff to greater than 0.95 (constituting a reduction
of the required margin to criticality) is not a credible event. The
requirement to maintain boron concentration in the spent fuel pool
water for reactivity control will have no effect on normal pool
operations and maintenance. There are no changes in equipment design
or in plant configuration. This new requirement will not result in
the installation of any new equipment or modification of any
existing equipment. Therefore, the proposed amendment will not
result in the possibility of a new or different kind of accident.
3. Will the change involve a significant reduction in a margin
of safety?
No. The proposed Technical Specification changes and the
resulting spent fuel storage operating limits will provide adequate
safety margin to ensure that the stored fuel assembly array will
always remain subcritical. Those limits are based on a plant
specific criticality analysis (Attachment 6) based on the
``Westinghouse Spent Fuel Rack Criticality Analysis Methodology''
described in Reference 1. The Westinghouse methodology for taking
credit for soluble boron in the spent fuel pool has been reviewed
and approved by the NRC (Reference 6). This methodology takes
partial credit for soluble boron in the spent fuel pool and requires
conformance with the following NRC Acceptance criteria for
preventing criticality outside the reactor:
(1) keff shall be less than 1.0 if fully flooded with
unborated water which includes an allowance for uncertainties at a
95% probability, 95% confidence (95/95) level; and
( 2) keff shall be less than or equal to 0.95 if
fully flooded with borated water, which includes an allowance for
uncertainties at a 95/95 level.
The criticality analysis utilized credit for soluble boron to
ensure keff will be less than or equal to 0.95 under
normal circumstances, and storage configurations have been defined
using a 95/95 keff calculation to ensure that the spent
fuel rack keff will be less than 1.0 with no soluble
boron. Soluble boron credit is used to provide safety margin by
maintaining keff less than or equal to 0.95 including
uncertainties, tolerances and accident conditions in the presence of
spent fuel pool soluble boron. The loss of substantial amounts of
soluble boron from the spent fuel pool which could lead to exceeding
a keff of 0.95 has been evaluated (Attachment 7) and
shown to be not credible. Accordingly, the required margin to
criticality is not reduced.
The evaluations in Attachment 7, which show that the dilution of
the spent fuel pool boron concentration from the conservative
assumed initial boron concentration (2475 ppm) to the minimum boron
concentration required to maintain keff 0.95
(730 ppm) is not credible, combined with the 95/95 calculation which
shows that the spent fuel rack keff will remain less than
1.0 when flooded with unborated water, provide a level of safety
comparable to the conservative criticality analysis methodology
required by References 2, 3 and 4.
Therefore the proposed changes in this license amendment will
not result in a significant reduction in the facility's 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: Ms. Lisa F. Vaughn, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina 28201-1006.
NRC Section Chief: Richard L. Emch, Jr.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of amendment requests: September 26, 2000.
Description of amendment requests: The proposed amendments would
revise the current licensing basis in the Updated Final Safety Analysis
Report by requiring operator action to mitigate the effects of a loss
of seal injection (LOSI) cooling to the reactor coolant pumps (RCPs).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated?
The proposed change to the licensing basis recognizes that if
RCP Number 1 seal leak-off rates are low, continuous RCP operation
following a sustained LOSI may no longer be permitted. Tripping the
plant, securing the affected RCPs, and maintaining hot standby

[[Page 62387]]

conditions following a sustained LOSI will permit adequate RCP seal
cooling by readily achievable process controls. These actions ensure
that the probability of developing excessive seal leakage equivalent
to that of a previously evaluated loss of coolant accident (LOCA),
has not been significantly increased. Plant and RCP tripping are
anticipated transients that do not involve plant operation outside
design limits.
The consequences of large- and small-break (SB) LOCAs have been
evaluated and it has been shown that the radiological consequences
of these events do not result in unacceptable exposures to members
of the public. Therefore, even if stopping of the RCPs following a
LOSI and control of process parameters as described above does not
preclude RCP seal failures, the consequences of such failure are
bounded by the current accident analysis.
Therefore, the probability of occurrence or the consequences of
accidents previously evaluated are not significantly increased.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The leakage resulting from failed RCP seals may be large enough
to be considered a SBLOCA and industry data on SBLOCA initiating
frequencies includes the contribution from failed RCP seals. SBLOCAs
are a previously evaluated class of accidents. There is no new or
different kind of accident created as a result of this change.
Therefore, the change does 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?
The original design objective for the controlled leakage seal
assemblies in the RCPs was to permit sufficient controlled leakage
following a LOSI, such that cooling of the leakage provided by the
thermal barrier heat exchanger would be sufficient to continue RCP
operation unabated without challenging seal integrity. This is an
implied margin of safety for seal integrity, even if not explicitly
defined in the basis for any Technical Specification. It has been
postulated that the reduced seal leak-off will no longer permit
continuous RCP operation following a LOSI. The proposed change to
the licensing basis recognizes this condition and requires pump
tripping if seal injection cannot be restored prior to receiving
high temperature alarms in the leak-off return lines. Pump tripping
reduces the heat generated in the pump and permits readily
achievable process controls to maintain adequate seal cooling and an
adequate margin to seal failure.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: Claudia M. Craig.

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit 2, Berrien County, Michigan

Date of amendment request: September 30, 2000.
Description of amendment request: The proposed amendment would
allow an extension of the steam generator tube inspection surveillance
requirements of Technical Specification (T/S) Surveillance Requirement
4.4.5.3. The proposed amendment would prevent a mid-cycle shutdown to
meet the required 40-calendar month inspection interval of SR 4.4.5.3
and would allow the steam generator tube inspection to be performed
during the refueling outage following the current operating cycle.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated?
The accident considered applicable to the proposed change is a
steam generator tube rupture (SGTR). The precursors/initiators of a
SGTR (degraded, defective, or leaking tubes) are not known or
expected to be present in the CNP [Cook Nuclear Plant] Unit 2 steam
generators. These steam generators were newly installed in 1988, and
include corrosion prevention design features not included in
previous generations of steam generators.
There are no active degradation mechanisms present in the Unit 2
steam generators. Any tube imperfections that may be present or that
may be initiated during the current operating cycle are not expected
to progress to the point of tube failure before the next refueling
outage.
Considering the condition of the steam generators and the
operational time between inspections, the proposed change will not
significantly increase the probability of occurrence of an accident.
The proposed change will not affect the scope, methodology,
acceptance limit, or corrective measures of the existing steam
generator examination program.
Unit 2 recently completed an extended shutdown that effectively
limited the operational time that is the basis for the surveillance
frequency. When the reactor is shut down and the reactor coolant
system is at a reduced temperature, the steam generators are not
subject to conditions that lead to significant tube degradation.
Based on power operation time, the proposed extension will not
increase the operating interval between surveillances beyond that
currently allowed by [the] T/S.
The steam generator tube inspection interval is not used in the
SGTR accident analysis. The proposed change will, therefore, not
affect the accident analysis or methodology.
The severity of an analyzed tube rupture event is not related to
the time interval between inspections. The proposed change does not
affect allowable leakage rates or source terms, and does not change
the duration of an SGTR or the response to the event. Because the
severity of an accident is not increased by the proposed change,
there is no impact on offsite dose considerations.
Therefore, the probability of occurrence or the consequences of
accidents previously evaluated are not significantly increased.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change will not result in a change in plant
configuration or operation. Plant systems and components will not be
operated in a different manner because of this change. The proposed
change does not affect or create new accident initiators or
precursors.
Therefore, the proposed change does 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?
The T/S limit of one gallon per minute total steam generator
tube leakage ensures the offsite dose from tube leaks is limited to
a small fraction of 10 CFR 100 limits. The T/S leakage limit of 500
gallons per day in one steam generator is based on ensuring tube
integrity in the event of a steam line rupture or loss of coolant
accident. Because the offsite dose considerations from steam
generator tube failures are limited by the primary-to-secondary leak
rate program and not the tube inspection program, the proposed
change has no impact on offsite dose.
There are no tubes in service in any of the Unit 2 steam
generators that were found to be degraded, and no active steam
generator tube degradation is known to be occurring. Therefore, the
available margin in tube wall thickness is not being significantly
reduced. During the last inspection, 50% of the tubes were inspected
(more than sixteen times the T/S requirement), and none were found
to exceed the plugging limit, providing additional assurance that
safety margins are not being reduced. The absence of tube
degradation, along with the material and design features and
chemistry controls, provide reasonable assurance that tube repair
limits will not be approached during the current operating cycle.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.

[[Page 62388]]

amendment requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: Claudia M. Craig.

