[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.

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

Calvert Cliffs Nuclear Power Plant, Inc., Docket 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.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the

[[Page 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:

    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 plant design or method of 
operation. This change only affects the plant in a cold shutdown or 
refueling condition and will not alter the basic operation of 
process variables, structures, systems, or components as described 
in the safety analyses. No new equipment is introduced.
    The proposed change does not affect the ability of low pressure 
ECCS injection/spray systems to perform their required safety 
functions. The essential safety function of providing water to 
reflood the reactor vessel following an inadvertent vessel draindown 
is 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.
    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 intended safety 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.

[[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