[Federal Register Volume 65, Number 240 (Wednesday, December 13, 2000)]
[Notices]
[Pages 77913-77934]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-31541]



[[Page 77913]]

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

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 November 17, 2000, through December 1, 2000. 
The last biweekly notice was published on November 29, 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.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC's Public Document 
Room, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. The filing of requests for a hearing 
and petitions for leave to intervene is discussed below.
    By January 12, 2001, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, located at One 
White Flint North, 11555 Rockville Pike (first Floor), Rockville, 
Maryland 20852. Publicly available records will be accessible 
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

[[Page 77914]]

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, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, by the above date. A copy of 
the petition should also be sent to the Office of the General Counsel, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to 
the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 
Publicly available records will be accessible electronically from the 
ADAMS Public Library component on the NRC Web site, http://www.nrc.gov 
(the Electronic Reading Room).

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

    Date of amendment request: July 5, 2000.
    Description of amendment request: The proposed amendment would 
revise the maximum power level specified in each unit's license; revise 
the value of rated thermal power of each unit from 3411 megawatts 
thermal (MWt) to 3586.6 MWt and the reference source for conversion 
factors in the calculation of Dose Equivalent Iodine 131 in the 
technical specification (TS) definitions; add a Departure from Nucleate 
Boiling Ratio (DNBR) limit specifically for a thimble cell; increase 
the minimum limit for reactor coolant system (RCS) total flow; revise 
the steam generator laser welded sleeve plugging limit; and reduce the 
peak calculated containment internal pressure Pa for the 
design basis loss-of-coolant accident (LOCA).
    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 proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    A. Evaluation of the Probability of Previously Evaluated 
Accidents.
    Plant systems and components have been verified to be capable of 
performing their intended design functions at uprated power 
conditions. Where necessary, some components will be modified prior 
to implementation of uprated power operations to accommodate the 
revised operating conditions. The analysis has concluded that 
operation at uprated power conditions will not adversely affect the 
capability or reliability of plant equipment. Current TS 
surveillance requirements ensure frequent and adequate monitoring of 
system and component operability. All systems will continue to be 
operated in accordance with current design requirements under 
uprated conditions, therefore no new components or system 
interactions have been identified that could lead to an increase in 
the probability of any accident previously evaluated in the Updated 
Final Safety Analysis Report (UFSAR). No changes were required to 
the Reactor Trip or Engineered Safety Features (ESF) setpoints.
    B. Evaluation of the Consequences of Previously Evaluated 
Accidents.
    The radiological consequences were reviewed for all design basis 
accidents (DBAs) (i.e., both Loss of Coolant Accident (LOCA) and 
non-LOCA accidents) previously analyzed in the UFSAR. The analysis 
showed that the resultant radiological consequences for both LOCA 
and non-LOCA accidents remain either unchanged or have not 
significantly increased due to operation at uprated power 
conditions. The radiological consequences of all DBAs continue to 
meet established regulatory limits.
    The proposed addition of Table E-7 of NRC Regulatory Guide 
1.109, ``Calculation of Annual Doses to Man from Routine Releases of 
Reactor Effluents for the Purpose of Evaluating Compliance with 10 
CFR Part 50, Appendix I,'' Revision 1, 1977, or International 
Commission on Radiological Protection (ICRP) 30, ``Limits for 
Intakes of Radionuclides by Workers,'' Supplement to Part 1, page 
192-212, Table titled, ``Committed Dose Equivalent in Target Organs 
or Tissues per Intake of Unit Activity,'' for thyroid dose 
conversion factors, will not significantly increase the consequences 
of an accident previously evaluated. If Regulatory Guide 1.109, or 
ICRP 30, Supplement to Part 1, are used to calculate maximum dose 
equivalent iodine specific activity, the total RCS iodine activity 
may increase, depending on the iodine nuclide mix, and this activity 
is used to calculate the doses resulting from a Main Steam Line 
Break (MSLB) or other analyzed accident. The calculated thyroid 
doses resulting from an MSLB or other analyzed accident would not 
increase as the corresponding dose conversion factors would be used 
to calculate the offsite thyroid doses. For a given Dose Equivalent 
I-131 concentration in the RCS, the offsite dose predicted using the 
dose conversion factors in either Table E-7 of Regulatory Guide 
1.109, or ICRP 30, Supplement to Part 1, is less than that predicted 
by Table III of Atomic Energy Commission (AEC) Technical Information 
Document TID-14844, ``Calculation of Distance Factors for Power and 
Test Reactor Sites,'' which is currently referenced in the TS 
definition of Dose Equivalent I-131.
    ICRP-30 is the updated reference for thyroid dose conversion 
factors used in the power uprate accident analysis radiological 
evaluation. The current version of 10 CFR 20, ``Standards for 
Protection Against Radiation,'' also utilizes ICRP-30 data.
    Based on the analysis, it is concluded that the proposed TS 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The configuration, operation and accident response of the Byron 
Station and the Braidwood Station systems, structures or components 
are unchanged by operation at uprated power conditions or by the 
associated proposed TS changes. Analyses of transient events have 
confirmed that no transient event results in a new sequence of 
events that could lead to a new accident scenario.
    The effect of operation at uprated power conditions on plant 
equipment has been

[[Page 77915]]

evaluated. No new operating mode, safety-related equipment lineup, 
accident scenario, or equipment failure mode was identified as a 
result of operating at uprated conditions. In addition, operation at 
uprated power conditions does not create any new failure modes that 
could lead to a different kind of accident. Minor plant 
modifications, to support implementation of uprated power 
conditions, will be made as required to existing systems and 
components. The basic design of all systems remains unchanged and no 
new equipment or systems have been installed which could potentially 
introduce new failure modes or accident sequences. No changes have 
been made to any Reactor Trip or ESF actuation setpoints.
    Based on this analysis, it is concluded that no new accident 
scenarios, failure mechanisms or limiting single failures are 
introduced as a result of the proposed changes. The proposed TS 
changes do not have an adverse effect on any safety-related system. 
Therefore, the proposed TS changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    A comprehensive analysis was performed to support the power 
uprate program at the Byron Station and the Braidwood Station. This 
analysis identified and defined the major input parameters to the 
NSSS [Nuclear Steam Supply System], reviewed NSSS design transients, 
and reviewed the capabilities of the NSSS fluid systems, NSSS/BOP 
[balance-of-plant] interfaces, NSSS control systems, and NSSS and 
BOP components. All appropriate NSSS accident analysis was 
reperformed to confirm acceptable results were maintained and that 
the radiological consequences remained within regulatory limits. The 
nuclear and thermal hydraulic performance of nuclear fuel was also 
reviewed to confirm acceptable results. The analysis confirmed that 
all NSSS and BOP systems and components are capable, some with minor 
modifications, to safely support operations at uprated power 
conditions.
    To support the operation of Byron Station, Units 1 and 2, and 
Braidwood Station, Units 1 and 2 at uprated power conditions, 
nuclear fuel Departure from Nucleate Boiling Ratio (DNBR) reanalysis 
was required to define new core limits, axial offset limits, and 
Condition II, ``Faults of Moderate Frequency,'' acceptability. This 
analysis included review of the following events: loss of RCS flow, 
reactor coolant pump locked rotor, feedwater malfunction, dropped 
control rod, steamline break, and control rod withdrawal from a 
subcritical condition. DNB design criteria was met for all events.
    NUREG-1431, ``Standard Technical Specifications, Westinghouse 
Plants,'' Revision 1, dated April 1995, allows Dose Equivalent I-131 
to be calculated using any one of three dose conversion factors; 
Table III of TID-14844, 1962, Table E-7 of NRC Regulatory Guide 
1.109, Revision 1, 1977, or ICRP 30, Supplement to Part 1. Using 
thyroid dose conversion factors other than those given in TID-14844 
results in lower doses and higher allowable activity but is 
justified by the discussion given in the Federal Register (i.e., 
Federal Register (FR) page 23360 Vol. 56, May 21, 1991). This 
discussion accompanied the final rulemaking on 10 CFR 20, 
``Standards for Protection Against Radiation,'' by the NRC. In that 
discussion, the NRC stated that it was incorporating modifications 
to existing concepts and recommendations of the ICRP into NRC 
regulations. Incorporation of the methodology of ICRP-30 into the 10 
CFR 20 revision was specifically mentioned with the changes being 
made resulting from changes and updates in the scientific techniques 
and parameters used in calculating dose. This FR reference clearly 
shows that the NRC was updating 10 CFR 20 to incorporate ICRP-30 
recommendations and data. Regulatory Guide 1.109 thyroid dose 
conversion factors are higher than the ICRP-30 thyroid dose 
conversion factors for all five iodine isotopes of concern. 
Therefore, using Regulatory Guide 1.109 thyroid dose conversion 
factors to calculate Dose Equivalent I-131 is more conservative than 
ICRP-30 and is therefore acceptable. For a given Dose Equivalent I-
131 concentration in the Reactor Coolant, the offsite dose predicted 
using the dose conversion factors in either Table E-7 of Regulatory 
Guide 1.109, Revision 1, NRC, 1977, or ICRP 30, Supplement to Part 
1, is less than that predicted by Table III of TID-14844 which is 
currently referenced in the TS definition of Dose Equivalent I-131.
    ICRP-30 is the updated reference source used in the power uprate 
accident analysis radiological evaluation. All regulatory acceptance 
criteria continue to be met and adequate safety margin is 
maintained.
    Revising the minimum limit for RCS total flow from greater than 
or equal to 371,400 gpm to greater than or equal to 380,900 gpm does 
not represent a significant reduction in the margin of safety. The 
reactor coolant pumps run at full flow and have a total flow 
capacity greater than 380,900 gpm. The analysis has shown that DNBR 
criteria has been met for all normal operational transients and loss 
of flow accident scenarios.
    The margin of safety of the reactor coolant pressure boundary is 
maintained under uprated power conditions. The design pressure of 
the reactor pressure vessel and reactor coolant system will not be 
challenged as the pressure mitigating systems were confirmed to be 
sufficiently sized to adequately control pressure under uprated 
power conditions.
    The proposed change revises the plugging limit for laser welded 
sleeves from 40% to 38.7% of nominal wall thickness. The analysis 
performed in support of the power uprate effort, indicated that it 
is necessary to remove steam generator (SG) tubes with laser welded 
sleeves from service upon discovering an imperfection depth of 38.7% 
wall thickness to ensure the structural integrity of SG tubes which 
have been sleeved thereby precluding the occurrence of an SG tube 
rupture of sleeved tubes under all operating conditions. The 
previous laser welded sleeve plugging limit was based on an analysis 
that used lower tolerance limit material strength values. The new 
analysis methodology, required for laser welded sleeves, uses 
minimum strength properties from the American Society of Mechanical 
Engineers Code. As determined by the new analysis, reducing the 
plugging limit from 40% to 38.7% maintains a comparable margin of 
safety to the previous analysis.
    Reanalysis of containment structural integrity under Design 
Basis Accident (DBA) conditions indicated that the safety margin 
improved, even though the mass and energy release due to a LOCA 
under uprated power conditions increases. Based on new and improved 
analytical methodologies, Pa, the peak calculated 
containment internal pressure for the design basis LOCA, is 42.8 
psig as compared to the current value of 47.8 psig for Unit 1; and 
is 38.4 psig as compared to the current value of 44.4 psig for Unit 
2, for both Byron Station and Braidwood Station.
    Radiological consequences of the following accidents were 
reviewed: Main Steamline Break, Locked Reactor Coolant Pump (RCP) 
Rotor, Locked RCP Rotor with Power-Operated Relief Valve Failure, 
Rod Ejection, Small Line Break Outside Containment, Steam Generator 
Tube Rupture, Large Break Loss of Coolant Accident, Small Break Loss 
of Coolant Accident, Waste Gas Decay Tank Rupture, Liquid Waste Tank 
Failure, and Fuel Handling Accident. The resultant radiological 
consequences for each of these accidents did not show a significant 
change due to uprated power conditions and 10 CFR 100 limits 
continue to be met.
    The analyses supporting the power uprate program have 
demonstrated that all systems and components are capable of safely 
operating at uprated power conditions. All design basis accident 
acceptance criteria will continue to be met. Therefore, it is 
concluded that the proposed TS changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck 
Plant

    Date of amendment request: July 7, 2000.
    Description of amendment request: The proposed amendment would add 
a new license condition which would approve the License Termination 
Plan dated July 7, 2000, and allow the licensee to make changes to the 
approved License Termination Plan without prior Nuclear Regulatory 
Commission (NRC) approval if certain