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

Date of amendment request: June 30, 2000, as supplemented September
22, 2000.
Description of amendment request: The proposed changes would modify
Sections 2.4.13.5, ``Design Bases for Subsurface Hydrostatic Loading''
2.5.4.6.1, ``Design Basis for Groundwater'' 3.4.1.2, ``Permanent
Dewatering System'' 3.8.1.6.4, ``Waterproofing Membrane'' 3.8.1.6.5,
``Steel Liner and Penetrations'' 9.3.3.1, ``Reactor Plant Vent and
Drain Systems, Design Bases'' 9.3.3.2.4, ``Reactor Plant Aerated Drains
System'' 9.3.3.2.4.1, ``Safety-Related Containment Recirculation
Cubicle Sump'' 9.3.3.3, ``Safety Evaluation'' 9.3.3.4, ``Tests and
Inspections'' and 12.3.1.3.2, ``Post-Accident Access to Vital Areas''
Tables 1.8-1, 3.2-1, 8.3-3, 12.3-3, and 12.3-4; and Figures 3.8-67 and
9.3-6 of the Final Safety Analysis Report (FSAR) to reflect the
addition of the new subsystem and its impact on other safety-related
systems. The new sump pump system creates the possibility of a
malfunction of a different type than previously evaluated in the FSAR
because of the system's dependence on electrical power; only one non-
environmentally qualified, non-safety-related pump is provided; and
portions of the Engineered Safety Feature Building structure are now
credited with preventing Recirculation Spray System (RSS) cubicle
flooding. Additionally, the proposed changes involve deviations from
safety classification and ``code & standards,'' Standard Review Plan
3.4.1 and Regulatory Guide (RG) 1.26.
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.
This license amendment request deals with changes in Millstone
Unit No. 3 Final Safety Analysis Report (FSAR) due to the
installation of a new sump pump system in the Engineered Safety
Features Building (ESFB). The sump pump system which prevents
inleakage through the containment basemat is not connected to and is
fully independent of the reactor coolant system. Therefore, the
proposed changes to this system will not increase the probability of
occurrence of a Loss of Coolant Accident (LOCA). The new system is a
support system for the Recirculation Spray System (RSS) and
containment protective boundary which are mitigation design
features. Therefore, the new system does not increase the
probability of occurrence of accidents previously evaluated.
The proposed changes to the groundwater sump system separate the
sump from the RSS pump cubicle. As such, the proposed changes would
preclude flooding of the RSS cubicles and a potential malfunction of
the RSS pumps. The RSS pumps function to provide containment and
core cooling, as early as 11 minutes and 30 minutes, respectively,
post LOCA. Operability of the RSS pumps is required long term. Since
the changes do not affect the operation of the RSS pumps, they will
not increase the consequences of a LOCA.
The new collection tank 3SRW-TK1 will be installed in the
location of the existing abandoned in place Chemical Addition Tank
(CAT) 3QSS*TK2, by the Refueling Water Storage Tank (RWST). The tank
will be seismically supported utilizing similar struts and
attachments to the RWST as the removed CAT. A calculation has
confirmed that there is no impact on the seismic qualification of
the RWST as a result of the new tank. The RWST provides water to the
Emergency Core Cooling System (ECCS) and Containment Quench Spray
(QSS) which are credited to mitigate the consequences of a LOCA.
Therefore, the proposed changes do not increase the consequences of
a LOCA.
In the proposed design, the installation of the new safety
related collection sump and casing pipe will result in a change in
the Supplemental Leak Collection and Release System (SLCRS) boundary
within the ESFB. This modification will be performed to meet the
SLCRS design requirements. Testing will be performed post
modification and routinely to satisfy SLCRS Technical Specification
3/4.6.6 requirements. Per Technical Specification 3/4.6.6 basis, the
SLCRS is credited post LOCA to limit the release of fission products
from the containment. Since the proposed changes do not affect
operability of SLCRS, it does not increase the consequences of a
LOCA.
In the proposed changes, sumps 3DAS*SUMP7A/B inflow pathways
will be restored such that it may become potentially contaminated.
Emergency operating procedures shall contain operator actions to
ensure that power to 3DAS*SUMP7A/B sump pumps 3DAS-P8A/B is isolated
post LOCA. As such, the proposed changes will continue to ensure
that potentially contaminated water is not discharged from
3DAS*SUMP7A/B. Therefore, the changes will not increase the
consequences of a LOCA.
The design change, per NUREG-0737 Section II.B.2 requirement,
has been evaluated by a calculation to ensure that the required
operator actions post LOCA can be performed within a 5 rem whole
body dose requirement, and has been found to be acceptable.
Therefore, these changes will not significantly increase the
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
This license amendment request is associated with the
installation of a new sump pump system in the ESFB. The current and
new sump pump systems are not accident initiators since neither
system is connected to, and both are fully independent of any system
that could cause an accident to occur. The new system, which
collects groundwater from beneath the Containment Structure and
ESFB, is a support system for the RSS and the containment protective
boundary, which are design basis accident mitigation design
features. Therefore, the proposed changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Therefore, the proposed changes do 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 Millstone Unit No. 3 FSAR changes reflect the installation
of a new sump pump system in the ESFB. The proposed changes do not
affect operation of the RWST, ECCS, QSS, RSS, SLCRS, Containment,
EDG or any Class 1E component required for safety. The additional
load on the Train A EDG and fuel oil consumption are within the
calculated allowance. Therefore, these changes do not significantly
reduce the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Section Chief: James W. Clifford.

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

Date of amendment request: July 28, 2000.
Description of amendment request: The proposed amendment would
revise the Fort Calhoun Station Unit 1 Technical Specifications to
allow installation of tube sleeves as an alternative to plugging to
repair defective steam generator tubes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards

[[Page 62389]]

consideration, which is presented below:

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

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

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

Date of amendment request: July 31, 2000.
Description of amendment request: The proposed changes will revise
LGS Technical Specifications (TSs) to replace the existing Automatic
Depressurization System (ADS) TS Surveillance Requirement (SR)
4.5.1.d.1, a 31-day channel functional test of the accumulator backup
compressed gas system low pressure alarm system, with a 31-day
verification of the ADS accumulator gas supply header pressure. The
existing TS SR 4.5.1.d.1 and SR 4.5.1.d.2.c, a 24-month channel
calibration of the accumulator backup compressed gas system low
pressure alarm system, will be relocated 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. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed TS changes have no physical impact on plant
equipment or the normal operation of plant systems. The ADS and the
ADS accumulator backup compressed gas system affected by the
proposed testing changes are normally in a standby mode and there
are no existing credible system failures that are accident
initiators. The ability of the ADS to depressurize the vessel
following a small break Loss of Coolant Accident (LOCA) so that flow
from low pressure Emergency Core Cooling Systems (ECCS) can enter
the core in time to limit fuel cladding temperatures is maintained
by the operability of the ADS accumulators and their inlet check
valves. The ADS accumulator backup compressed gas low pressure alarm
system has no impact on the ability of the ADS accumulators and
associated check valves to maintain an adequate gas supply required
to mitigate an accident. Therefore, the removal of the alarm system
testing from the TS has no impact on the ability of the ADS to cope
with the small break LOCA as previously evaluated. The replacement
of the monthly alarm channel functional test with the monthly
verification of the ADS accumulator gas supply header pressure will
assure that the ADS accumulators are pressurized as required to
support ADS operability and the ability of ADS to mitigate the
accident as previously analyzed is maintained. Therefore, the
proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes have no physical impact on plant equipment
or the normal operation of plant systems. The changes are limited to
changes in administrative testing requirements for the existing ADS
and ADS accumulator backup compressed gas low pressure alarm
systems, and the long term gas supply to the ADS valves. The changes
do not impact the methods of operation or manipulation of these
systems or components. The impact of these changes has been
evaluated to assure that the changes are in conformance with the
required design and licensing basis, and that system performance is
not degraded. The changes do not affect the operation of the ADS or
the ADS accumulator backup gas system and do not create any new
system failure modes or

[[Page 62390]]

accident initiators not previously considered. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed changes maintain the safety design basis of the ADS
and the ADS accumulator backup gas systems. The ADS accumulator
backup compressed gas low pressure alarm system does not support the
operability of the ADS accumulators which are required to maintain
an adequate gas supply for ADS vessel depressurization. Therefore,
the Channel Functional Test and Channel Calibration of backup gas
system alarms can be removed from the TS and have no impact on the
ability of the ADS to depressurize the reactor and maintain current
safety margins defined in the design basis for this TS. The
availability of the ADS accumulator backup gas system to perform its
long term cooling function after an accident or other event is not
addressed in any TS or Bases. The proposed changes in testing also
do not impact any of the Inservice Inspections or Tests currently
performed on the ADS or ADS accumulator backup gas system
components. Therefore, the proposed changes do not involve a
significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Section Chief: James W. Clifford.

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

Date of amendment request: February 29, 2000.
Description of amendment request: The amendment would incorporate
Supplement 3 to PL-NF-90-001, ``Application of Reactor Analysis Methods
for BWR Design and Analysis: Application Enhancements,'' into Technical
Specification Section 5.6.5, Core Operating Limits Report. The
supplement describes alternative methods for the analysis of the
rotated bundle event, the control rod withdrawal error event, and the
recirculation flow controller event.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
[The] proposed alternative analysis methods do not involve an
increase in the probability or consequences of an accident
previously evaluated. The alternative analysis methods affect the
analysis methods used to perform the Rotated Bundle Analysis, the
Rod Withdrawal Error Analysis, and the Recirculation Flow Controller
Failure Analysis. These events are analyzed on a cycle specific
basis to ensure that the operating limits contained in the COLR
[Core Operating Limits Report] will provide acceptable consequences
to the health and safety of the public consistent with NRC
guidelines. No physical changes are being made to plant systems,
structures or components. The alternative analysis methods ensure
that the [offsite] dose consequences of the postulated events remain
within the NRC guidelines.
Based on the above, it is concluded that the alternative
analysis methods do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The alternative analysis methods do not create the possibility
of a new or different kind of accident from any accident previously
evaluated. The proposed alternative analysis methods affect the
analysis methods for the Rotated Bundle, Rod Withdrawal Error and
Recirculation Flow Controller Failure Events. Since these
alternative analysis methods affect analytical methods and do not
affect any plant systems, structures, or components, it is concluded
that the proposed alternative analysis methods do not create the
possibility for any new or different kind of accident from any
accident previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The alternative analysis methods do not involve a significant
reduction in the margin of safety.
The Rotated Bundle Methodology is currently analyzed as a
moderate frequency event. The alternative methods will instead
analyze the Rotated Bundle Event as an Infrequent Event. Analysis of
this event as an infrequent event is consistent with NRC guidance
(provided in the Standard Review Plan) and the frequency
classification of the event as described in the SSES [Susquehanna
Steam Electric Station] FSAR [Final Safety Analysis Report]. The
proposed analysis methodology limits the analytical [offsite] dose
to a small fraction of 10 CFR 100 guidelines consistent with the NRC
guidelines. Therefore, the proposed alternative analysis methods do
not represent a significant reduction in the margin of safety.
The Rod Withdrawal Error Analysis currently does not credit the
Rod Block Monitor System to limit the extent of the inadvertent rod
withdrawal. The alternative proposed methods will allow credit in
the analysis for the Rod Block Monitor to limit the extent of the
inadvertent control rod withdrawal. Several plant and procedural
improvements have been implemented that have improved the
reliability of the Rod Block Monitor System. The analytical
acceptance criteria for the event is not affected. Therefore, the
proposed alternative analysis methods do not affect the margin of
safety.
The Recirculation Flow Controller Failure analysis is currently
analyzed using the RETRAN code. The proposed alternative analysis
methods use [PPL Susquehanna, LLC's] approved steady state nodal
simulation methodology instead of the RETRAN code. The [PPL
Susquehanna, LLC,] steady state nodal simulation methodology
produces final operating limits that are consistent with the RETRAN
methodology. The analytical acceptance criteria is not affected by
the alternative analysis methodology. Use of [PPL Susquehanna,
LLC's] methodology does not affect the margin of safety.

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

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

Date of amendment request: July 31, 2000.
Description of amendment request: The amendment would remove the
phrase ``maximum pathway'' from Surveillance Requirement 3.6.1.3.12 in
Technical Specification Section 3.6.1.3, ``Primary Containment
Isolation Valves.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change to eliminate the words ``maximum pathway''
does not affect any plant system or component. The change does not
impact operator performance or procedures. The leak rate testing of
the MSIVs [main steam isolation valves] will continue to be
performed in accordance with

[[Page 62391]]

10 CFR 50 Appendix J. The change does not impact the design basis
accident analyses presented in the FSAR [Final Safety Analysis
Report]. The change only affects how the as-found leakage is used to
evaluate operability and reportability. This change is consistent
with the guidance on leak rate testing presented in NEI 94-01
[Nuclear Energy Institute Guideline for Implementing Performance-
Based Option of 10 CFR Part 50, Appendix J] and the Standard
Technical Specifications. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
As discussed above, the proposed change to the Technical
Specifications does not affect any plant system or component and
does not affect plant operation. The consequences of accidents will
remain within the accident analysis described in the FSAR.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The proposed change does not affect any plant system or
component, and does not have any impact on plant operation. The
proposed change does not involve a significant reduction in the
margin of safety as currently defined in the bases of the applicable
Technical Specification section. Therefore, the proposed change does
not involve a significant reduction in the margin of safety.

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

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

Date of amendment requests: September 6, 2000 (PCN-274, Supplement
1). This application supersedes the licensee's application of November
24, 1999.
Description of amendment requests: The U.S. Nuclear Regulatory
Commission (the Commission) has granted the request of Southern
California Edison Company to withdraw its November 24, 1999,
application for proposed amendments. The Commission had previously
issued a Notice of Consideration of Issuance of Amendments published in
the Federal Register on December 29, 1999 (64 FR 73098). However, by
letter dated September 6, 2000, the licensee withdrew the proposed
change. TAC Nos. MA7289 and MA7290 used for the review of the November
24, 1999, application have been closed.
As submitted by the licensee on September 6, 2000, the proposed
amendments would modify the Technical Specifications (TSs) for the San
Onofre Nuclear Generating Station, Units 2 and 3, to revise TS 3.3.11,
``Post Accident Monitoring Instrumentation (PAMI).'' Specifically, the
proposed change would extend the PAMI channel calibration surveillance
frequency from 18 months to 24 months to accommodate a 24-month fuel
cycle for all PAMI instruments with the exception of the reactor
coolant system (RCS) temperature instrumentation. Surveillance
Requirement (SR) 3.3.11.4 relating to RCS temperature instrumentation
channel calibration every 18 months will remain in place.
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:

Do the proposed amendments:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No
The proposed license amendment to extend the calibration
surveillance frequency of Post Accident Monitoring Instrumentation
(PAMI) (excluding RCS temperature instrumentation) is being made to
support plant operation with a 24-month fuel cycle. Increasing the
calibration intervals for PAMI instrumentation to 30 months [24
months plus the 25 percent surveillance interval extension allowed
by SR 3.0.2] (excluding RCS temperature instrumentation) does not
affect the initiation or probability of any previously analyzed
accident. Increasing the calibration interval will not affect the
integrity of any of the principal barriers against radiation release
(fuel cladding, reactor vessel, and containment building). The
ability of the plant to mitigate the consequences of any previously
analyzed accidents is not adversely affected.
PAMI instrumentation provides to the operators both qualitative
and quantitative information used in accident mitigation and for the
safe shutdown of the plant. Instrumentation which provides
qualitative information is unaffected by a change in instrument
accuracy induced by drift due to the increased surveillance interval
because no explicit value is required by the Emergency Operating
Instructions (EOIs). Instrumentation that provides quantitative
information (i.e., decision points) in the EOIs have been evaluated.
This evaluation resulted in no changes to any operating
instructions. This evaluation of the proposed change to the
surveillance interval demonstrates that licensing basis safety
analyses acceptance criteria and San Onofre Nuclear Generating
Station (SONGS) Units 2 and 3 EOI criteria will continue to be met.
The proposed new surveillance frequency for these instrument
channels was evaluated using the guidance of Generic Letter 91-04
[``Changes in Technical Specification Surveillance Intervals To
Accommodate a 24-Month Fuel Cycle'']. The basis for the change
includes a quantitative evaluation of instrument drift for PAMI
instrumentation (excluding RCS temperature instrumentation)
providing quantitative information to the EOIs. Also, loop accuracy/
setpoint calculations for these instruments were updated to
accommodate the extended surveillance period. Analyses and
evaluations completed to assess the proposed increase in the
surveillance interval demonstrate that the effectiveness of these
instruments in fulfilling their respective functions is maintained.
Technical Specifications Channel Checks and Channel Functional
Checks for the subject channels, will continue to be performed to
provide assurance of instrument channel OPERABILITY.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of any previously
analyzed accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated?
Response: No
The increased calibration surveillance interval for PAMI
instrumentation (excluding RCS temperature instrumentation) is
justified based on evaluation of past equipment performance and does
not require any plant hardware changes or changes in normal system
operation. Changing the calibration interval for this
instrumentation has no means of creating the possibility of a new or
different kind of accident. There are no new decision points or
operator responses required to support existing accident mitigation
strategies.
Therefore, there are no new failure modes introduced as a result
of extending these surveillance intervals, and the proposed
amendment does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety?
Response: No
The proposed change to the calibration surveillance interval
(excluding RCS temperature instrumentation) was evaluated using the
criteria of 95% probability/95% confidence level for process sensor
drift.
PAMI instrumentation are used to provide indication following
certain hypothetical accident conditions and are used in EOIs for
trending and to initiate operator action at certain decision points.
Instrument uncertainty calculations have been updated for PAMI
instrumentation used for EOI

[[Page 62392]]

decision points as appropriate. Updated calculations show that the
total loop uncertainty for PAMI evaluated either decreased or
remained the same. These updated calculations demonstrate that
applicable accuracy requirements for SONGS 2 and 3 are satisfied
with the proposed new surveillance intervals.
Changing the calibration interval for these channels does not
affect the margin of safety for previously analyzed accidents.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.