[[Page 77916]]

criteria specified in the license condition are met.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Connecticut Yankee 
Atomic Power Company (CYAPCO) has provided its analysis of the issue of 
no significant hazards consideration, which is presented below:
    CYAPCO has reviewed the proposed change to the Operating License 
in accordance with the requirements of 10 CFR 50.92, ``Issuance of 
Amendment,'' and concluded that the change does not involve a 
significant hazards consideration (SHC). The proposed change does 
not involve an SHC because the change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Currently, the bounding airborne radioactivity event given in 
the Haddam Neck Plant UFSAR [Updated Final Safety Analysis Report] 
is the resin container accident. Whereas previously doses associated 
with gaseous waste system accidents would have bounded those 
associated with solid waste system failures, the small amount of 
radioactivity contained within the gaseous radioactive waste system 
with the plant in the permanently defueled condition results in this 
system's failure no longer being bounding. The curie content of the 
resin container was based on the actual radioactivity inventory 
collected on the resin from the reactor coolant system 
decontamination. This corresponded to approximately 90% of the NRC 
Class C burial limits. Consistent with NUREG-0782 for a resin fire, 
one percent of the activity of the container was assumed to be 
released to the environment. The 1% bounds the potential airborne 
release fraction from various resin incidents, such as an exothermic 
reaction during dewatering, dropping of a high integrity container, 
or a resin spill. Other airborne particulate radwaste or radioactive 
material accidents considered in the UFSAR but bounded by the resin 
container fire are as follows:
     a fire in the radwaste storage facility,
     a drop of a component (e.g., steam generator, reactor 
vessel, or heat exchanger) being removed from the site,
     a van of radioactive waste materials consumed by a fire 
while stored in the yard area on-site,
     a radiological HEPA [High-Efficiency Particulate Air] 
filter rupture,
     segmentation of components or structures during loss of 
local engineering controls,
     an oxyacetylene tank explosion, or
     an explosion of liquid propane gas leaked from a front-
end loader.
    The UFSAR also discusses a fuel handling accident in the fuel 
building, involving the drop of a spent fuel assembly onto the fuel 
racks. The postulated drop assumes the rupture of all fuel rods in 
the associated assembly. The probability or consequences of this 
accident would not be increased during any future fuel transfer 
operations in the spent fuel pool related to decommissioning. 
Transfer of the spent fuel to canisters for dry cask storage will 
involve additional restrictions contained in the cask certificate of 
compliance in order to maintain decommissioning activities within 
the assumption and consequences of the fuel handling accident.
    The requested license amendment is consistent with plant 
activities described in the Post Shutdown Decommissioning Activities 
Report (PSDAR) and the HNP [Haddam Neck Plant] Decommissioning 
UFSAR. Accordingly, no systems, structures, or components that could 
initiate the previously evaluated accidents or are required to 
mitigate these accident are adversely affected by this proposed 
change. Therefore, the proposed change does not involve an increase 
in the probability or consequences of any previously evaluated 
accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Accident analyses related to decommissioning activities are 
addressed in the UFSAR. The requested license amendment is 
consistent with the plant activities described in the HNP 
Decommissioning UFSAR and the PSDAR. Thus, the proposed change does 
not affect plant systems, structures, or components in a way not 
previously evaluated. No new failure mechanisms will be created by 
this activity, and the proposed activity does not create the 
possibility of a new or different kind of accident than those 
previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The License Termination Plan (LTP) is a plan for demonstrating 
compliance with the radiological criteria for license termination as 
provided in 10 CFR 20.1402. The margin of safety defined in the 
statements of consideration for the final rule on the Radiological 
Criteria for License Termination is described as the margin between 
the 100 mrem/yr public dose limit established in 10 CFR 20.1301 for 
licensed operation and the 25 mrem/yr dose limit to the average 
member of the critical group at a site considered acceptable for 
unrestricted use (one of the criteria of 10 CFR 20.1402). This 
margin of safety accounts for the potential effect of multiple 
sources of radiation exposure to the critical group. Since the 
License Termination Plan was designed to comply with the 
radiological criteria for license termination for unrestricted use, 
the LTP supports this margin of safety.
    In addition, the LTP provides the methodologies and criteria 
that will be used to perform remediation activities of residual 
radioactivity to demonstrate compliance with the ALARA [as low as 
reasonably achievable] criterion of 10 CFR 20.1402.
    Additionally, the LTP was designed with recognition that (a) the 
methods in MARSSIM (Multi-Agency Radiation Survey and Site 
Investigation Manual) and (b) the building surface contamination 
levels are not directly applicable to use with complex nonstructural 
components. Therefore, the LTP states that nonstructural components 
remaining in buildings (e.g., pumps, heat exchangers, etc.) will be 
evaluated against the criteria of RG [Regulatory Guide] 1.86 to 
determine if the components can be released for unrestricted use. 
The LTP also states that materials, surveyed and evaluated as a part 
of normal decommissioning activities and prior to implementation of 
the final status survey, will be surveyed for release using current 
site procedures to demonstrate compliance with the ``no detectable'' 
criteria. Such materials that do not pass these criteria will be 
controlled as contaminated.
    Also, as previously discussed, the bounding accident for 
decommissioning is the resin container accident. Since the bounding 
decommissioning accident results in more airborne radioactivity than 
can be released from other decommissioning events, the margin of 
safety associated with the consequences of decommissioning accidents 
is not reduced by this activity.
    Thus, 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: Mr. Robert K. Gad, III, Ropes & Gray, One 
International Plaza, Boston, Massachusetts 02110-2624.
    NRC Section Chief: Michael T. Masnik.

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

    Date of amendment request: November 2, 2000.
    Description of amendment request: The proposed amendment would 
revise the WNP-2 Technical Specifications (TS) to incorporate long-term 
power stability solution requirements. The proposed changes reflect: 
(1) The addition of a new TS Section 3.3.1.3, ``Oscillation Power Range 
Monitoring (OPRM) Instrumentation,'' (2) a revision to TS Section 
3.4.1, ``Recirculation Loops Operating,'' to remove monitoring 
specifications that would no longer be necessary upon activation of the 
automatic OPRM instrumentation, and (3) a revision to TS 5.6.5 to 
include in the Core Operating Limits Report (COLR) the applicable 
operating limits for the OPRMs, and also reference the topical report 
which describes the analytical methods used to determine the setpoint 
values for the OPRM.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 77917]]


    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change specifies limiting conditions for operation, 
required actions and surveillance requirements for the OPRM system 
and allows operation in regions of the power-to-flow map currently 
restricted by the requirements of interim corrective actions (ICAs) 
and certain limiting conditions of operation of Technical 
Specification 3.4.1. The restrictions of the ICAs and Technical 
Specification 3.4.1 were imposed to ensure adequate capability to 
detect and suppress conditions consistent with the onset of thermal-
hydraulic oscillations that may develop into a thermal-hydraulic 
instability event. A thermal-hydraulic instability event has the 
potential to challenge the minimum critical power ratio (MCPR) 
safety limit. The OPRM system can automatically detect and suppress 
conditions necessary for thermal-hydraulic instability. With the 
installation of the OPRM system, the restrictions of the ICAs and 
Technical Specification 3.4.1 are no longer required to prevent a 
potential challenge to the MCPR safety limit during an anticipated 
instability event.
    The probability of a thermal-hydraulic event is dependent on 
power-to-flow conditions such that only during operation inside 
specific regions of the power-to-flow map, in combination with power 
shape and inlet enthalpy conditions, can the occurrence of an 
instability event be postulated to occur. Operation in these regions 
may increase the probability that operation with conditions 
necessary for a thermal-hydraulic instability can occur. When the 
OPRM system is operable, conditions consistent with the imminent 
onset of oscillations are automatically detected and the conditions 
necessary for oscillations are suppressed, which decreases the 
probability of an instability event. In the event the trip 
capability of the OPRM is not maintained, the proposed change limits 
the period of time before an alternate method to detect and suppress 
thermal-hydraulic oscillations is required. The probability of a 
thermal-hydraulic instability event may be increased during the 
limited period of time that operation is allowed at conditions 
otherwise requiring the trip capability of the OPRM to be 
maintained. However, since the duration of this period of time is 
limited, the increase in the probability of a thermal-hydraulic 
instability event is not significant.
    The proposed change requires the OPRM system to be operable and, 
thereby, ensures mitigation of thermal-hydraulic instability events 
with a potential to challenge the MCPR safety limit when initiated 
from anticipated conditions, by detection of the onset of 
oscillations and actuation of an RPS [reactor protection system] 
trip signal. The OPRM also provides the capability of an RPS trip 
being generated for thermal-hydraulic instability events initiated 
from unanticipated, but postulated conditions. These mitigating 
capabilities of the OPRM system will become available as a result of 
the proposed change and have the potential to reduce the 
consequences of anticipated and postulated thermal-hydraulic 
instability events. The OPRM installation has been evaluated and 
does not alter the function or capability of any other installed 
equipment such as the average power range monitoring (APRM) system 
or the RPS to mitigate the consequences of postulated events.
    Therefore, operation of WNP-2 in accordance with the proposed 
amendment will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change specifies limiting conditions for operation, 
required actions and surveillance requirements of the OPRM system 
and allows operation in regions of the power-to-flow map currently 
restricted by the requirements of ICAs and Technical Specification 
3.4.1. The OPRM system uses input signals shared with APRM and rod 
block functions to monitor core conditions and generate an RPS trip 
when required. Quality requirements for software design, testing, 
implementation and module self-testing of the OPRM system provide 
assurance that no new equipment malfunctions due to software errors 
are created. The design of the OPRM system also ensures that neither 
operation nor malfunction of the OPRM system will adversely impact 
the operation of other systems and no accident or equipment 
malfunction of these other systems could cause the OPRM system to 
malfunction or cause a different kind of accident.
    Operation in regions currently restricted by the requirements of 
ICAs and Technical Specification 3.4.1 is within the nominal 
operating domain and ranges of plant systems and components, and 
within the range for which postulated equipment and accidents have 
been previously evaluated.
    Therefore, operation of WNP-2 in accordance with the proposed 
amendment will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change specifies limiting conditions for operation, 
required actions and surveillance requirements of the OPRM system 
and allows operation in regions of the power-to-flow map currently 
restricted by the requirements of ICAs and Technical Specification 
3.4.1.
    The OPRM system monitors small groups of LPRM [local power range 
monitor] signals for indication of local variations of core power 
consistent with thermal-hydraulic oscillations and generates an RPS 
trip when conditions consistent with the onset of oscillations are 
detected. An unmitigated thermal-hydraulic instability event has the 
potential to result in a challenge to the MCPR [minimum critical 
power ratio] safety limit. The OPRM system provides the capability 
to automatically detect and suppress conditions which might result 
in a thermal-hydraulic instability event and, thereby, maintains the 
margin of safety by providing automatic protection for the MCPR 
safety limit while significantly reducing the burden on the control 
room operators. In the event the trip capability of the OPRM is not 
maintained, the proposed change limits the period of time before an 
alternate method to detect and suppress thermal-hydraulic 
oscillation is required. The alternate method to detect and suppress 
oscillations would be comparable to current actions required by the 
interim corrective actions and no significant reduction in the 
margin of safety would result in the event that an unmitigated 
instability event occurred.
    Operation in regions currently restricted by the requirements of 
ICAs and Technical Specification 3.4.1 is within the nominal 
operating domain and ranges of plant systems and components, and 
within the range assumed for initial conditions considered in the 
analysis of anticipated operational occurrences and postulated 
accidents.
    Therefore, operation of WNP-2 in accordance with the proposed 
amendment will 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: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: August 29, 2000.
    Description of amendment request: The proposed changes to the 
Arkansas Nuclear One, Unit 1 (ANO-1) Technical Specifications (TS) 
provide for the use of an Alternate Repair Criteria (ARC) for steam 
generator tubes with indications of outer diameter intergranular attack 
(ODIGA) within the upper tube sheet region of the once-through steam 
generators (OTSGs). Amendment 202 to the ANO-1 TS dated October 4, 
1999, allowed the ARC for ODIGA indications only during Operating Cycle 
16 at ANO-1. The proposed change would allow continued operation beyond 
Cycle 16 for ANO-1 with OTSG tubes that have ODIGA indications that are 
located in a defined area of the upper tube sheet.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 77918]]


    An evaluation of the proposed change has been performed in 
accordance with 10 CFR 50.91(a)(1) regarding no significant hazards 
considerations using the criteria in 10 CFR 50.92(c). A discussion 
of these criteria as they relate to this amendment request follows:

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

    The purpose of the periodic surveillance performed on the OTSGs 
in accordance with ANO-1 Technical Specification (TS) 4.18 is to 
ensure that the structural integrity of this portion of the reactor 
coolant system will be maintained. The TS plugging limit of 40% of 
the nominal tube wall thickness requires tubes to be repaired or 
removed from service because the tube may become unserviceable prior 
to the next inspection. Unserviceable is defined in the TS as the 
condition of a tube if it leaks or contains a defect large enough to 
affect its structural integrity in the event of an operating basis 
earthquake, a loss-of-coolant accident, or a steam line or feedwater 
line break. The proposed TS change allows OTSG tubes with ODIGA 
indications contained within a defined area of the UTS [upper tube 
sheet] to remain in service with existing degradation exceeding the 
existing 40% through-wall (TW) plugging limit.
    Extensive testing and plant experience has illustrated that 
ODIGA flaws confined to this area within the OTSG will not result in 
tube burst and tube leakage is unlikely. Therefore, allowing ODIGA 
flaws in this specific region to remain in service will not alter 
the conditions assumed in the current ANO-1 accident analysis for 
OTSG tube failures under postulated accident conditions. In 
addition, the condition of the OTSG tubes in this region are 
monitored during regular inspection intervals to assess for evidence 
of growth. Any growth noted will be addressed through testing and 
the operational assessment. Therefore, ANO-1 has determined that the 
identification, testing, monitoring, assessment, and corrective 
action programs provided in ANO [Arkansas Nuclear One] Engineering 
Report No. 00-R-1005-01, sufficiently supports this change request.
    Application of the ODIGA alternate repair criteria will allow 
leaving tubes with ODIGA indications found in the defined area of 
the UTS in service while ensuring safe operation by monitoring and 
assessing the present and future conditions of the tubes. ANO-1 has 
operated since 1984 with ODIGA affected tubes in service with no 
appreciable effect on structural integrity or indications of tube 
leakage from ODIGA sources within the UTS. Through the inspection, 
testing, monitoring, and assessment program previously mentioned, 
and the on-line leak detection capabilities available during plant 
operation, continued safe operation of ANO-1 is reasonably assured.
    Therefore, the application of the ODIGA alternate repair 
criteria does not involve a significant increase in the probability 
or consequences of any accident previously evaluated.

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

    The implementation of the ODIGA alternate repair criteria will 
not result in any failure mode not previously analyzed. The OTSGs 
are passive components. The intent of the TS surveillance 
requirements are being met by these proposed changes in that 
adequate structural and leak integrity will be maintained. 
Additionally, the proposed change does not introduce any new modes 
of plant operation.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The application of an alternate repair criteria for ODIGA 
provides adequate assurance with margin that ANO-1 steam generator 
tubes will retain their integrity under normal and accident 
conditions. The structural requirements of ODIGA affected tubes have 
been evaluated satisfactorily and meet or exceed regulatory 
requirements. Leakage rates for these tubes within the defined 
region of the upper tubesheet are essentially zero and are 
reasonably assured to remain within the assumptions of the accident 
analysis by proper application of the ODIGA alternate repair 
criteria program. Assuming high differential pressures following an 
ATWS [Anticipated Transient Without Scram] or MSLB [Main Steam Line 
Break], if the ODIGA patches leak, the leakage would be less than 
the normal makeup capacity of the reactor coolant system. Since no 
appreciable impact is evidenced on the tubes structural integrity or 
its resulting leak rate, the margin to safety remains effectively 
unaltered.

    Therefore, this change does not involve a significant reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: September 28, 2000.
    Description of amendment request: The proposed changes to the 
Arkansas Nuclear One, Unit 1 (ANO-1) technical specifications revise 
the safety-related 4160 Volt (V) bus loss-of-voltage and 480 V bus 
degraded voltage relay allowable values.
    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:

    An evaluation of the proposed change has been performed in 
accordance with 10 CFR 50.91(a)(1) regarding no significant hazards 
considerations using the standards in 10 CFR 50.92(c). A discussion 
of these standards as they relate to this amendment request follows:

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

    The two 4160 V vital bus loss-of-voltage protection relays that 
are provided on each of the 4160 V safety buses act to mitigate the 
consequences of an accident by detecting a loss of voltage, 
isolating the safety buses, initiating load shedding schemes, and 
starting the associated emergency diesel generator (EDG). The safety 
function of the relays is unchanged by the proposed setpoint 
revisions. The revised settings for the loss-of-voltage protection 
relays will continue to provide the safety function with no 
appreciable additional time delay. The proposed time delays are 
within those assumed in the ANO-1 safety analyses. Additionally, the 
lower voltage settings will aid in preventing unnecessary isolation 
from the off-site power sources, which in turn will reduce the 
probability of a loss of off-site power to the unit due to off-site 
power system transients. Since the proposed change does not 
adversely impact the mitigating function of the relays, the 
consequences of an accident previously evaluated remains unchanged.
    The two degraded voltage protection relays that are provided on 
each of the 480 V safety buses act to mitigate the consequences of 
an accident by detecting a sustained undervoltage condition, 
isolating the safety buses from offsite power, and starting the 
associated EDG. This safety function is unchanged by the proposed 
setpoint revisions. The revised settings for the degraded voltage 
protection relays will continue to provide the safety function of 
protecting the associated Class IE equipment from the effects of a 
low voltage condition. There is no proposed change to the existing 
timer setting and the time delays remain within those assumed in the 
ANO-1 safety analyses. Additionally, the revised allowable voltage 
settings will not result in any unnecessary isolation from the off-
site power sources. Since the proposed change does not adversely 
impact the mitigating function of the relays, the consequences of an 
accident previously evaluated remains unchanged.
    The ANO-1 technical specifications will continue to require the 
4160 V bus loss-of-voltage functions and 480 V bus degraded voltage 
functions to be surveillance tested at their present frequency 
without changing the modes in which the surveillance is required or 
the modes of applicability for these components. The technical 
specifications will continue to require the same actions as 
currently exist for the inoperability of one or more of the 4160 V 
bus loss-of-voltage channels or the 480 V bus degraded voltage 
channels.

[[Page 77919]]

    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.

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

    The proposed change introduces no new modes of plant operation 
or new plant configuration that could lead to a new or different 
kind of accident from any previously evaluated being introduced. The 
4160 V vital bus loss-of-voltage protection relays are required to 
operate following a complete loss of off-site power to initiate the 
bus power source transfer to on-site power, i.e., the EDGs, to 
prevent a loss of all AC power. Likewise, the 480 V bus degraded 
voltage relays are required to operate upon detection of a sustained 
undervoltage condition to protect the Class IE components from 
damage from low voltage by initiating transfer of the 4160 V safety 
bus power source to the EDG. These safety functions are unchanged by 
the proposed setpoint revisions, and the proposed setpoints continue 
to provide the required actions consistent with the ANO-1 safety 
analysis.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The two undervoltage relays located on each 4160 V safety bus 
are provided to detect loss-of-voltage, isolate the safety buses, 
initiate load shedding, and start the EDGs. The two undervoltage 
relays located on each 480 V safety bus are provided to detect 
sustained undervoltage, isolate the safety buses, and start the 
EDGs. These safety functions are unchanged by the proposed setpoint 
revisions. The proposed changes to the allowable values for both 
loss-of-voltage and degraded voltage relays incorporate channel 
uncertainties and calibration tolerances, while fully meeting their 
required safety functions of loss-of-voltage and degraded voltage 
protection without resulting in undesired tripping of the offsite 
power source.
    The lower loss-of-voltage values do not affect the margin of 
safety since there is no appreciable time difference in reaching the 
lower setpoints during a loss-of-voltage event. The maximum proposed 
time delay allowable value with the minimum loss-of-voltage relay 
allowable value is within that used in the ANO-1 safety analysis. 
The revised allowable values for the loss-of-voltage relays will 
continue to provide the safety function with no appreciable 
additional time delay. Additionally, the lower voltage settings will 
help to prevent unnecessary isolation from the off-site power 
sources due to off-site perturbations in the electrical grid, and 
thus contribute to increasing the margin of safety. Also, the 
slightly higher range of allowable values for the degraded voltage 
settings allows enhanced protection of the Class IE components, but 
does not result in undesired tripping of the offsite power source 
for the analyzed grid minimum normal condition. The degraded voltage 
relays, therefore, also act to contribute to an increased margin of 
safety.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Therefore, based upon the reasoning presented above and the 
previous discussion of the amendment request, Entergy Operations, 
Inc. has determined that the requested change does not involve a 
significant hazards consideration.

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

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: September 28, 2000.
    Description of amendment request: The proposed changes to Arkansas 
Nuclear One, Unit 1 (ANO-1), Technical Specifications (TS) provide for 
the implementation of a revised reroll repair process for ANO-1 Once-
Through Steam Generators (OTSG). The current TSs limit application of 
the reroll repair process to repair tubes with defects in the upper 
tubesheet area only, using a 1 inch roll length, and allow the reroll 
repair process to be performed only once per steam generator tube. The 
requested amendment would allow the reroll repair process to be used 
multiple times for a single tube and would allow the repairs in both 
the upper and lower tubesheets.
    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:

    An evaluation of the proposed change has been performed in 
accordance with 10 CFR 50.91(a)(1) regarding no significant hazards 
considerations using the criteria in 10 CFR 50.92(c).
    OTSG tubesheet areas where reroll installation is excluded are 
specified in Appendix A of topical report BAW-2303P, Revision 4 [A 
non-proprietary version of the report, BAW-2303NP, Revision 4, 
``OTSG Repair Roll Qualification Report,'' was submitted on October 
26, 2000.]. The following discussion applies to areas of the OTSG 
tubesheets where installation of reroll repairs is permitted:

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

    Two types of repair rolls have been developed for installation 
in the OTSGs, a single 1-inch roll expansion and an overlapping roll 
consisting of two 1-inch roll expansions. The overlapping roll 
provides a minimum of 1\5/8\ inch effective roll expansion. There is 
an additional \1/4\-inch roll transition region on each end of the 
roll expansion and a new leak-limiting pressure boundary is created 
by the repair roll. Applicable OTSG transient conditions were 
evaluated to develop a set of bounding test conditions for 
application to both types of repair rolls. Testing included 
examination of the effects of crevice deposits, cyclic loading, tube 
yield strength, differential dilations, axial loads and internal 
pressure.
    Test results conclude that the single 1-inch minimum repair roll 
is structurally adequate to prevent tube slip during all non-faulted 
operating transients. A small amount of slippage is acceptable 
provided the tube does not slip out of the tubesheet and tube bow 
due to post-faulted transient heatup does not result in tube 
failure. Exclusion areas are established in the tubesheets to 
provide assurance that tube will not slip out of the tubesheet . The 
1\5/8\ inch minimum overlapping roll is structurally acceptable 
based on the bounding evaluation of the single 1-inch repair roll.
    Bounding leak rates are applied based on tubesheet depth and 
radial position. A post-slip leak rate is applied to any location 
where there is potential for repair roll slip during a postulated 
accident. The bounding leak rates are very conservative because the 
leakage is based on test samples with a full circumferential sever 
outboard of the repair roll. The majority of the degradation in the 
tubesheets is comprised of short, axial cracks for which the leakage 
would be much less under axial tensile loads than for the tested 
severed tube. In addition, repair rolls will actually slip only if 
the tube is severed outboard of the repair roll. Since the majority 
of the degradation in the region of the roll joints has been 
identified as small axial cracks, the probability of the repair roll 
maintaining structural integrity is very high and the potential for 
a joint to slip is very low. The leakage from each repair roll that 
serves as a pressure boundary is added to the leakage from all other 
sources and the total leakage must be within current accident 
analysis limits.
    The application of the reroll repair process as described in 
topical report BAW-2303P, Revision 4 will not alter the conditions 
assumed in the current ANO-1 accident analysis for OTSG tube 
failures under postulated accident conditions. In addition, the 
condition of the OTSG tubes in this region are monitored during 
regular inspection intervals to assess for evidence of degradation. 
Any degradation noted will be addressed in the operational 
assessment and appropriate actions taken.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.

Criterion 2--Does Not Create the Possibility

[[Page 77920]]

of a New or Different Kind of Accident from any Previously 
Evaluated.

    The reroll process establishes a new pressure boundary for the 
associated tube in the tubesheet region inboard of the flaw. The new 
roll transition may eventually develop primary water stress 
corrosion cracking (PWSCC) and require additional repair. Industry 
experience with roll transition cracking has shown that PWSCC in 
roll transitions are normally short axial cracks, with extremely low 
leak rates. The standard MRPC eddy current inspection during the 
refueling outages have proven to be successful in detecting these 
defects.
    In the unlikely event the rerolled tube failed and severed 
completely at the heel transition of the reroll region, the tube 
would retain engagement in the tubesheet bore, preventing any 
interaction with neighboring tubes. In this case, leakage is 
minimized and is well within the assumed leakage of the design basis 
tube rupture accident. In addition, the possibility of rupturing 
multiple steam generator tubes is unaffected. Therefore, this change 
does not create the possibility of a new or different kind of 
accident from any previously evaluated.