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

Date of application for amendments: August 31, 2000 (TS 99-17).
Brief description of amendments: The proposed amendment would
revise the Sequoyah Nuclear Plant (SQN) Technical Specifications (TSs).
The revision would revise TS Section 5.6, ``Fuel Storage,'' to allow
credit for soluble boron in the fuel storage pools.
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 change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The presence of soluble boron in the spent fuel pool (SFP) water
for criticality control does not increase the probability of a fuel
assembly misplacement accident in the SFP. The handling of the fuel
assemblies in the SFP has always been performed in borated water.
The proposed change does allow greater flexibility for fuel storage
configurations in the SFP. The increased flexibility does not
introduce any greater complexity than the 3-zone configuration now
in use. Fuel assembly placement will continue to be controlled
pursuant to approved fuel handling procedures and will be in
accordance with the TS limitations. There is no increase in the
probability of a fuel placement accident.
The criticality analysis shows the consequences of the most
serious fuel assembly misplacement accident in the SFP are not
affected when considering the presence of soluble boron. Under
normal conditions, the rack keff [k effective] remains
subcritical as required by 10 CFR 50.68 [Section 50.68 of Title 10
of the Code of Federal Regulations], and is less than 0.95 with only
300 ppm [parts per million] soluble boron concentration. In the
event of a postulated fuel assembly misplacement, the presence of
sufficient soluble boron in the SFP precludes criticality as a
result of the misplacement. The criticality analysis demonstrates
that the pool keff will remain less than 0.95 following
an accidental misplacement due to 2000 parts per million (ppm) boron
concentration of the pool. In fact, concentration of only 700 ppm
soluble boron is sufficient to maintain keff less than
0.95 with 95% probability at 95% confidence level for the most
serious fuel assembly misplacement. The proposed TS will ensure that
an adequate SFP boron concentration is maintained. There is no
significant increase in the consequences of the accidental
misplacement of spent fuel assemblies in the SFP.
There is no increase in the probability of the loss of normal
cooling to the SFP water when considering the presence of soluble
boron in the pool water for subcriticality control since a high
concentration of soluble boron has always been maintained in the SFP
water.
Reactivity changes due to SFP temperature changes have been
evaluated. The base case criticality analysis used a SFP temperature
of 20 deg.C. The SFP reactivity uncertainty due to temperature
changes was considered for SFP temperatures ranging from 4 deg.C to
120 deg.C. The reactivity increment between 4 deg.C and 20 deg.C is
taken into account as additional uncertainty in the analysis. In all
spent fuel temperature cases, the temperature (and void)
coefficients of reactivity are negative. Therefore there is no
requirement for additional soluble boron above the base case level.
Because the coefficients of reactivity are negative, the
consequences of the loss of normal cooling to the SFP will not be
increased.
Therefore, based on the conclusions of the above analysis, the
proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
B. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Spent fuel handling accidents are not new or different types of
accidents and have been evaluated in the criticality analysis,
Reference 1.
The boron concentration in the SFP water is maintained at a
minimum of 2000 ppm. The proposed changes to the TS do not change
boron concentration requirements for the SFP water. A dilution of
the SFP soluble boron has always been a possibility; however, it was
shown in the SFP dilution evaluation (Reference 2) that there are no
credible dilution events for which the SFP keff could
reach criticality. Therefore, the implementation of proposed changes
to the TS will not result in the of a new kind of accident.
The proposed changes for re-rack storage management continue to
specify requirements for the spent fuel rack configurations. Since
the proposed SFP storage configuration limitations are comparable to
those used in the past, the new limitations will not have any
significant effect on normal SFP operations and maintenance and will
not create any possibility of a new or different kind of accident.
Verifications will continue to be performed to ensure that the SFP
loading configuration meets specified requirements.
The misplacement of a fuel assembly in the revised storage
configurations has been evaluated. In all cases, the rack
keff remains subcritical and less than 0.95 with 700 ppm
boron in the water.
As discussed above, the proposed changes will not create the
possibility of a new or different kind of accident. There is no
significant change in plant configuration, equipment design, or
equipment.
Under the proposed amendment, no changes are being made to the
racks themselves, any other systems, or to the physical fuel
handling structures in the Auxiliary Building itself. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
C. The proposed change does not involve a significant reduction
in a margin of safety.
The TS changes proposed by this License Amendment Request and
the resulting spent fuel storage configuration limitations will
provide adequate safety margin to ensure that the storage fuel
assembly array will always remain subcritical. Those limits are
based on a plant specific criticality analysis (Reference 1)
performed in accordance with accepted spent fuel rack criticality
analysis methodology.
While the criticality analysis utilized partial credit for
soluble boron, storage configurations have been defined to ensure
that the spent fuel rack keff will be less than 1.0 with
no soluble boron. Soluble boron credit is used to provide
subcritical margin such that the SFP keff is maintained
less than 0.95 under all credible conditions.
The loss of substantial amounts of soluble boron from the SFP,
which could lead to keff exceeding 0.95, has been
evaluated (Reference 2) and shown to be not credible. This
evaluation also shows that dilution of the SFP boron concentration
from 2000 ppm to 800 ppm is not credible. Also, the spent fuel
storage pool keff remains less than 1.0 at a 95/95
probability/confidence level with the pool filled with unborated
water. Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
Based on the above evaluation, TVA concludes that the proposed
changes to the TSs does [sic] not result in a significant reduction
in a margin of safety.

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,

[[Page 62393]]

400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.

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

Date of amendment request: September 14, 2000, as supplemented on
September 22, 2000.
Description of amendment request: This proposed change revises the
Technical Specification to clarify the valve isolation signal
information in Table 4.7.2 and makes an administrative change to the
table main steam isolation valves component identification to include
all the valves.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:

1. The operation of the Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
No changes are being made to plant design, method of operation
or method of testing. This change will not alter the basic operation
of process variables, systems, or components as described in the
safety analysis. No new equipment is introduced.
The proposed change does not affect the ability of the primary
containment isolation system or ECCS [emergency core cooling system]
systems to perform their required safety functions. The essential
safety functions of providing primary containment integrity and
providing water to cool the core in the event of an accident are
maintained. There is no physical or operational change being made
which would alter the sequence of events, plant response, or
conclusions of existing safety analyses. This proposed change
results in no impact on analyzed accident event precursors or
effects.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change does not involve any physical alteration of
plant equipment and does not change the method by which any safety-
related system performs its function. As such, no new or different
types of equipment will be installed, and the basic operation of
installed equipment is unchanged. There is no change in plant
operation that involves failure modes other than those previously
evaluated. The methods governing plant operation and testing remain
consistent with current safety analysis assumptions.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
No changes are being made to plant design, method of operation
or method of testing. This change will not alter the basic operation
of process variables, systems, or components as described in the
safety analysis. No new equipment is introduced.
The proposed change does not affect the ability of the primary
containment isolation system or ECCS systems to perform their
required safety functions. The essential safety functions of
providing primary containment integrity and providing water to cool
the core in the event of an accident are maintained. There is no
physical or operational change being made which would alter the
sequence of events, plant response, or conclusions of existing
safety analyses. This proposed change results in no impact on
analyzed accident event precursors or effects.
This proposed change does not alter the physical design of the
plant, methods or modes of operation, testing or analyses, thereby
resulting in no impact on safety functions. Since the proposed
change does not alter the means by which primary containment
isolation is maintained and containment cooling valves are isolated
in support of RHR [residual heat removal] LPCI [low pressure coolant
injection] actuation, there is no 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: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.

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

Date of amendment request: September 19, 2000.
Description of amendment request: This proposed change revises
Technical Specification (TSs) 3.5.H.3 and 3.5.H.4 related to low
pressure Emergency Core Cooling System (ECCS) injection/spray subsystem
operability during cold shutdown and refueling conditions. Two
circumstances are considered: (1) when no operations with the potential
for draining the reactor vessel (OPDRV) are in progress (addressed in
TS 3.5.H.3), and (2) when OPDRVs are in progress (addressed in TS
3.5.H.4). The proposed change provides completeness in the TS for the
defined conditions and also provides for the operation of an
alternative combination of low pressure ECCS injection/spray subsystems
to ensure adequate coolant inventory and sufficient heat removal
capability for the irradiated fuel during cold shutdown and refueling
conditions when OPDRVs are in progress.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:

[[Page 62394]]

During refueling and cold shutdown conditions with operations
having the potential for draining the reactor vessel (OPDRV) in
progress, any one ECCS injection/spray subsystems is adequate to
reflood the reactor vessel in the event of an inadvertent draindown.
Since the proposed change provides an equivalent means for achieving
this safety function, there is no reduction in reflood capability.
The additional flexibility, to maintain a combination of one core
spray subsystem and one LPCI [low pressure coolant injection]
subsystem (provided by this change), is equivalent to the safety
margin provided by the existing TS since a single active failure
affecting one subsystem results in the same remaining capability of
one ECCS subsystem.
Since the changed TS provides equivalent low pressure ECCS
injection/spray capability and protection from loss of coolant
inventory, the risk of an inadvertent draindown event is unchanged,
thus preserving previously existing margins of safety.
For circumstances involving no OPDRVs during refueling and cold
shutdown conditions, no ECCS or containment cooling equipment is
required to meet safety objectives. Thus, the margins of safety for
such situations are maintained.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.