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

    The repair roll is applicable to repairing axial, volumetric, or 
circumferential indications. Testing was conservatively performed 
with the assumption that the tube is severed at the heel transition 
(360 degree and 100% through-wall circumferential defect). The joint 
strength margin (actual load/limiting load) was calculated for each 
tubesheet depth and radial position for the cooldown transient to 
ensure margin against slip for non-faulted conditions. All locations 
showed a joint strength margin less than 0.65 with an acceptable 
margin less than 1.0.
    A tube with degradation can be kept in service through the use 
of the reroll process. The new roll expanded interface created with 
the tubesheet satisfies all of the necessary structural and leakage 
requirements. Since the joint is constrained within the tubesheet 
bore, there is no additional risk associated with tube rupture. 
Therefore, the analyzed accident scenarios remain bounding, and the 
proposed modifications to the reroll process do not reduce the 
margin of safety.
    Therefore, based upon the reasoning presented above and the 
previous discussion of the amendment request, Entergy Operations has 
determined that the requested change does not involve a significant 
hazards consideration.

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

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

    Date of amendment request: November 9, 2000.
    Description of amendment request: The proposed amendment requests 
fourteen of the simpler, generic administrative/editorial/consistency 
improvements agreed upon between the Nuclear Energy Institute Technical 
Specification Task Force (TSTF) and the NRC, subsequent to the 
conversion of the Perry Technical Specifications to the improved 
Standard Technical Specifications. The proposed amendment requests 
Perry-specific versions of TSTF 5, ``Delete Notification, Reporting, 
and Restart Requirements if a Safety Limit is Violated;'' TSTF 32, 
``Slow/Stuck Control Rod Separation Criteria;'' TSTF 38, ``Revise 
Visual Surveillance of Batteries to Specify Inspection is for 
Performance Degradation;'' TSTF 52, ``Implement 10 CFR Part 50, 
Appendix J, Option B;'' TSTF 65, ``Use of Generic Titles for Utility 
Positions;'' TSTF 104, ``Relocate to the Bases the Discussion of 
Exceptions to Limiting Condition for Operation (LCO) 3.0.4;'' TSTF 106, 
``Change to Diesel Fuel Oil Testing Program;'' TSTF 118, 
``Administrative Controls Program Exceptions;'' TSTF 152, ``Revise 
Reporting Requirements to be Consistent with 10 CFR Part 20;'' TSTF 
153, ``Clarify Exception Notes to be Consistent with the Requirement 
being Excepted;'' TSTF 166, ``Correct Inconsistency between LCO 3.0.6 
and the Safety Functional Determination Program (SFDP) Regarding 
Performance of an Evaluation;'' TSTF 258, ``Changes to Section 5.0, 
Administrative Controls;'' TSTF 278, ``Battery Cell Parameters (LCO 
3.8.6) includes more than Table 3.8.6-1 Limits;'' and TSTF 279, 
``Remove the Words `Including Applicable Supports' from the Description 
of the Inservice Testing Program.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. This proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes involve reformatting and rewording of the 
existing Technical Specifications to be consistent with regulations 
or other existing Technical Specifications, or the changes do not 
involve a change in intent. The proposed changes also involve 
Technical Specification requirements that are administrative rather 
than technical in nature. As such, this change does not affect 
initiators of previously evaluated events, or assumed mitigation of 
accident or transient events. Therefore, this change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. This proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or 
changes in methods governing normal plant operation. The proposed 
changes will not impose new or eliminate old requirements on design 
or operation of the plant. The administrative changes also do not 
introduce new initiators of events. Thus, this change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. This proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change has no impact on any safety analysis 
assumptions or design basis margins. This change is administrative 
in nature. The proposed changes will not impose new or eliminate old 
requirements on design or operation of the plant. Therefore, the 
change does not involve a significant reduction in a margin of 
safety.

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

GPU Nuclear, Inc., Three Mile Island Nuclear Station, Unit 2, Docket 
No. 50-320, Dauphine County, Pennsylvania

    Date of amendment request: July 25, 2000.
    Description of amendment request: The proposed technical 
specifications change request (TSCR) is to revise Three Mile Island 
Nuclear Generating Station, Unit 2 (TMI-2), Technical Specification 
(TS) Section 6.7.2 to eliminate a change associated with periodic 
reviews of procedures. Currently, TS 6.7.2 states that required 
procedures shall be reviewed periodically as required by American 
National Standards Institute (ANSI) N18.7-1976 (a biennial review). 
This TSCR proposes to revise the wording for TS 6.7.2 to be essentially 
identical with the Three Mile Island, Unit 1 (TMI-1), TS requirements 
for procedure reviews, which states that

[[Page 77921]]

required procedures shall be revised periodically, as set forth in 
administrative procedures (currently a biennial review). This TSCR 
would also be consistent with the TMI-2 Post-Defueling Monitored 
Storage (PDMS) Quality Assuance (QA) Plan, which states that 
``Procedural documentation shall be periodically reviewed for adequacy 
as set forth in administrative procedures.''
    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:

    Applying the three standards set forth in 10 CFR 50.92, the 
proposed changes to the Technical Specifications involve no 
significant hazards consideration. The proposed changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators or 
assumptions are affected. The proposed changes have no effect on any 
plant systems. All Limited Conditions for PDMS and Safety Limits 
specified in the Technical Specifications will remain unchanged.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because no accident conditions or 
assumptions are affected. The proposed changes do not alter the 
source term, containment isolation, or allowable radiological 
consequences. The change in specified periodic procedure review 
requirements will have no adverse effect on any plant system.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new 
accident initiators or assumptions are introduced by the proposed 
changes. The proposed changes have no direct effect on any plant 
systems. The changes do not affect any system functional 
requirements, plant maintenance, or operability requirements.
    3. Not involve a significant reduction in the margin of safety 
because the proposed changes do not involve significant changes to 
the initial conditions contributing to accident severity or 
consequences. The proposed changes have no direct effect on any 
plant systems.

    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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 0037.
    NRC Section Chief: Michael T. Masnik.

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

    Date of amendment request: October 16, 2000.
    Description of amendment request: The proposed amendment would 
incorporate new pressure and temperature (P/T) curves into the 
Technical Specifications. The reactor pressure vessel P/T limit curves 
would be updated for inservice leakage and hydrostatic testing, non-
nuclear heatup and cooldown, and criticality.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The P/T [pressure and temperature] limits are 
not derived from Design Basis Accident (DBA) analyses. They are 
prescribed by the ASME [American Society of Mechanical Engineers] 
Code and 10 CFR 50 Appendix G and H as restrictions on operation to 
avoid encountering pressure, temperature, and temperature rate of 
change conditions that might cause undetected flaws to propagate and 
cause non-ductile failure of the reactor coolant pressure boundary.
    The changes to the calculational methodology for the P/T limits 
based upon Code Case N-640 continue to provide adequate margin in 
the prevention of a non-ductile type fracture of the reactor 
pressure vessel (RPV). The Code Case was developed based upon the 
knowledge gained through years of industry experience. P/T curves 
developed using the allowances of Code Case N-640 indeed yield more 
operating margin. However, the experience gained in the areas of 
fracture toughness of materials and pre-existing undetected defects 
shows that some of the existing assumptions used for the calculation 
of P/T limits are unnecessarily conservative and unrealistic. 
Therefore, providing the allowances of the Code Case in developing 
the P/T limit curves will continue to provide adequate protection 
against non-ductile type fractures of the RPV.
    The proposed change will not affect any other system or piece of 
equipment designed for the prevention or mitigation of previously 
analyzed events. The change does not adversely affect the integrity 
of the reactor coolant system such that its function in control of 
radiological consequences is affected.
    (2) The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The amendment will revise the P/T curves which are 
established to the requirements of 10 CFR 50, Appendix G to assure 
that non-ductile fracture of the reactor vessel is prevented.
    The proposed change provides more operating margin in the P/T 
limit curves for inservice leakage and hydrostatic pressure testing, 
non-nuclear heatup and cooldown, and criticality, with benefits 
being primarily realized during the pressure tests. The proposed 
change does not result in any new or unanalyzed operation of any 
system or piece of equipment important to safety, and as a result, 
the possibility of a new type event is not created.
    (3) The proposed amendment will not involve a significant 
reduction in a margin of safety. 10 CFR 50, Appendix G specifies 
fracture toughness requirements to provide adequate margins of 
safety during operation over the service lifetime. The values of 
adjusted reference temperature and upper shelf energy are expected 
to remain within the limits of Regulatory Guide 1.99, Revision 2 and 
Appendix G of 10 CFR 50 (less than 200 degrees F and greater than 50 
ft-lbs respectively) for at least 32 effective full power years 
(EFPY) of operation.
    The proposed change reflects an update of P/T curves based on 
the latest ASME guidance. The revised P/T curves provide more 
operating margin and thus, more operational flexibility than the 
current P/T curves. With the increased operational margin, a 
reduction in the safety margin results with respect to the existing 
curves. However, industry experience since the inception of the P/T 
limits in 1974 confirms that some of the existing methodologies used 
to develop P/T curves are unrealistic and unnecessarily 
conservative. Accordingly, ASME Code Case N-640 takes into account 
the acquired knowledge and establishes more realistic methodologies 
for the development of P/T curves. Therefore, operational 
flexibility is gained and an acceptable margin of safety to RPV non-
ductile type fracture is maintained.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Al Gutterman, Morgan, Lewis & Bockius, 1800 
M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Claudia M. Craig.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: November 29, 1999, as supplemented on 
November 10, 2000.
    Description of amendment request: The licensee submitted a proposed 
amendment to Kewaunee Nuclear Plant's Technical Specifications (TSs) 
modifying the TSs to incorporate requested changes per Generic Letter 
99-02, ``Laboratory Testing of Nuclear-Grade Activated Charcoal,'' 
dated June 3, 1999.