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

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

Date of amendment request: September 26, 2000.
Description of amendment request: This proposed change revises
Technical Specification (TS) requirements regarding secondary
containment systems, including the Standby Gas Treatment System
(SBGTS). The affected TS sections are 1.0, Definitions; 3/4.7.B,
Standby Gas Treatment System; and 3/4.7.C, Secondary Containment
System. In addition, a new TS section, 3/4.7.E, Reactor Building
Automatic Ventilation System Isolation Valves (RBAVSIVs), is proposed.
Some of the proposed changes are administrative in nature and do not
affect the technical aspects of the requirements. Associated changes to
the TS Bases are also being made to conform to the changed TS. The
proposed changes provide certain additional flexibility in operations
when equipment is made or found to be inoperable, while also ensuring
appropriate actions are taken to place the plant in a safe condition
under such conditions.
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.
No changes are being made to the plant design, physical system
configuration, or basic method of operation as a result of the
proposed amendment. The Standby Gas Treatment System (SBGTS) and
secondary containment are not assumed to be initiators of any
analyzed event. The circumstances for which operability of SBGTS and
secondary containment are required are unchanged, and would not
occur at any greater frequency as a result of this change.
Therefore, the probability of a design basis loss-of-coolant
accident or fuel handling accident (the applicable accidents)
previously evaluated is not increased.
The proposed change does not increase the consequences of an
accident because system operability requirements are being
maintained. In lieu of suspending refueling activities when one
train of SBGTS is inoperable beyond seven days, placing the operable
train of SBGTS in operation ensures that no failures that could
prevent automatic actuation have occurred and that any other failure
would be readily detected. Operation of one train of the SBGTS is
sufficient to mitigate the consequences of any analyzed event. The
secondary containment systems assumed to operate following a design
basis accident continue to function as assumed in accident analyses
to mitigate the consequences of postulated accidents.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change does not involve any physical alteration to
the plant structures, systems, or components (SSCs), or the basic
manner in which these SSCs are operated or maintained. The methods
by which these systems perform their safety function are unchanged
and remain consistent with current safety analysis assumptions.
There is no change in plant operation that involves failure modes
other than those previously evaluated. Therefore, the proposed
change will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. 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 result in a significant reduction
in a margin of safety because restrictions placed on operations
which have the potential for releasing radioactive material to the
secondary containment continue to be in accordance with the
assumptions and conditions of existing safety analyses. Operations
with inoperable equipment have the proper restrictions to maintain
existing margins or to place the plant in a safe condition such that
inoperable equipment is not required to meet safety analysis
assumptions. Ensuring operability of one train of SBGTS together
with required secondary containment integrity is sufficient to
mitigate the consequences of any analyzed event. Since current
analyses are unaffected in this regard, margins of safety are
maintained.

Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.

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

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

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412,
Beaver Valley Power Station, Unit 2, Shippingport, Pennsylvania

Date of amendment request: September 1, 2000.

[[Page 62395]]

Description of amendment request: The proposed amendment would
revise certain 18-month surveillance requirements in the technical
specifications by eliminating the condition that testing be conducted
during shutdown, or during cold shutdown or refueling mode. The systems
that would be affected are the emergency core cooling system,
containment depressurization and cooling system, chemical addition
system, and containment isolation valve system.
Date of publication of individual notice in Federal Register:
September 12, 2000 (65 FR 55056).
Expiration date of individual notice: October 12, 2000.

Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of application for amendment: April 27, 2000.
Brief description of amendment request: The amendment seeks to
extend the applicability of the current pressure-temperature and
overpressure protection system limit curves from 13.3 effective full-
power years (EFPY) to 16.2 EFPYS.
Date of publication of individual notice in Federal Register:
August 29, 2000 (65 FR 52451).
Expiration date of individual notice: September 28, 2000.

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

Date of amendment request: August 29, 2000, as supplemented by
letter dated September 8, 2000.
Description of amendment request: The amendment proposes to change
Technical Specifications 3.0.D and 4.0.D to be equivalent to the
Boiling-Water Reactor NUREG-1433 guidance for the Improved Technical
Specifications limiting condition for operation 3.0.4, which is
currently under review.
Date of publication of individual notice in Federal Register:
September 14, 2000 (65 FR 55650).
Expiration date of individual notice: October 16, 2000.

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

Date of amendment request: September 14, 2000.
Brief description of amendment request: The amendment would clarify
the valve isolation signal information in the Technical Specification
Table 4.7.2 and make an administrative change to the table main steam
isolation valves component identification.
Date of publication of individual notice in Federal Register:
September 27, 2000 (65 FR 68111).
Expiration date of individual notice: October 27, 2000.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

Date of application for amendment: June 19, 2000, as supplemented
August 8, 2000.
Brief description of amendment: The amendment allows some emergency
diesel generator Technical Specification surveillance requirements to
be performed during plant operation instead of during plant shutdown.
Date of issuance: October 2, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 132.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 26, 2000 (65 FR
46006). The supplemental information did not change the application or
affect the proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 2, 2000.
No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

Date of application for amendment: April 1, 1999, as supplemented
June 14, and July 27, 2000.
Brief description of amendment: The amendment revised the TMI-1
Technical Specifications (TSs) 1.4.2, 1.4.3, 1.4.4, 3.3.1.2.b,
3.3.1.3.b, and c, 3.3.2.1, Table 4.1-1 (Items 14, 25, 31, and 32),
Table 4.1-3 (Items 4 and 6), Table 4.1-5, and TSs 4.1.5, 4.5.2.1.a and
b, 4.5.2.3.a, and 4.5.3.1.b.1 and 2, to: add limiting condition for
operation (LCO) action statements and make LCOs and surveillance
requirements more consistent with the revised ``Standard Technical
Specifications for Babcock & Wilcox Plants,'' (NUREG-1430, Revision 1);
correct conflicts or inconsistencies; and revise spent fuel pool
sampling frequency from monthly and after adding chemicals, to weekly.
TS 3.3.1.2.d is deleted as a result of the LCO additions described
above. Also, a Bases statement for surveillance testing was added to
Section 4.1 of the TSs and a revised Bases to Section 4.4.4 is included
as well.
Date of issuance: September 25, 2000.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 225.

[[Page 62396]]

Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 28, 1999 (64 FR
40906) and August 23, 2000 (65 FR 51349).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 25, 2000.
No significant hazards consideration comments received: No.

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

Date of application for amendments: December 1, 1999 (102-04378).
Brief description of amendments: The amendments to the operating
licenses delete or update outdated administrative information and
delete license conditions that are no longer applicable or have been
completed.
Date of issuance: September 29, 2000.
Effective date: September 29, 2000, to be implemented within 30
days of the date of issuance.
Amendment Nos.: Unit 1--128, Unit 2--128, Unit 3--128.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Operating Licenses.
Date of initial notice in Federal Register: March 8, 2000 (65 FR
12288).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 29, 2000.
No significant hazards consideration comments received: No.

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

Date of application for amendments: June 6, 2000, as supplemented
June 29 and July 3, 2000.
Brief description of amendments: The amendments restrict the
emergency diesel generator (DG) acceptance criteria for steady-state
voltage and frequency in several surveillance requirements (SRs)
involving DG starts in Technical Specification (TS) 3.8.1, ``AC
Sources--Operating,'' of the TSs for the three units. The amendments
also add a note to each SR that states ``The steady state voltage and
frequency limits are analyzed values and have not been adjusted for
instrument error.'' The restricted acceptance criterion is to ensure
proper DG operation.
Date of issuance: October 4, 2000.
Effective date: October 4, 2000, to be implemented within 45 days
of the date of issuance. For surveillance requirements associated with
the revised steady-state voltage and frequency limits in Technical
Specifications 3.8.1 and 3.8.2, the first performance is due at the end
of the first surveillance interval that began on the date the
surveillance was last performed prior to the date of implementation of
the amendments.
Amendment Nos.: Unit 1-129, Unit 2-129, Unit 3--129.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 12, 2000 (65 FR
43043).
The June 29 and July 3, 2000, supplements provided clarifying
information that was within the scope of the application and the
Federal Register notice, and did not change the staff's initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 4, 2000.
No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Date of application for amendments: February 18, 2000.
Brief description of amendments: The amendments remove the
anticipatory reactor scram signal for turbine electro-hydraulic control
(EHC) low oil pressure trip from the reactor protection system (RPS)
trip function.
Date of issuance: September 27, 2000.
Effective date: Immediately, to be implemented within 90 days.
Amendment Nos.: 181 and 176.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 5, 2000 (65 FR
17910).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 27, 2000.
No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois

Date of application for amendments: April 25, 2000.
Brief description of amendments: The amendments revised Technical
Specification 3/4.9.5, ``Communications'' to allow the movement of a
control rod in a fueled core cell in Operational Condition 5 to be
exempt from the requirement that direct communication be maintained
between the control room and the refueling platform personnel when the
rod is moved with its normal drive system.
Date of issuance: October 5, 2000.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 141 and 127.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 14, 2000 (65 FR
37422).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 5, 2000.
No significant hazards consideration comments received: No.