[[Page 77922]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The Shield Building Ventilation, the Auxiliary Building 
Ventilation, the Spent Fuel Pool Sweep Systems and the Control Room 
Post Accident Recirculation System are not accident initiators. 
Therefore, the proposed change will not increase the probability of 
an accident. The purpose of each of these systems is to mitigate the 
consequences of an accident once it has occurred. Based upon a 
comparison, the later version of ASTM D3803, ASTM D3803-89 was found 
to test the efficiency of the charcoal material under more 
conservative conditions. By testing the charcoal absorber material 
under more conservative conditions, the charcoal will require 
replenishment sooner. Therefore, the consequences will not be 
increased.
    The changes to the basis sections are to promote clarity and 
uniformity. These statements were previously contained in the basis 
section or clarify which revision of Regulatory Guide 1.52 that 
should be used. This change provides acceptable guidelines for the 
qualification of replacement charcoal absorbent. Therefore, these 
changes will not increase the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    This amendment request does not change any component at the 
plant. It is changing the testing requirements for material already 
installed. The material being tested has not changed. By testing the 
charcoal material under this revised protocol the material will be 
replaced with fresh charcoal sooner. This will ensure the equipment 
performs as described in the USAR.
    The changes to the basis sections are to promote clarity and 
uniformity. These statements were previously contained in the basis 
section or clarify which revision of Regulatory Guide 1.52 that 
should be used. This change provides acceptable guidelines for the 
qualification of replacement charcoal adsorbent. Therefore, these 
changes will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    There is no reduction in the margin of safety. The efficiency of 
the charcoal material assumed by the USAR will not change as a 
result of this amendment and the functioning of the system will not 
change. Therefore, the original margin of safety is maintained.
    The changes to the basis sections are to promote clarity and 
uniformity. These statements were previously contained in the basis 
section or clarify which revision of Regulatory Guide 1.52 that 
should be used. This change provides acceptable guidelines for the 
qualification of replacement charcoal adsorbent. Therefore, these 
changes will not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: Claudia M. Craig.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: October 12, 2000.
    Description of amendment request: In accordance with 10 CFR 50.59, 
the topical report WPSRSEM-NP, ``Reload Safety Evaluation Methods for 
Application to Kewaunee,'' Revision 3, is being submitted for the 
staff's review and approval since the licensee determined the revision 
of the report involved an unreviewed safety question. The topical 
report is intended to be applicable to Kewaunee reload cycles after and 
including Cycle 25, presently scheduled to commence in the fall of 
2001. The topical report reflects:
     Editorial changes, including corrections to the limiting 
directions of core physics parameters and clarification of the 
definition of core physics parameters.
     Changes made to incorporate the CONTEMPT code for 
containment analysis. CONTEMPT is currently described for this purpose 
in the Kewaunee updated safety analysis report (USAR).
     The adoption of the GOTHIC code for containment analysis.
     Changes in Reload Safety Evaluation Methods due to Large 
Break Loss-of-Coolant Accident Upper Plenum Injection Analysis.
     The adoption of RETRAN-3D for use in the 2D mode for 
system analysis.
     The extension of the VIPRE-01 code to reflect changes in 
fuel design.
    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.
    Analysis methods are not accident initiators, therefore, changes 
in analysis methods will not increase significantly the probability 
of occurrence or the consequences of an accident previously 
evaluated.
    The changed analysis methods are conservative and conform to 
industry standards for analysis methods that are applied to design 
basis safety analyses. Benchmark analyses have demonstrated good 
agreement between the changed analysis methods and the current 
analysis of record (AOR) methods. The safety analysis results using 
the changed analysis methods are shown to satisfy all applicable 
design and safety analysis acceptance criteria. The demonstrated 
adherence to safety analysis acceptance criteria precludes new 
challenges to components and systems that could adversely affect the 
ability of existing components and systems to mitigate the 
consequences of any accident or adversely affect the integrity of 
any fission product barrier.
    Analysis methods changes will not impact plant equipment 
important to safety. Equipment important to safety will continue to 
operate within its design capabilities. The analysis methods changes 
also do not affect the plant configuration or the overall plant 
performance capabilities.
    Therefore, the changes will not increase probability or the 
consequences of an accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed change is a change to the analysis methods, which 
are applied to Kewaunee. Analysis methods are not accident 
initiators. The changed analysis methods are applied to the 
accidents that are the established design basis accidents for 
Kewaunee. Analysis methods changes will not impact plant equipment 
important to safety. Equipment important to safety will continue to 
operate within its design capabilities. The analysis methods changes 
also do not affect the plant configuration or the overall plant 
performance capabilities.
    As demonstrated by the benchmark reports the methodologies 
provide a more accurate but still conservative representation of 
expected plant response following a design basis accident. Since the 
new methodologies are conservative with respect to actual expected 
plant response the changes will not create the possibility of an 
accident of a different type than any previously evaluated.
    (3) Involve a significant reduction in the margin of safety.
    The proposed changes are changes to the analysis methods, which 
are applied to Kewaunee design basis safety analyses. The revised 
analysis methods have been verified through benchmark analyses 
against the current Analysis of Record methods. The analysis methods 
are conservative and appropriate for application to Kewaunee design 
basis analyses. Safety analysis acceptance criteria are satisfied 
when the changed analysis methods are applied to the Kewaunee design 
basis safety analyses. Demonstrated adherence to safety analysis 
acceptance criteria using the new analysis methods assures that 
Technical Specification limits will be satisfied during operation 
with the changed analysis methods.

[[Page 77923]]

    Therefore, the margin of safety as defined in the basis of any 
Technical Specification will not be reduced significantly because of 
these changes.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: Claudia M. Craig.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: November 10, 2000.
    Description of amendment request: The proposed amendment is to 
revise several sections of the Kewaunee Nuclear Power Plant (KNPP) 
Technical Specifications (TSs). These sections include administrative 
changes, Table 4.1-1, and Sections 1.0, 6.4, and 6.10.
    Administrative changes are submitted with this proposed amendment 
to correct minor typographical errors in the Table of Contents and 
among these changes are renumbering the index section pages and the 
addition of previously omitted sections.
    The proposed changes will modify TS Table 4.1-1, ``Minimum 
Frequencies for Checks, Calibrations and Test of Instrument Channels.'' 
This proposed change will decrease the calibration frequency for 
Turbine First Stage Pressure to support KNPP's 18-month operating 
cycle, and modify the table to eliminate a note that could lead to non-
conservative calibration frequency.
    The proposed TS Section 1.0, ``Definitions,'' will incorporate a 
line item improvement to provide additional clarification on channel 
calibration.
    The proposed TS Section 6.4, ``Training,'' will remove the title of 
director for the KNPP training program and relocate the title reference 
to the Operational Quality Assurance Program Description (OQAPD).
    The proposed TS Section 6.10, ``Record Retention,'' will revise the 
off-site review committee title.
    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.

Table of Contents

    The proposed changes are administrative in nature and, 
therefore, have no impact on accident initiators or plant equipment, 
and thus do not affect the probability or consequences of an 
accident.

TS Section 1.0, ``Definitions''

    A calibration will continue to ensure that a channel is within 
specification. Furthermore, calibration methodology is not an 
accident initiator. Therefore, the proposed change will not 
significantly raise the probability or consequences of an accident 
previously evaluated.

TS Table 4.1-1

    The proposed change amends the calibration interval of the 
turbine first stage pressure from 12 months to each refueling cycle 
to coincide with KNPP's operating cycle. Calibration frequency would 
not change the consequence of a failure of the first stage pressure 
channel. Calibration frequency is not an accident initiator. 
Therefore, the proposed changes will not significantly raise the 
probability or consequences of an accident previously evaluated. 
Additionally, this change is consistent with the turbine first stage 
pressure calibration frequency stated in STS.
    The proposed changes to the identified line items in Table 4.1-1 
will require calibration of the instruments on a refueling cycle 
interval without exception. These calibration frequencies are not 
accident initiators and thus do not affect the probability of an 
accident. These changes are more conservative than existing TS and, 
therefore, will not increase the consequences of an accident.

TS Section 6.4, ``Training''

    The proposed change will not change the intent of the TS. 
Removing the title from the TS is administrative in nature and, 
therefore, has no impact on accident initiators or plant equipment, 
and thus does not affect the probability or consequences of an 
accident.

TS Section 6.10, ``Record Retention''

    The proposed change will not change the intent of the TS. 
Changing the title of the off-site review committee is 
administrative in nature and, therefore, has no impact on accident 
initiators or plant equipment, and thus does not affect the 
probability or consequences of an accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.

Table of Contents

    The proposed changes do not involve changes to the physical 
plant or operations. Since these administrative changes do not 
contribute to accident initiation, they do not produce a new 
accident scenario or produce a new type of equipment malfunction. 
Also, these changes do not alter any existing accident scenarios; 
they do not affect equipment or its operation, and thus, do not 
create the possibility of a new or different kind of accident.

TS Section 1.0, ``Definitions''

    The proposed TS change to channel calibration will not introduce 
any new equipment or result in existing equipment functioning 
differently from that previously evaluated in the USAR or TS. 
Calibration will continue to ensure that the channel is within 
specification and capable of performing its design basis function. 
No new accident is introduced and no safety-related equipment or 
safety functions are altered. Therefore, the proposed change does 
not affect any of the parameters or conditions that contribute to 
initiation of any accident.

TS Table 4.1-1

    The proposed TS change will not introduce any new equipment or 
result in existing equipment functioning differently from that 
previously evaluated in the USAR or TS. The proposed change amends 
the calibration interval of the turbine first stage pressure from 12 
months to each refueling cycle to coincide with KNPP's operating 
cycle. Performing the surveillance during refueling will decrease 
the likelihood for an induced transient. Expanding the calibration 
frequency will not affect the performance of the first stage 
pressure channel. A review of turbine first stage pressure 
calibration results for the last three years concluded no adjustment 
of the instrument was necessary due to little or no drift. 
Furthermore, similar transmitters already calibrated on a refueling 
basis have remained within acceptable limits. These results indicate 
stable instrument performance to support extending calibration 
frequency from 12 months to each refueling cycle.
    The proposed changes will ensure that the affected channels are 
calibrated on a refueling basis. These changes will not introduce 
any new equipment or result in existing equipment functioning 
differently from that previously evaluated in the USAR or TS. No new 
accident is introduced and no safety-related equipment or safety 
functions are altered. The proposed changes do not affect any of the 
parameters or conditions that contribute to initiation of any 
accident.

TS Section 6.4, ``Training''

    The proposed change does not involve a change to the physical 
plant or operations. Since an administrative change does not 
contribute to accident initiation, it does not produce a new 
accident scenario or produce a new type of equipment malfunction.

TS Section 6.10, ``Record Retention''

    The proposed change does not involve a change to the physical 
plant or operations. Since an administrative change does not 
contribute to accident initiation, it does not produce a new 
accident scenario or produce a new type of equipment malfunction.
    3. Involve a significant reduction in the margin of safety.

Table of Contents

    Administrative changes do not involve a significant reduction in 
the margin of safety. The proposed changes do not affect plant 
equipment or operation. Safety limits and limiting safety system 
settings are not affected by these changes.

[[Page 77924]]

TS Section 1.0, ``Definitions''

    Operation of the facility in accordance with this proposed TS 
change would not involve a significant reduction in a margin of 
safety. The specification will still ensure the operability of 
channels requiring calibration.

TS Table 4.1-1

    Operation of the facility in accordance with the proposed TS 
changes would not involve a significant reduction in the margin of 
safety. The calibration will continue to verify the operability of 
the turbine first stage pressure channels. Therefore, the proposed 
change does not involve a significant reduction in the margin of 
safety.
    Operation of the facility in accordance with the proposed TS 
changes would not involve a significant reduction in a margin of 
safety. The proposed changes will ensure the continued reliability 
of the instruments. This change is more conservative than existing 
TS and is consistent with STS.

TS Section 6.4, ``Training''

    Administrative changes do not involve a significant reduction in 
the margin of safety. The proposed change does not affect plant 
equipment or operation. Safety limits and limiting safety system 
settings are not affected by this change.

TS Section 6.10, ``Record Retention''

    Administrative changes do not involve a significant reduction in 
the margin of safety. The proposed change does not affect plant 
equipment or operation. Safety limits and limiting safety system 
settings are not affected by this 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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: Claudia M. Craig.

PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric 
Station, Unit 2, Luzerne County, Pennsylvania

    Date of amendment request: March 20, 2000.
    Description of amendment request: The proposed amendment would 
revise the Susquehanna Steam Electric Station (SSES), Unit 2, Technical 
Specification 2.1.1.2, minimum critical power ratio (MCPR) safety 
limits. These safety limits are being revised to reflect planned 
changes to the core composition for the next operating cycle and to 
support a separate license amendment proposing an increase in the SSES, 
Unit 1 and 2, rated thermal power.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes in MCPR Safety Limits do not affect any 
plant system or component (except the reactor core) and therefore 
does not increase the probability of an accident previously 
evaluated.
    A Unit 2 Cycle 11 MCPR Safety Limit analysis was performed for 
PPL by SPC [Siemens Power Corporation]. This analysis used NRC 
approved methods as required by SSES Technical Specifications. For 
Unit 2 Cycle 11 [U2C11], the critical power performance of the 
ATRIUMTM-10 fuel was determined using the NRC approved 
ANFB-10 correlation. Also, the analysis for U2C11 supports a Core 
Thermal Power of 3493 MWt which is a 1.5% increase over U2C10 (3441 
MWt). The Safety Limit MCPR calculations statistically combine 
uncertainties on feedwater flow, feedwater temperature, core flow, 
core pressure, core power distribution, and uncertainties in the 
Critical Power Correlation. The SPC analysis used cycle specific 
power distributions and calculated MCPR values such that at least 
99.9% of the fuel rods are expected to avoid boiling transition 
during normal operation or anticipated operational occurrences. The 
resulting two-loop and single-loop MCPR Safety Limits are included 
in the proposed Technical Specification change. Thus, the cladding 
integrity and its ability to contain fission products are not 
adversely affected. It is therefore concluded that the proposed 
change does not increase the 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 changes to the Unit 2 Technical 
Specifications (MCPR Safety Limits) do not affect any plant system 
or component and do not affect plant operation. The consequences of 
transients and accidents will remain within the criteria approved by 
the NRC. 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 a margin of safety.
    Since the proposed changes do not affect any plant system or 
component, and do not have any impact on plant operation, the 
proposed changes will not affect the function or operation of any 
plant system or component. The consequences of transients and 
accidents will remain within the criteria approved by the NRC. The 
proposed MCPR Safety Limits do not involve a significant reduction 
in the margin of safety as currently defined in the bases of the 
applicable Technical Specification sections. 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 No. 50-206, San 
Onofre Nuclear Generating Station, Unit 1, San Diego County, California