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

Date of application for amendment: June 27, 2000, as supplemented
August 18 and 30, 2000.
Brief description of amendment: The amendment changes Improved
Technical Specification Sections 3.5.1, ``Safety Injection Tanks
(SITs),'' and 3.5.2, ``ECCS [Emergency Core Cooling System]--
Operating,'' regarding completion times for restoring an inoperable SIT
and for restoring a low-pressure safety injection train.
Date of issuance: October 2, 2000.
Effective date: As of the date of issuance and shall be implemented
on or before December 31, 2000.
Amendment No.: 191.
Facility Operating License No. DPR-20: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 26, 2000 (65 FR
46007) (two notices).
The August 18 and 30, 2000, supplemental letters provided
clarifying information that was within the scope of the original
application and did not change the staff's initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 2, 2000.
No significant hazards consideration comments received: No.

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

Date of application for amendments: June 29, 2000, as supplemented
by

[[Page 62397]]

letters dated July 27, and August 10, 2000.
Brief description of amendments: The amendments revised the
Technical Specifications (TS) to reference the Westinghouse Best
Estimate Large Break Loss-of-Coolant Accident analysis methodology
described in WCAP-12945-P-A, March 1998. These amendments also address
corresponding TS Bases changes.
Date of issuance: October 2, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 188 and 181.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 23, 2000 (65 FR
51350).
The letter dated August 10, 2000, provided additional information
that did not change the scope of the application and the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 2, 2000.
No significant hazards consideration comments received: No.

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

Date of application for amendments: May 25, 2000, as supplemented
by letters dated July 31, August 8, and August 17, 2000.
Brief description of amendments: The amendments temporarily revise
TS 3.5.2, ``Emergency Core Cooling System;'' TS 3.6.6, ``Containment
Spray System;'' TS 3.6.17, ``Containment Valve Injection Water
System;'' TS 3.7.5, ``Auxiliary Feedwater System;'' TS 3.7.7,
``Component Cooling Water System;'' TS 3.7.8, ``Nuclear Service Water
System;'' TS 3.7.10, ``Control Room Area Ventilation System;'' TS
3.7.12, ``Auxiliary Building Filtered Ventilation Exhaust System;'' and
TS 3.8.1, ``AC Sources''.
Date of issuance: October 4, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 189 and 182.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 25, 2000 (65 FR
51860).
The supplements dated July 31, August 8, and August 17, 2000,
provided clarifying information that did not change the scope of the
May 25, 2000, application and the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 4, 2000.
No significant hazards consideration comments received: No.

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

Date of application of amendments: September 7, 2000.
Brief description of amendments: The amendments revise Surveillance
Requirement 3.8.1.9.a by adding a note stating that the upper limits on
frequency and voltage are not required to be met for the annual test of
the Keowee Hydro Units until the NRC issues an amendment that removes
the note in response to an amendment request to be submitted no later
than April 5, 2001.
Date of Issuance: October 4, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 316, 316, & 316.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes (65 FR 56600 dated September 19, 2000). That notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by October 19, 2000, but indicated that if the Commission makes
a final no significant hazards consideration.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, and a final no significant hazards consideration
determination are contained in a Safety Evaluation dated October 4,
2000.

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

Date of amendment request: May 8, 2000, as supplemented by letter
dated August 30, 2000.
Brief description of amendment: The amendment revises Technical
Specifications to remove the fuel building (FB) and the FB ventilation
system from the requirements associated with secondary containment
during power operation (except during movement of recently irradiated
fuel assemblies in the FB).
Date of issuance: September 22, 2000.
Effective date: As of the date of issuance and shall be implemented
30 days from the date of issuance.
Amendment No.: 113.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 14, 2000 (65 FR
37424).
The August 30, 2000, supplemental letter provided additional
information to support staff review of the original application, and
did not affect the initial finding of no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 22, 2000.
No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
Nos. 1 and 2 (ANO-1 and ANO-2), Pope County, Arkansas

Date of application for amendments: September 17, 1999, as
supplemented by letters dated June 29, August 3, and September 15,
2000.
Brief description of amendment: The amendments change heavy load
handling requirements and transportation provisions that would permit
the movement of the original and replacement steam generators (SGs)
through the ANO-2 containment construction opening during the SG
replacement outage.
Date of issuance: September 25, 2000.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: 209 & 221.
Facility Operating License Nos. DRP-51 and NPF-6: The amendments
revise the licenses.
Date of initial notice in Federal Register: February 23, 2000 (65
FR 9004).
The additional information provided in the June 29 and August 3,
2000, supplemental letters was noticed in the Federal Register on
August 23, 2000 (65 FR 51352). The September 15, 2000, supplement
provided clarifying information that was within the scope of the
Federal Register notice published August 23, 2000, and did not change
the staff's initial no significant hazards consideration determination.

[[Page 62398]]

The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 25, 2000.
No significant hazards consideration comments received: No.

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

Date of application for amendment: November 29, 1999, as
supplemented by letters dated January 26, May 17 (2 letters), May 31,
and August 4, 2000.
Brief description of amendment: The amendment revised the License
and Technical Specifications (TSs), and corresponding Bases have been
changed to maintain consistency with the transient and accident
analyses which evaluated the impact of the replacement steam generators
(SGs) that are being used for Cycle 15 operation. The License was
revised to incorporate a new methodology employed in calculating
radiological doses for some non-loss-of-coolant accident events. TS
changes were made to the Reactor Protection System (RPS) and Engineered
Safety Features Actuation System (ESFAS) low pressurizer pressure
setpoints, the RPS and ESFAS low SG pressure setpoints, the RPS and
ESFAS low SG level setpoints, the reactor coolant flow rate limit, and
the high linear power trip setpoints with inoperable main steam safety
valves (MSSVs). The amendment also made changes to the TSs and
corresponding Bases have been changed that are not directly related to
the replacement SGs. These changes revised the allowed outage time of
the MSSVs in Modes 1 and 2 to allow up to 12 hours to reduce the high
linear power level-high trip setpoint when one or more MSSVs are
inoperable, and revised the action statement in Mode 3 to maintain at
least two MSSVs operable on each SG.
Date of issuance: September 29, 2000.
Effective date: As of the date of issuance to be implemented prior
to startup from the 2R14 refueling outage.
Amendment No.: 222.
Facility Operating License No. NPF-6: Amendment revised the License
and TSs.
Date of initial notice in Federal Register: February 9, 2000 (65 FR
6405).
The January 26, May 17 (2 letters), May 31, and August 4, 2000,
supplemental letters provided clarifying information that was within
the scope of the original Federal Register notice and did not change
the staff's initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 29, 2000.
No significant hazards consideration comments received: No.

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

Date of amendment request: January 12, 2000, as supplemented by
letters dated June 15, 2000, and September 7, 2000.
Brief description of amendment: The proposed changes modify
Technical Specification (TS) 3.9.4, ``Containment Building
Penetrations,'' to allow the containment equipment door, airlocks, and
other penetrations to remain open, but capable of being closed, during
core alterations or movement of irradiated fuel in containment.
Additionally, a note, Bases changes, and Surveillance Requirements
changes provide further enhancements to clarify equipment door,
airlock, and penetration closure capability.
Date of issuance: October 2, 2000.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 169.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 23, 2000 (65
FR 9008). The June 15, 2000, and September 7, 2000, supplemental
letters provided clarifying information that did not expand the scope
of the original Federal Register notice, or change the scope of the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 2, 2000.
No significant hazards consideration comments received: No.