    Date of amendment requests: October 30, 2000-PCN 268.
    Description of amendment requests: This amendment application 
requests to delete license condition 2.C(3) related to fuel 
transshipments between San Onofre Nuclear Generating Station, Unit 1 
(SONGS 1), which is in the process of decommissioning, and SONGS Units 
2 or 3 since such transshipments will no longer be made. In addition, 
the amendment application requests revisions to the Unit 1 defueled 
Technical Specifications to (1) remove the spent fuel pool (SFP) 
temperature limits and related cooling system operability requirements, 
(2) remove the SFP auxiliary feedwater storage tank makeup water 
requirements and related surveillance requirements, (3) change the SFP 
water level limit for conditions other than spent fuel movement, and 
(4) change the operator staffing requirements for the decommissioning 
control room. As a result of these proposed changes, the licensee also 
proposes to delete the definitions of FUNCTIONAL and SPENT FUEL POOL 
COOLING TRAIN and revise the table of contents and list of tables 
according to the above changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. The proposed change is a request to revise the San Onofre 
Nuclear Generating Station, Unit 1 (SONGS 1) license and permanently 
defueled technical specifications. The license condition for 
transshipment is being deleted since there is no safety-related 
equipment to protect and no plans for transshipment of SONGS 1 fuel 
to SONGS 2 or 3. Since the purpose of removing this license 
condition is that this activity will

[[Page 77925]]

no longer be performed, there is no impact on accident probability 
or consequences. Deleting the technical specifications for spent 
fuel pool temperature and makeup are based on the current benign 
status of the spent fuel and spent fuel pool. The requirements and 
surveillances provided by these technical specifications no longer 
provide appropriate limits for the safe storage of the spent fuel. 
The spent fuel temperature limit cannot be reached. Makeup water is 
available from various sources onsite and offsite in a timely 
manner. Deleting these technical specifications has no impact on the 
probability or consequences of an accident. Modifying the spent fuel 
pool water level requirements provides two levels for maintaining 
water: One water level (elevation 28' [feet]) for just storage and a 
higher water level (elevation 40' 3" [inches]) for fuel movement. 
Lowering the water level for storage of spent fuel does not affect 
the accident probability. The fuel handling accident will not occur 
when the pool water level is at elevation 28 feet since spent fuel 
will only be handled when the pool water level is at elevation 40' 
feet 3". Removing the restrictions for having one individual of the 
minimum shift crew located in the control room will not have any 
impact on the fuel handling accident since a certified fuel handler 
is still required to be present.
    Therefore, this change does not involve an increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different type of accident 
from any accident previously evaluated?
    No. This proposed change is a request to revise the SONGS 1 
license and permanently defueled technical specifications. The 
transshipment license condition is being deleted since there is no 
safety-related equipment to protect and no plans for transshipment 
of Unit 1 spent fuel to Units 2 or 3. The technical specifications 
for spent fuel pool temperature and makeup are being deleted since 
these requirements no longer provide limits appropriate for 
maintaining the spent fuel pool. Removing these requirements does 
not create the possibility for a new or different accident since the 
associated limits are no longer attainable by the spent fuel pool. 
The only potential accident remaining is the spent fuel handling 
accident. Lowering the level of the spent fuel pool to elevation 28 
feet has no impact on accident initiations since fuel handling will 
not be allowed at this water level. Removing the restrictions in the 
location of the minimum shift crew has no impact on accident 
initiation, and the certified fuel handler will be present during 
fuel handling operations.
    Therefore, this change does not involve the possibility of a new 
or different type of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety?
    No. This proposed change is a request to delete requirements 
from the license and the technical specifications and modify the 
spent fuel pool level requirements. Deleting the transshipment 
license condition has no impact on margin since there no longer is 
any safety-related equipment to protect and there are no plans for 
transshipment of Unit 1 spent fuel to Units 2 or 3. Deleting the 
spent fuel pool temperature and makeup requirements has no effect on 
margin since the status of the spent fuel pool is such that the 
margins associated with these requirements have increased and with 
time will continue to increase. Modifying the level requirement to 
allow the water level to be at elevation 28 feet for spent fuel 
storage has no impact on margin since the spent fuel has cooled 
significantly and fuel movement will not occur at this level. Since 
the status of the spent fuel pool is such that the margins are 
improving with time, removing the restrictions in the location of 
the minimum shift crew has no effect on the margins.

    Therefore, this change does not involve a reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. The staff also reviewed the proposed administrative changes 
to delete definitions and conform the table of contents and list of 
tables to the proposed changes for no significant hazards 
consideration. These administrative changes do not affect the design or 
operation of the facility and satisfy the three standards of 10 CFR 
50.92(c). Therefore, the NRC staff proposes to determine that the 
amendment request involves 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: Michael Masnik (Unit 1).

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

    Date of application for amendments: October 5, 2000.
    Brief description of amendments: The amendments revise the VEGP 
Updated Final Safety Analysis Report (UFSAR) Chapters 11 and 15 to 
incorporate changes due to an updated Dose Equivalent Iodine analysis. 
The new analysis was performed in response to Westinghouse Nuclear 
Safety Advisory Letter, ``NSAL-00-04: Nonconservatisms in Iodine 
Spiking Calculations.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes do not significantly increase the 
probability or consequences of an accident previously evaluated in 
the [Updated Final Safety Analysis Report] UFSAR. The comprehensive 
engineering review included evaluations or reanalysis of all 
accident analyses. The letdown flow rate does not initiate any 
accident; therefore, the probability of an accident has not been 
increased. All dose consequences have been analyzed or evaluated 
with respect to the proposed changes, and all acceptance criteria 
continue to be met. Therefore, these changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated?
    The proposed changes do not create the possibility of a new or 
different kind of accident than any accident already evaluated in 
the UFSAR. No new accident scenarios, failure mechanisms or limiting 
single failures are introduced as a result of the proposed changes. 
The changes have no adverse effects on any safety-related system and 
do not challenge the performance or integrity of any safety-related 
system. Therefore, all accident analyses criteria continue to be 
met, and these 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 proposed changes do not involve a significant reduction in a 
margin of safety. All analyses and evaluations using these inputs 
have been revised to reflect the proposed values. The evaluations 
and analyses results demonstrate that applicable acceptance criteria 
are met. Therefore, the proposed changes do not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: Richard L. Emch, Jr.

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 amendment request: November 3, 2000.

[[Page 77926]]

    Description of amendment request: The proposed amendments would 
revise Technical Specification 5.5.11, ``Technical Specification Bases 
Control Program,'' to provide consistency with the changes to 10 CFR 
50.59 which were published in the Federal Register (64 FR 53582) on 
October 4, 1999.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change deletes the reference to unreviewed safety 
question as defined in 10 CFR 50.59. Deletion of the definition of 
unreviewed safety question was approved by the NRC with the revision 
of 10 CFR 50.59. Consequently, the probability of an accident 
previously evaluated is not significantly increased. Changes to the 
TS Bases are still evaluated in accordance with 10 CFR 50.59. As a 
result, the consequences of any accident previously evaluated are 
not significantly affected. Therefore, this change does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously analyzed?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. 
Therefore, this 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 proposed change will not reduce a margin of safety because 
it has no direct effect on any safety analyses assumptions. Changes 
to the TS Bases that result in meeting the critieria in paragraph 10 
CFR 50.59(c)(2) will still require NRC approval pursuant to 10 CFR 
50.59. This change is administrative in nature based on the revision 
to 10 CFR 50.59. 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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of application for amendments: November 6, 2000.
    Brief description of amendments: The request revises the VEGP 
Technical Specification (TS) Limiting Conditions for Operation 3.7.10, 
3.7.11, and 3.7.13 to address degraded pressure boundaries. The changes 
revise the TS to allow the pressure boundaries of ventilation systems 
such as the Control Room Emergency Filtration System (CREFS) and the 
Piping Penetration Area Filtration and Exhaust System (PPAFES) to be 
opened intermittently under administrative controls. A new condition is 
also added that allows 24 hours to restore inoperable CREFS and PPAFES 
pressure boundaries before requiring the units to perform an orderly 
shutdown.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The control room emergency filtration system (CREFS) and the 
piping penetration area filtration and exhaust system (PPAFES) are 
not assumed to be initiators of any analyzed accident. Therefore, 
the proposed changes do not affect the probability of any accident 
previously evaluated. The proposed changes for the CREFS and PPAFES 
Technical Specifications (TS) would permit the subject pressure 
boundaries to be opened intermittently under administrative control. 
Based on the proposed compensatory measures in the form of a 
dedicated individual who is in communication with the control room, 
and his ability to rapidly restore the pressure boundary, the 
capability to mitigate a design basis event will be maintained. In 
addition, the proposed changes would add a new condition that would 
permit a 24-hour period to take action to restore an inoperable 
pressure boundary to operable status, modify existing conditions to 
accommodate the new condition (so as to maintain the requirements of 
the existing conditions), and correct a typographical error. With 
respect to CREFS, the proposed changes do not involve a significant 
increase in the consequences of an accident previously evaluated 
based on the availability of a self-contained breathing apparatus to 
minimize radiological dose due to iodine and the ability to operate 
more than one train as the need arises to maintain positive pressure 
or at least maintain an outflow of air from the control room 
environment. With respect to the PPAFES, it has been demonstrated by 
analysis that a breach of the pressure boundary will not result in 
control room or offsite doses that exceed their respective limits. 
The correction of the typographical error is an administrative 
change that has no technical impact.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    No. The proposed changes for the CREFS and PPAFES TS would 
permit the subject pressure boundaries to be opened intermittently 
under administrative control. In addition, the proposed changes 
would add a new condition that would permit a 24-hour period to take 
action to restore an inoperable pressure boundary to operable 
status, modify existing conditions to accommodate the new condition 
(so as to maintain the requirements of the existing conditions), and 
correct a typographical error. The proposed changes do not alter the 
operation of the plant or any of its equipment, introduce any new 
equipment, or result in any new failure mechanisms or single 
failures. Therefore, there is no potential for a new accident and no 
changes to the way that an analyzed accident will progress. The 
correction of the typographical error is an administrative change 
that has no technical impact.
    3. Do the proposed changes result in a significant reduction in 
a margin of safety?
    No. The proposed changes for the CREFS and PPAFES TS would 
permit the subject pressure boundaries to be opened intermittently 
under administrative control. In addition, the proposed changes 
would add a new condition that would permit a 24-hour period to take 
action to restore an inoperable pressure boundary to operable 
status, modify existing conditions to accommodate the new condition 
(so as to maintain the requirements of the existing conditions), and 
correct a typographical error. The proposed changes do not adversely 
affect the ability of the fission product barriers to perform their 
functions. The only safety-related equipment affected by the 
proposed changes is the CREFS and the PPAFES. It has been 
demonstrated by analysis that a breach in the pressure boundary of 
the PPAFES will not cause the control room or offsite doses to 
exceed their respective limits. Adequate compensatory measures are 
available to mitigate a breach in the CREFS pressure boundary. The 
probabilities of design bases accidents that would place demands on 
these systems during a period that the ventilation system pressure 
boundaries would be allowed to be inoperable have been shown to be 
negligible. In addition, the proposed changes avoid the potential of 
placing one or both units in TS Limiting Condition for Operation 
(LCO) 3.0.3 solely due to a breach of the ventilation system 
pressure boundary. The correction of the typographical error is an 
administrative change that has no technical impact.

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

[[Page 77927]]

involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of application for amendments: November 16, 2000.
    Brief description of amendments: The request proposes to amend 
Technical Specification 5.5.1, `` Technical Specification Bases Control 
Program'' to provide consistency with the changes to 10 CFR 50.59 as 
published in the Federal Register (64 FR 53582) dated October 4, 1999.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change deletes the reference to unreviewed safety 
question as defined in 10 CFR 50.59. Deletion of the definition of 
unreviewed safety question was approved by the NRC with the revision of 
10 CFR 50.59. Consequently, the probability of an accident previously 
evaluated is not significantly increased. Changes to the TS Bases are 
still evaluated in accordance with 10 CFR 50.59. As a result, the 
consequences of any accident previously evaluated are not significantly 
affected. Therefore, this change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously analyzed?
    The proposed change does not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. Therefore, this 
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 proposed change will not reduce a margin of safety because it 
has no direct effect on any safety analyses assumptions. Changes to the 
TS Bases that result in meeting the criteria in paragraph 10 CFR 50.59 
(c)(2) will still require NRC approval pursuant to 10 CFR 50.59. This 
change is administrative in nature based on the revision to 10 CFR 
50.59. 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: Richard L. Emch, Jr.

Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear 
Plant, Unit 2, Limestone County, Alabama

    Date of amendment request: November 21, 2000 (TSC-396).
    Description of amendment request: The proposed amendment would 
revise the reactor core Safety Limit Minimum Critical Power Ratio 
(SLMCPR) specified in Technical Specification (TS) Section 2.1.1.2 from 
1.10 to 1.07 for two reactor recirculation loop operation and from 1.12 
to 1.10 for single loop operation. The change is based on use of newly 
approved analytical methodology for the Cycle 12 reload analysis. This 
methodology is described in Global Nuclear Fuels (GNF) licensing 
document, ``General Electric Standard Application for Reactor Fuel, 
GESTAR-II, Amendment 25,'' dated June 2000, which has been approved by 
NRC.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    A. The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The proposed amendment establishes revised SLMCPR values for two 
recirculation loop operation and for single recirculation loop 
operation. The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
proposed SLMCPRs preserve the existing margin to transition boiling and 
the probability of fuel damage is not increased. Since the change does 
not require any physical plant modifications or physically affect any 
plant components, no individual precursors of an accident are affected 
and the probability of an evaluated accident is not increased by 
revising the SLMCPR values.
    The consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those consequences. 
The revised SLMCPRs have been performed using NRC-approved methods and 
procedures. The basis of the MCPR [minimum critical power ratio] Safety 
Limit is to ensure no mechanistic fuel damage is calculated to occur if 
the limit is not violated. These calculations do not change the method 
of operating the plant and have no effect on the consequences of an 
evaluated accident. Therefore, the proposed TS change does not involve 
an increase in the probability or consequences of an accident 
previously evaluated.
    B. The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed license amendment involves a revision of the SLMCPR 
for two recirculation loop operation and for single loop operation 
based on the results of an analysis of the Cycle 12 core. Creation of 
the possibility of a new or different kind of accident would require 
the creation of one or more new precursors of that accident. New 
accident precursors may be created by modifications of the plant 
configuration, including changes in the allowable methods of operating 
the facility. This proposed license amendment does not involve any 
modifications of the plant configuration or changes in the allowable 
methods of operation. Therefore, the proposed TS change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    C. The proposed amendment does not involve a significant reduction 
in a margin of safety.
    The margin of safety as defined in the TS bases will remain the 
same. The new SLMCPRs are calculated using NRC-approved methods and 
procedures which are in accordance with the current fuel design and 
licensing criteria. The SLMCPRs remain high enough to ensure that 
greater than 99.9% of all fuel rods in the core are

[[Page 77928]]

expected to avoid transition boiling if the limit is not violated, 
thereby preserving the fuel cladding integrity. Therefore, the proposed 
TS changes do not involve a reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
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: October 25, 2000.
    Description of amendment request: The proposed change would revise 
the 125 volt DC (Vdc) station battery system Technical Specifications 
(TSs) to reflect the availability of a second, fully qualified charger, 
for each main station battery system. The licensee also proposed 
corresponding changes to the listing of components in the TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. 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.
    There is no change in the method of operation of the 125 Vdc 
main station battery systems by this change. The battery chargers 
will function the same, except that an additional battery charger 
will be available to each system. No change to accident assumptions 
or precursors are involved with this change. Likewise, no change in 
system operation or response to analyzed events is affected.
    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 new chargers to be installed will provide additional 
charging capability. No reduction in DC system equipment operation 
or capability is involved. The methods by which the DC systems 
perform their safety functions are unchanged and remain consistent 
with current safety analysis assumptions. There is no change in 
system or 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.
    No adverse affect on equipment operation or capability will 
result from this change. The installation of additional chargers in 
fact enhances the reliability of the battery charging function. The 
equipment fed by the DC systems involved in this change will 
continue to provide adequate power to safety related loads in 
accordance with analysis assumptions.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: November 1, 2000.
    Description of amendment request: The proposed change would revise 
the operability requirement for high pressure coolant injection (HPCI) 
and reactor core isolation cooling (RCIC) low steam line pressure 
isolation instrumentation to coincide with system operability 
requirements. The proposed change eliminates the need to open manual 
containment isolation valves under administrative control during 
reactor heatup, reduces the potential for operator error when closing 
these valves (potential for leaving valve mispositioned) and clarifies 
the steam line low pressure isolation function description. An 
administrative change to correct the HPCI High Steam Line d/p 
instrument component numbers was also proposed to ensure the accuracy 
of isolation instrumentation information.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The operation of the Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change clarifies the equipment protection purpose 
of the HPCI and RCIC low steam line isolation function. [The 
proposed change would require] operability of the steam supply 
pressure instrumentation [ ] whenever the systems are required to be 
operable. This change does not significantly alter the function of 
containment isolation actuation instruments nor does it 
significantly alter containment integrity requirements. The proposed 
change does not alter the basic operation of process variables, 
systems, or components as described in the safety analysis. No new 
equipment is being introduced.
    The proposed change does not affect the ability of the primary 
containment isolation system or high pressure core cooling systems 
to perform their safety functions. The essential safety function of 
providing primary containment integrity is maintained since 
operability of the primary instrumentation associated with detection 
of a HPCI or RCIC steam line break outside containment will continue 
to be required when primary containment integrity is required. The 
essential safety function of providing water to cool the core in the 
event of a small break in the nuclear system is maintained. The 
operational change being made would not 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, initiators 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. No new or different types of 
equipment will be installed. Operation with the HPCI and RCIC steam 
line isolation valves open between 212  deg.F and 150 psig does not 
alter the input or result of existing accident analyses. The change 
in plant operation does not involve 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 77929]]

    The change involves operation with the HPCI and RCIC systems 
with steam line isolation valves open between 212  deg.F and 150 
psig. 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 maintains design margins of the primary 
containment isolation system or high pressure core cooling 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 a small break in the nuclear 
system are maintained. There is no physical or operational change 
being made which would alter the sequence of events, plant response, 
or margins in 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. The change in method of operation results in no significant 
impact on safety functions or assumed responses. The proposed change 
does not alter the means by which primary containment isolation is 
maintained and high pressure core cooling systems are operated.
    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: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: September 27, 2000, as supplemented 
November 21, 2000.
    Description of amendment request: The proposed changes will 
increase the fuel enrichment limit from 4.3 weight percent to 4.6 
weight percent Uranium\235\ (U\235\), establish Technical 
Specifications to control the boron concentration in the spent fuel 
pool (SFP) and impose restrictions on the storage locations for some 
spent fuel assemblies, and change the method of criticality calculation 
used to evaluate the effect of a fuel enrichment change on the SFP.
    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.] Criterion 1. The proposed increase in maximum fuel 
enrichment and the changes to the SFP design basis will not 
significantly increase the probability of or consequences of an 
accident previously evaluated in the North Anna Units 1 and 2 UFSAR 
[Updated Final Safety Analysis Report].
    The only accidents for which the probability of occurrence is 
potentially affected by the fuel enrichment and SFP changes involve 
criticality events during fuel handling and storage (e.g., fuel 
mispositioning). The proposed Technical Specifications establish 
additional restrictions on the placement of each fuel assembly in 
the SFP to ensure subcriticality. However, criticality safety 
analyses have been performed that demonstrate that the 
Keff during the handling and storage of both new and 
spent fuel remains low enough to ensure subcriticality during 
postulated accident conditions. In addition, analyses of the 
dilution of the spent fuel pool have been performed to ensure that 
there is adequate time for a dilution event to be detected and 
mitigated, such that the required subcritical margin is maintained 
in the spent fuel pool. Therefore the probability of occurrence of 
criticality during fuel handling or storage is not significantly 
increased. In addition the consequences of the operating reactor 
accident scenarios are also unchanged, because the source terms used 
to determine the releases from fuel during accidents are a function 
of burnup, rather than initial enrichment.
    [2.] Criterion 2. The proposed increase in maximum fuel 
enrichment or the change in the SFP design basis does not create a 
new or different kind of accident from any already discussed in the 
North Anna Units 1 and 2 UFSAR.
    Although there are new restrictions on placement of fuel in the 
SFP, the administrative controls on fuel movement to specified 
locations in the pool are unchanged. The higher enrichment fuel and 
the new Technical Specifications for the spent fuel pool do not 
require any new or different plant equipment, and do not change the 
manner in which currently installed equipment is operated. There are 
no changes to normal core operation, and the units will meet all 
applicable design criteria and will operate within existing 
Technical Specifications limits. No new failure modes have been 
created for any system, component, or piece of equipment, and no new 
single failure mechanisms have been introduced. No new or different 
plant equipment is introduced, and the operation of currently 
equipment is not changed. The use of a higher maximum fuel 
enrichment will not cause the design criteria for fuel operation or 
storage to be exceeded. No new modes or limiting single failures are 
created by the use of a higher fuel enrichment. Safety analyses for 
the fuel storage area have demonstrated that subcriticality will be 
maintained during fuel handling and storage, including fuel 
mispositioning and pool dilution scenarios.
    [3.] Criterion 3. The proposed increase in maximum fuel 
enrichment and the changes to the SFP design basis will not 
significantly reduce the margin of safety.
    The use of higher enriched fuel and the changes to the SFP 
design basis have the potential to affect only criticality events 
during fuel handling and storage. Criticality analyses demonstrate 
that the limits on Keff for the new and spent fuel 
storage areas will be satisfied. Therefore, there is adequate margin 
to ensure subcriticality during the storage and handling of fuel. 
The requirements of 10 CFR 50 Appendix A General Design Criterion 62 
are satisfied. Safety analyses demonstrated that Keff 
will remain sufficiently low to ensure subcriticality, so no new 
releases will result and there is no impact on radiological 
consequences of accidents. The safety analyses of record will remain 
applicable for the operation of fuel with a higher initial U\235\ 
enrichment and changes to the spent fuel pool. Therefore, the margin 
of safety is not affected by the proposed increase in initial fuel 
enrichment or changes to the spent fuel pool design basis.
    Based on the evaluations and analyses results presented in the 
foregoing safety significance evaluation, it has been demonstrated 
that increasing the North Anna Units 1 and 2 maximum initial fuel 
enrichment to 4.6 weight percent U\235\ and changing the design 
basis of the spent fuel pool to eliminate any credit for Boraflex 
but take credit for soluble boron in the pool will not result in a 
significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Donald P. Irwin, Esq., Hunton and 
Williams, Riverfront Plaza East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Section Chief: Richard L. Emch, Jr.

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,

[[Page 77930]]

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).

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

    Date of application for amendments: July 27, 2000.
    Brief Description of amendments: The amendments change the 
Technical Specifications to allow one of each unit's Direct Current 
power subsystems to be inoperable when in Modes 4 and 5, and during 
movement of irradiated fuel assemblies in the secondary containment.
    Date of issuance: November 29, 2000.
    Effective date: November 29, 2000.
    Amendment Nos.: 211 and 238.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: September 20, 2000 (65 
FR 56948). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 29, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 1, 2000.
    Brief description of amendments: The amendments revise TS Section 
3.7.15 and associated Bases, and Section 4.0 for the McGuire Nuclear 
Stations, Units 1 and 2, to allow the use of credit for soluble boron 
in spent fuel pool criticality analyses. The request is based on the 
NRC-approved Westinghouse Owners Group Topical Report WCAP-14416-NP-A, 
which provides generic methodology for crediting soluble boron.
    Date of issuance: November 27, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 197 and 178.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 18, 2000 (65 FR 
62385). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 27, 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: October 18, 2000.
    Brief description of amendments: The amendments revise the 
implementation date of Amendment Nos. 312, 312, and 312 from November 
30, 2000, so that implementation will be on or before implementation of 
amendments resulting from the application that must be submitted by 
April 5, 2001. This submittal will be based on an engineering study 
that is being conducted to evaluate both the appropriate Keowee Hydro 
Unit out-of-tolerance surveillance criteria and resolve overshoot 
concerns.
    Date of Issuance: November 27, 2000.
    Effective date: As of the date of issuance.
    Amendment Nos.: 317/317/317.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Implementation Date.
    Date of initial notice in Federal Register: October 25, 2000 (65 FR 
63896). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 27, 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: May 25, 2000.
    Brief description of amendment: The amendment changed the action 
statements for Technical Specification (TS) 3.8.2.2, A.C. 
Distribution--Shutdown, and TS 3.8.2.4, DC Distribution--Shutdown, by 
replacing the requirement to establish containment integrity within 
eight hours with a requirement to immediately suspend core alterations, 
the movement of irradiated fuel assemblies, and any operations 
involving positive reactivity additions. Related changes to the 
associated Bases were also made.
    Date of issuance: November 28, 2000.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 227.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 12, 2000 (65 FR 
43045). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 28, 2000.
    No significant hazards consideration comments received: No.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: May 31, 2000.
    Brief description of amendment: The Technical Specifications (TS) 
were revised by adding an additional Condition to ITS 3.3.11, Emergency 
Feedwater Initiation and Control System Instrumentation, regarding the 
required action to be taken for one or more Emergency Feedwater 
Initiation and Control System channels when up to two Reactor Coolant 
Pump status signals are inoperable.
    Date of issuance: November 21, 2000
    Effective date: November 21, 2000
    Amendment No.: 194.
    Facility Operating License No. DPR-72: Amendment revised the TS.
    Date of initial notice in Federal Register: July 12, 2000 (65 FR 
43047). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 21, 2000.
    No significant hazards consideration comments received: No.