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

Date of application for amendments: September 20, 1999, as
supplemented May 12, 2000.
Brief description of amendments: The amendments revised the
standard to which the control room ventilation charcoal and
Supplementary Leak Collection and Release System (SLCRS) charcoal must
be laboratory tested as specified in: BVPS-1 Technical Specification
(TS) 4.7.7.1.1.c.2 for the Control Room Emergency Habitability Systems;
BVPS-1 TS 4.7.8.1.b.3 for the SLCRS; BVPS-2 TS 4.7.7.1.d for the
Control Room Emergency Air Cleanup and Pressurization System; and BVPS-
2 TS 4.7.8.1.b.3 for the SLCRS. Nuclear Regulatory Commission Generic
Letter 99-02, ``Laboratory Testing of Nuclear-Grade Activated
Charcoal,'' dated June 3, 1999, requested licensees to revise their TS
criteria associated with laboratory testing of ventilation charcoal to
a valid test protocol, which included American Society for Testing and
Materials (ASTM) D3803-1989. These license amendments revised the
charcoal laboratory standard to follow ASTM D3803-1989 for each BVPS
Unit. These license amendments also: (1) Revised the minimum amount of
output in kilowatts needed for the control room emergency ventilation
system heaters at each BVPS unit; (2) revised BVPS-1 SLCRS surveillance
testing criteria to be consistent with American Nuclear Standards
Institute/American Society of Mechanical Engineers N510-1980, the BVPS-
1 control room ventilation testing, and BVPS-2 SLCRS/control room
ventilation testing; and (3) made minor typographical corrections.
Date of issuance: September 29, 2000.
Effective date: As of date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 234 and 117.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 29, 2000 (65 FR
52449).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 29, 2000.
No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412,
Beaver Valley Power Station (BVPS-2), Unit 2, Shippingport,
Pennsylvania

Date of application for amendment: May 1, 2000, as supplemented
July 21, 2000.
Brief description of amendment: The amendment: (1) Revised
Technical Specification (TS) requirements regarding the minimum number
of radiation monitoring instrumentation channels required to be
operable during movement of fuel within the containment; (2) revised
the Modes in which the surveillance specified by Table 4.3-3,
``Radiation Monitoring Instrumentation Surveillance Requirements,''
Item 2.c.ii is required; (3) revised TS 3.9.4, ``Containment Building
Penetrations,'' to allow both personnel air lock (PAL) doors and

[[Page 62399]]

other containment penetrations to be open during movement of fuel
assemblies within containment, provided certain conditions are met; (4)
revised applicability and action statement requirements of TS 3.9.4. to
be for only during movement of fuel assemblies within containment; (5)
revised periodicity and applicability of Surveillance Requirement (SR)
4.9.4.1; (6) revised SR 4.9.4.2 to verify flow rate of air to the
supplemental leak collection and release system (SLCRS) rather than
verifying the flow rate through the system; (7) added two new SRs,
4.9.4.3 and 4.9.4.4, for verification and demonstration of SLCRS
operability; (8) modified TS 3/4.9.9 for the containment purge exhaust
and isolation system to be applicable only during movement of fuel
assemblies within containment; (9) revised associated TS Bases and made
editorial and format changes; and, (10) revised the BVPS-2 Updated
Final Safety Analysis Report (UFSAR) description of a fuel-handling
accident (FHA) and its radiological consequences. The changes to the
BVPS-2 UFSAR reflect a revised FHA analysis that the licensee performed
to evaluate the potential consequences of having containment
penetrations and/or the PAL open during movement of fuel assemblies
within containment. These UFSAR revisions include potential exclusion
area boundary, low population zone, and control room operator doses as
a result of an FHA.
Date of issuance: September 28, 2000.
Effective date: As of date of issuance. Technical Specification
changes shall be implemented within 60 days.
Amendment No: 116.
Facility Operating License No. NPF-73. Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: August 23, 2000 (65 FR
51342).
The July 21, 2000, letter provided clarifying information that did
not expand the scope of the amendment and did not change the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 28, 2000.
No significant hazards consideration comments received: No

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 application for amendments: February 16, 2000.
Brief description of amendments: (1) Accident monitoring
instrumentation for both St. Lucie Units 1 and 2, (2) motor operated
valve thermal overload protection bypass device TS for Unit 2, and (3)
an administrative change to the Unit 2 Technical Specification (TS)
Index.
Date of Issuance: October 4, 2000.
Effective Date: October 4, 2000.
Amendment Nos.: 165 and 109.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the TS.
Date of initial notice in Federal Register: April 5, 2000 (65 FR
17916).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 4, 2000.
No significant hazards consideration comments received: No.

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

Date of application for amendment: November 22, 1999, as
supplemented August 14, 2000.
Brief description of amendment: The amendment would adopt selected
NRC approved generic changes to the Improved Technical Specifications
(ITS) NUREGs. The 16 changes come from the Technical Specification Task
Force (TSTF) process developed by the Industry and the NRC. Three of
these changes are Bases-only changes but are included for completeness
relative to the TSTF process.
Date of issuance: October 3, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 234.
Facility Operating License No. DPR-49: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 12, 2000 (65 FR
1924).
The supplemental information contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice. The Commission's related evaluation of the amendment
is contained in a Safety Evaluation dated October 3, 2000.
No significant hazards consideration comments received: No.

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

Date of application for amendment: November 29, 1999, as
supplemented by letter dated May 2, 2000.
Brief description of amendment: The amendment changes Technical
Specification (TS) 3/4.6.6, ``Supplementary Leak Collection and Release
System''; TS 3/4.7.7, ``Control Room Emergency Ventilation System''; TS
3/4.7.9, ``Auxiliary Building Filter System''; and TS 3/4.9.12, ``Fuel
Building Exhaust System''; in response to Generic Letter 99-02,
``Laboratory Testing of Nuclear-Grade Activated Charcoal.''
Date of issuance: October 4, 2000.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 184.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 26, 2000 (65 FR
4287).
The letter dated May 2, 2000, provided clarifying information and
did not change the staff's initial proposed no significant hazards
consideration determination or expand the scope of the application as
published in the Federal Register.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated October 4, 2000.
No significant hazards consideration comments received: No.

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

Date of application for amendment: July 18, 2000.
Brief description of amendment: The amendment changes the Technical
Specifications to add operability requirements for the No. 12 residual
heat removal service water pump.
Date of issuance: October 2, 2000.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 113.
Facility Operating License No. DPR-22: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 23, 2000 (65 FR
51361).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 2, 2000.
No significant hazards consideration comments received: No.

[[Page 62400]]

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

Date of application for amendment: June 14, 2000, as supplemented
August 9, 2000.
Brief description of amendment: This amendment revised the TSs for
safety limit Minimum Critical Power Ratio from its current value of
1.10 to 1.09 for two recirculation-loop operation, and from 1.12 to
1.10 for single recirculation-loop operation.
Date of issuance: September 22, 2000.
Effective date: As of date of issuance, and shall be implemented
prior to startup for Cycle 14 operations, scheduled for October 2000.
Amendment No.: 236.
Facility Operating License No. DPR-44: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 26, 2000 (65 FR
46012). The August 9, 2000, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination or expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 22, 2000.
No significant hazards consideration comments received: No.

Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of application for amendment: April 27, 2000.
Brief description of amendment: The amendment would extend the
applicability of the current pressure-temperature limit curves and
overpressure protective system setpoints from 13.3 to 16.2 effective
full-power years.
Date of issuance: October 5, 2000.
Effective date: October 5, 2000.
Amendment No.: 202.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 29, 2000 (65 FR
52431).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 5, 2000.
No significant hazards consideration comments received: No.

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

Date of application for amendment: August 29, 2000, as supplemented
September 8, 2000.
Brief description of amendment: The amendment adapts the provisions
of the Boiling Water Reactor Standard Technical Specifications (STS)
regarding applicability of Technical Specifications 3.0.D and 4.0.D in
the event of plant shutdown.
Date of issuance: September 29, 2000.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 262.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes September 14, 2000 (65 FR 55650). That notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by October 16, 2000, but indicated that if the Commission makes
a final no significant hazards consideration determination any such
hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment finding of
exigent circumstances, state consultation, and final determination of
no significant hazards consideration determination are considered in a
Safety Evaluation dated September 29, 2000.

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

Date of application for amendment: April 27, 2000, as supplemented
September 5, 2000.
Brief description of amendment: The amendment changes the Trip
Level Settings for the Residual Heat Removal and Core Spray Start
Timers as well as the Automatic Depressurization System Auto-Blowdown
Timer.
Date of issuance: October 4, 2000.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 263.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 14, 2000 (65 FR
37428).
The September 5, 2000, supplement did not change the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 4, 2000.
No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

Date of application for amendments: March 2, 2000.
Brief description of amendments: The amendments modify the
requirements in Technical Specifications Section 3/4.6.3, ``Containment
Isolation Valves,'' by changing limiting conditions for operation (LCO)
3.6.3.1 and 3.6.3 for Unit Nos. 1 and 2, respectively. The changes
delete the asterisk (*) modifying the word OPERABLE in LCO 3.6.3.1
(Unit 1) and LCO 3.6.3 (Unit 2), and relocate its associated footnote
to the Action portion of the LCO.
Date of issuance: October 2, 2000.
Effective date: As of the date of issuance, and shall be
implemented within 60 days of issuance.
Amendment Nos.: 235 and 216.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 28, 2000 (65 FR
39959). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 2, 2000.
No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

Date of application for amendments: March 13, 2000
Brief description of amendments: The amendments revise TS Table
3.3-6, ``Radiation Monitoring Instrumentation,'' by changing the
Containment Gaseous Activity Monitor (R12A) alarm and trip setpoint for
the containment purge and pressure relief system isolation for Mode 6
(Refueling) operations.
Date of issuance: October 2, 2000.
Effective date: As of the date of issuance, and shall be
implemented within 60 days of issuance.
Amendment Nos.: 236 and 217.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 26, 2000 (65 FR
46013).

[[Page 62401]]

The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated October 2, 2000.
No significant hazards consideration comments received: No.

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

Date of application for amendments: May 3, 2000 (PCN-516), as
supplemented August 25, 2000.
Brief description of amendments: The amendments consist of changes
to the Technical Specifications that revise the pressure temperature
(P-T) limits for 20 effective full power years and reduce the minimum
boltup temperature from 86 deg.F to 65 deg.F. The P-T limits
calculations are based on the 1989 American Society of Mechanical
Engineers Appendix G methodology.
Date of issuance: September 28, 2000.
Effective date: September 28, 2000, to be implemented within 30
days of issuance.
Amendment Nos.: Unit 2--172; Unit 3--163.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 31, 2000 (65 FR
34749). The supplemental letter dated August 25, 2000, provided
clarifying information that was within the scope of the May 3, 2000,
application and the original Federal Register notice and did not change
the staff's initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 28, 2000.
No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia

Date of application for amendments: November 17, 1999, as
supplemented by letter dated August 21, 2000.
Brief description of amendments: The amendments revise TS 5.5.7,
``Ventilation Filter Testing Program'' to include the requirements for
laboratory testing of Engineered Safety Feature Ventilation System
charcoal samples in accordance with American Society Testing and
Materials D3803-1989 and the application of a safety factor of 2.0 to
the charcoal filter efficiency assumed in the plant design-basis dose
analyses. In addition, editorial revisions are being made to some
portions of TS Section 5.0 to reference the correct sections of
Regulatory Guide 1.52.
Date of issuance: October 3, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 223 and 164.
Facility Operating License Nos. DPR-57 and NPF-5: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 15, 1999 (64
FR 70091). The supplemental letter dated August 21, 2000, provided
clarifying information that did not change the scope of the November
17, 1999, application and the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 3, 2000.
No significant hazards consideration comments received: Yes. One
comment was received, and is addressed in the above-referenced Safety
Evaluation.

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

Date of amendment request: September 28, 1998, as supplemented on
April 22, 1999, April 27, 2000, and August 15, 2000.
Brief description of amendments: The amendments revise the
technical specifications (TSs) to eliminate the need to enter TS 3.0.3
when multiple trains of either the control room makeup and cleanup
filtration system or the fuel handling building exhaust air system are
inoperable by providing an allowed outage time of up to 12 hours to
restore at least one train to an operable status.
Date of issuance: September 26, 2000.
Effective date: September 26, 2000, to be implemented within 60
days.
Amendment Nos.: Unit 1--125; Unit 2--113.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 26, 2000 (65 FR
46016). The August 15, 2000, submittal provided clarifying information
that was within the scope of the revised application and Federal
Register notice and did not change the staff's revised proposed no
significant hazards considerations determination issued on July 26,
2000. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 26, 2000.
No significant hazards consideration comments received: No.

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

Date of application for amendments: May 16, 2000.
Brief description of amendments: These amendments change the
Technical Specifications (TSs) by replacing Surveillance Requirement
(SR) 4.8.1.1.2.c, for evaluating fuel oil for the emergency diesel
generators, with a Diesel Fuel Oil Program in Section 6. The revision
also deletes the portion of the SRs that specifies the use of sodium
hypochlorite solution in cleaning of the fuel oil storage tanks,
deletes the SR to perform a pressure test on the diesel generator fuel
oil system designed to American Society of Mechanical Engineers Section
III requirements, and corrects various typographical errors in the TS
and Bases. Two Bases pages are also added to each units TS. The
applicable TS Bases are also revised.
Date of issuance: October 2, 2000.
Effective date: October 2, 2000.
Amendment Nos.: 261 and 252.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the TS.
Date of initial notice in Federal Register: August 9, 2000 (65 FR
48758). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 2, 2000.
No significant hazards consideration comments received: No.

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

Date of application for amendments: February 4, 2000.
Brief description of amendments: These amendments change the
Technical Specifications (TS) to revise the cold leg accumulator volume
and pressure limits based on instrumentation changes, instrument
inaccuracies, and instrumentation tap locations. The applicable TS
bases are also revised.
Date of issuance: October 6, 2000.
Effective date: October 6, 2000.
Amendment Nos.: 262 and 253.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the TS.
Date of initial notice in Federal Register: May 17, 2000 (65 FR
31360). The Commission's related evaluation of

[[Page 62402]]

the amendment is contained in a Safety Evaluation dated October 6,
2000.
No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri

Date of application for amendment: May 25, 2000 (ULNRC-04258).
Brief description of amendment: The amendment expands (1) The range
of acceptable lift settings for the pressurizer safety valves (PSVs),
and (2) the tolerance (from 1% to 2%) of the
as-found, measured lift settings of tested PSVs, to be operable.
Following testing, however, the lift settings of the PSVs would remain
nor more than the current 1%. The amendment revises
Technical Specifications (TS) 3.3.2, ``Engineered Safety Features
Actuation System (ESFAS) Instrumentation,'' 3.4.10, ``Pressurizer
Safety Valves,'' and 3.4.11, ``Pressurizer Power Operated Relief Valves
(PORVs),'' of the Callaway TS. For TS 3.3.2, a new Action H for one or
more trains inoperable is added, the note for surveillance requirement
(SR) 3.3.2.14 is revised to identify another slave relay that the SR
would be applicable to, and the automatic PORV actuation is added to
Table 3.3.2-1, ``Engineered Safety Features Actuation System
Instrumentation.'' For TS 3.4.10, the range of allowable PSV lift
settings in the limiting condition for operation (LCO) is expanded from
2460 and 2510 to 2411 and
2509, and SR 3.4.10.1 is revised to state that, following
testing, the lift settings shall be ``within 1% of 2460 psig'' instead
of simply ``within 1%.'' The nominal PSV lift setting would be changed
from 2485 psig to 2460 psig. For TS 3.4.11, Actions A and B is revised
to be actions for inoperable PORVs either solely due to excessive PORV
seat leakage (Action A) or for reasons other than excessive seat
leakage (Action B), and Action E would remain an action for two
inoperable PORVs, but would be only for reasons other than excessive
seat leakage.
Date of issuance: September 25, 2000.
Effective date: September 25, 2000, to be implemented (including
issuing the revised EOP E-O and training all the control room operator
crews on the revised procedure) before the restart from refueling
outage 11, the next refueling outage for Callaway Plant, Unit 1,
scheduled to begin in Spring 2001.
Amendment No.: 137.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 28, 2000 (65 FR
29964). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 25, 2000.
No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri

Date of application for amendment: July 21, 2000 (ULNRC-04285), as
supplemented August 16, 2000.
Brief description of amendment: The amendment revises Limiting
Condition for Operation (LCO) 3.9.4, ``Containment Penetrations,'' of
the Callaway Technical Specifications (TS) to allow containment
penetrations with direct access to the outside atmosphere to be open
under administrative controls during refueling operations, by adding a
note to the LCO that states ``containment penetration flow path(s)
providing direct access from the containment atmosphere to the outside
atmosphere may be unisolated under administrative controls.'' In
addition, there is a format and editorial correction to TS 3.8.3,
``Diesel Fuel Oil, Lube Oil, and Start Air,'' to correct an error in
the conversion to the improved TS issued May 28, 1999, in Amendment No.
133.
Date of issuance: September 26, 2000.
Effective date: September 26, 2000, to be implemented (including
the completion of the administrative procedures that ensure that open
containment penetrations, with direct acess to the outside atmosphere
during refueling operations with core alterations and irradiated fuel
movement inside containment, will be promptly closed in the event of a
fuel handling accident inside containment) before refueling operations
during refueling outage 11.
Amendment No.: 138.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 23, 2000 (65 FR
51364). The August 16, 2000, supplement provided additional clarifying
information, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 26, 2000.
No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri

Date of application for amendment: September 8, 1999.
Brief description of amendment: The amendment authorizes revisions
to the descriptions of the steam generator tube rupture and main steam
line break accidents in the Callaway Plant, Unit 1 Final Safety
Analysis Report (FSAR) to reflect increases in the radiological dose
consequences calculated by the licensee for these accidents.
Date of issuance: September 27, 2000.
Effective date: September 27, 2000, to be implemented in the next
periodic update to the FSAR in accordance with 10 CFR 50.71(e).
Implementation of the amendment is the incorporation into the FSAR the
changes to the description of the facility as described in the
licensee's application dated September 8, 1999, and evaluated in the
staff's Safety Evaluation attached to the amendment.
Amendment No.: 139.
Facility Operating License No. NPF-30: The amendment revised the
Final Safety Analysis Report.
Date of initial notice in Federal Register: October 6, 1999 (64 FR
54383). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 27, 2000.
No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 11th day of October 2000.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management. Office of Nuclear
Reactor Regulation.
[FR Doc. 00-26645 Filed 10-17-00; 8:45 am]
BILLING CODE 7590-01-P