[[Page 77931]]

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 application for amendments: June 8, 2000.
    Brief description of amendments: The amendments allow the use of 
probabilistic risk assessment (PRA) techniques in evaluating the need 
for tornado-generated missile barriers; this provides an alternative to 
installing physical missile protection for those structures, systems, 
and components that are not physically protected from tornado-generated 
missiles.
    Date of issuance: November 17, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 247 and 228.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
approved revision of the UFSAR.
    Date of initial notice in Federal Register: July 12, 2000 (65 FR 
43049). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 17, 2000.
    No significant hazards consideration comments received: No.

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 application for amendments: September 1, 2000, as 
supplemented October 27, 2000.
    Brief description of amendments: The licensee proposed the 
following three changes:
    (1) A one-time change to Unit 1 Technical Specification (TS) 
Surveillance Requirement (SR) 4.6.1.2 to add the following: ``A one-
time exception to the requirement to perform post-modification Type A 
testing is allowed for the steam generators and associated piping, as 
components of the containment barrier. For this case, American Society 
of Mechnical Engineers (ASME) Section XI leak testing will be used to 
verify leak tightness of the repaired or modified portions of the 
containment barrier. Entry into MODES 3 and 4 following the extended 
outage that commenced in 1997, may be made to perform this testing.''
    (2) A change to Unit 1 and Unit 2 TS SR 4.6.1.2 to add the phrase 
``except as modified by NRC-approved exemptions'' to the requirement to 
perform testing in accordance with 10 CFR Part 50, Appendix J, Option 
B, and the September 1995 version of Regulatory Guide 1.163.
    (3) A change to the Unit 1 and Unit 2 Bases TS SR 4.6.1.2 to add 
the phrase ``Regulatory Guide 1.163, dated September 1995, and Nuclear 
Energy Institute (NEI) document NEI 94-01, except as modified'' after 
the surveillance testing for measuring leakage rates are consistent 
with the Appendix ``J'' of 10 CFR Part 50.
    Date of issuance: November 17, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 248 and 229.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 20, 2000 (65 
FR 56953). 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 amendments is contained 
in a Safety Evaluation dated November 17, 2000.
    No significant hazards consideration comments received: No.

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 application for amendments: April 6, 2000, as supplemented 
November 13, 2000.
    Brief description of amendments: The amendments would approve 
changes involving unreviewed safety questions to the Updated Final 
Safety Analysis Report to incorporate new methodology to be used in the 
analysis of high-energy line breaks at D. C. Cook.
    Date of issuance: November 21, 2000.
    Effective date: As of the date of issuance.
    Amendment Nos.: 249 and 230.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 23, 2000 (65 FR 
51355). 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 amendments 
is contained in a Safety Evaluation dated November 21, 2000.
    No significant hazards consideration comments received: No.

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 application for amendments: October 18, 2000, as 
supplemented November 10, 2000.
    Brief description of amendments: The amendments revise Technical 
Specifications (TSs) 3/4.7.1.2, ``Auxiliary Feedwater [AFW] System,'' 
to change the description in the TSs surveillance requirement (SR) 
4.7.1.2.d of the position for each automatic valve in the AFW system 
from the ``fully open'' position to the ``correct'' position.
    Date of issuance: November 30, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 250 and 231.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 25, 2000 (65 FR 
63899) 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 amendments is contained 
in a Safety Evaluation dated November 30, 2000.
    No significant hazards consideration comments received: No.

North Atlantic Energy Service Corporation, et al., Docket No. 50-443, 
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: June 20, 2000, as supplemented on 
September 25, 2000.
    Description of amendment request: The amendment revises the 
Technical Specifications (TS) by removing the prescriptive requirement 
for determining the reactor coolant system flow rate by precision heat 
balance in Surveillance Requirement 4.2.5.3. The amendment also revises 
TS Table 2.2-1 to reflect the allowed calibration tolerance of the 
protection racks and noting that the Trip Setpoint for Functional Unit 
12, Reactor Coolant Flow-Low reactor trip is based on an indicated 
value rather than a measured value.
    Date of issuance: October 26, 2000.
    Effective date: As of its date of issuance, and shall be 
implemented at commencement of Cycle 8 operation (scheduled for 
November 2000).
    Amendment No.: 77.

[[Page 77932]]

    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 9, 2000 (65 FR 
48753) The supplemental letter provided clarifying information 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 26, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 6, 2000.
    Brief description of amendment: The amendment deletes or modifies 
license conditions and confirmatory orders to reflect the permanently 
defueled condition of the unit.
    Date of Issuance: November 15, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 108.
    Facility Operating License No. DPR-21: The amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: July 26, 2000 (65 FR 
46010). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 15, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 12, 2000.
    Brief description of amendment: The amendment revises the Technical 
Specification 4.6.E.1.d safety/relief valve bellows monitoring system 
test frequency from quarterly to once per operating cycle.
    Date of issuance: November 30, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 114.
    Facility Operating License No. DPR-22: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 28, 2000 (65 FR 
39959). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 30, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 21, 1999, as 
supplemented May 2, 2000
    Brief description of amendments: These amendments incorporate 
changes to the Technical Specifications (TSs) to more clearly define 
the requirements for service water (SW) system operability in 
accordance with the system configuration assumed in the SW system 
analysis. The application dated December 21, 1999, as supplemented May 
2, 2000, superceded an application dated July 30, 1998, in its 
entirety. The December 21, 1999, application was submitted because the 
licensee performed additional analyses of the SW system subsequent to 
the submittal of the July 30, 1998, application, which necessitated 
additional changes to the TSs.
    Date of issuance: November 17, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 199 and 204.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 23, 2000 (65 
FR 9014). The May 2, 2000, supplemental letter provided additional 
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 amendments is contained 
in a Safety Evaluation dated November 17, 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: June 14, 2000, as supplemented 
on October 12, 2000.
    Brief description of amendments: The amendments modify the Salem 
Unit Nos. 1 and 2 Technical Specifications (TS), and allow PSEG Nuclear 
to use the Best Estimate Analyzer For Core Operations--Nuclear (BEACON) 
system at Salem to fulfill certain TS surveillance requirements that 
involve core power distribution measurements. BEACON is a core power 
distribution monitoring and support system based on a three dimensional 
nodal code. The system is used to provide data reduction for incore 
neutron flux maps, core parameter analysis and follow, and core 
prediction.
    Date of issuance: November 6, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days of issuance.
    Amendment Nos.: 237 and 218.
    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 
46014). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 6, 2000.
    No significant hazards consideration comments received: No.

Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek 
Generating Station, Salem County, New Jersey

    Date of application for amendment: November 24, 1999, as 
supplemented September 14, 2000.
    Brief description of amendment: The amendment revises the Technical 
Specifications to implement Filtration, Recirculation, and Ventilation 
System and Control Room Emergency Filtration System charcoal filter 
testing requirements that are consistent with the U.S. Nuclear 
Regulatory Commission guidance delineated in Generic Letter 99-02, 
``Laboratory Testing of Nuclear-Grade Activated Charcoal.''
    Date of issuance: November 17, 2000
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 130.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73096). The September 14, 2000, supplement provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the scope of the original 
application. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 17, 2000.
    No significant hazards consideration comments received: No.

[[Page 77933]]

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: July 20, 2000 (PCN-488, 
supplement 1; supersedes application dated August 11, 1999).
    Brief description of amendments: The amendments revised Technical 
Specifications surveillance requirements (SRs) related to the 
acceptance criteria for TS 3.3.7, ``Diesel Generator (DG)--Undervoltage 
Start,'' SR 3.3.7.3, which verifies operability of the loss of voltage 
and degraded voltage actuation circuits. The amendments replaced the 
analytical limits currently specified as acceptance criteria with 
allowable values, and deleted SR 3.3.7.4 on the basis that it is 
redundant with SR 3.3.7.3.
    Date of issuance: November 29, 2000.
    Effective date: November 29, 2000, to be implemented within 30 days 
of issuance.
    Amendment Nos.: Unit 2-174; Unit 3-165.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 23, 2000 (65 FR 
51362). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 29, 2000.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama

    Date of application for amendments: August 28, 2000.
    Description of amendment request: The amendments revise the Units 
1, 2 and 3 Technical Specifications (TS) to incorporate TS Task Force 
(TSTF) Items Nos. TSTF-71, TSTF-208, TSTF-222, TSTF-284, TSTF-258 and 
TSTF-364. TSTFs are changes to the Improved Standard TS that were 
initiated by the nuclear power industry and submitted to the NRC staff. 
A description of each of the six TSTFs proposed for implementation at 
Browns Ferry follows: (1) TSTF-71, Revision 2, adds an example of the 
application of the Safety Function Determination Program to the Bases 
for Limiting Conditions for Operation (LCO) 3.0.6. (2) TSTF-208, 
Revision 0, extends the allowed time to reach MODE 2 in LCO 3.0.3 from 
7 hours to 10 hours. The change is based on plant experience regarding 
the time needed to perform a controlled shutdown in an orderly manner. 
(3) TSTF-222, Revision 1, clarifies Improved Technical Specification 
(ITS) Section 3.1.4, Control Rod Scram Times, Surveillance Requirements 
(SRs) to better delineate the requirements for testing control rods 
following refueling outages and for control rods requiring testing due 
to work activities. (4) TSTF-258, Revision 4, revises TS Section 5.0, 
Administrative Controls, to delete specific TS staffing requirement 
provisions for Reactor Operators (ROs), eliminates TS details for 
working hour limits, clarifies requirements for the Shift Technical 
Advisor position, adds regulatory definitions for Senior ROs and ROs, 
revises the Radioactive Effluent Controls Program to be consistent with 
the intent of 10 CFR Part 20, deletes periodic reporting requirements 
for mainsteam relief valve openings, and revises radiological area 
control requirements for radiation areas to be consistant with those 
specified in 10 CFR 20.1601(c). (5) TSTF-284, Revision 3, modifies 
Improved TS Section 1.4, Frequency, to clarify the usage of the terms 
``met'' and ``performed'' to facilitate the application of SR Notes. 
Two new SR Examples, 1.4-5 and 1.4-6, are added to illustrate the 
application of the terms. (6) TSTF-364, Revision 0, revises Section 
5.5.10, TS Bases Control Program, to reference 10 CFR 50.59 rather than 
``unreviewed safety question.'' Also, editorial change WOG-ED-24, which 
substitutes ``require'' for ``involve'' in 5.5.10.b is made for 
consistency in usage.
    Date of issuance: November 21, 2000.
    Effective date: November 21, 2000.
    Amendment Nos.: 239, 266, and 226.
    Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: 
Amendments revised the licenses.
    Date of initial notice in Federal Register: October 4, 2000 (65 FR 
59224)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 21, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: September 15, 2000.
    Brief description of amendments: The proposed change replaces the 
general references currently provided in Technical Specification 5.6.6 
for determining the reactor coolant system pressure and temperature 
limits with the requirement that the Pressure/Temperature Limits and 
Low Temperature Overpressure Protection System Setpoints shall not be 
revised without prior U.S. Nuclear Regulatory Commission approval.
    Date of issuance: November 27, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 81 & 81.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 1, 2000 (65 FR 
65351). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 27, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 19, 2000.
    Brief description of amendment: The amendment revises the Technical 
Specifications to establish operability requirements to ensure that 
adequate reactor coolant inventory and sufficient heat removal 
capability exist during cold shutdown and refueling conditions.
    Date of Issuance: November 17, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 195.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 18, 2000 (65 FR 
62393). The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated November 17, 2000.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of application for amendments: November 29, 1999, as 
supplemented August 31, 2000.
    Brief description of amendments: The amendments revise the testing 
requirements in Technical Specification (TS) 4.7.7.1 and TS 4.7.8.1 to 
incorporate the American Society for Testing and Materials D3803-1989 
standard and the application of a safety factor of 2.0 for the charcoal 
filter

[[Page 77934]]

efficiency assumed in Virginia Electric and Power Company's design-
basis dose analysis.
    Date of issuance: November 20, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 224 and 205.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: February 9, 2000 (65 FR 
6413). The August 31, 2000, supplement provided clarifying information 
only, and did not change the initial no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated November 20, 2000.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 6th day of December 2000.
    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-31541 Filed 12-12-00; 8:45 am]
BILLING CODE 7590-01-P