[Federal Register Volume 65, Number 212 (Wednesday, November 1, 2000)]
[Notices]
[Pages 65337-65356]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-27938]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from October 9, 2000, through October 20, 2000. 
The last biweekly notice was published on October 18, 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 
Review and Directives Branch, Division of Administration Services, 
Office of Administration, U.S. Nuclear Regulatory

[[Page 65338]]

Commission, Washington, DC 20555-0001, and should cite the publication 
date and page number of this Federal Register notice. Written comments 
may also be delivered to Room 6D22, Two White Flint North, 11545 
Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal 
workdays. Copies of written comments received may be examined at the 
NRC's Public Document Room, the Gelman Building, 2120 L Street, NW., 
Washington, DC through September 22, 2000. The NRC has relocated its 
Public Document Room to the NRC's headquarters building. Effective 
September 26, 2000, documents may be examined at the NRC's Public 
Document Room, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852. The filing of requests for a 
hearing and petitions for leave to intervene is discussed below.
    By December 1, 2000, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, located at One 
White Flint North, 11555 Rockville Pike (first floor), Rockville, 
Maryland 20852. Publicly available records will be accessible and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room). If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 
Publicly available records will be accessible and electronically from 
the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room).

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Nuclear Power Plant, Unit 1, Wake and Chatham 
Counties, North Carolina

    Date of amendment request: October 4, 2000.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to support the replacement of 
the current Westinghouse Model D4 steam generators (SGs) with 
Westinghouse Model Delta 75 replacement steam generators (RSGs), which 
is planned to occur during the fall 2001 refueling outage. The proposed 
changes to the TS

[[Page 65339]]

are required in part as a result of the physical differences between 
the currently installed Model D4 SGs and the Model Delta 75 RSGs. The 
Model Delta 75 RSGs have thermally treated alloy 690 tube material, 
which has been proven through laboratory testing and operational 
experience to provide increased corrosion resistance compared to the 
Inconel 600 tube material in the Model D4 SGs. The licensee has 
completed a comprehensive engineering review program in support of the 
SG replacement. This program evaluated the difference between the 
current Model D4 SGs and the Model Delta 75 RSGs in conjunction with 
restoration of the original nominal reactor coolant average temperature 
(Tavg) of 588.8 deg.F. The supporting analyses and evaluations were 
performed to support operations with the RSGs at the current licensed 
core power of 2775 MWt and also, where possible, at an uprated core 
power level of 2900 MWt. This license amendment application, however, 
does not include a request for a power uprate.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not affect accident initiators. The 
change in RCS [reactor coolant system] volume, an input in certain 
accident analyses, is acceptable since applicable acceptance 
criteria continue to be met. The change to the calculated residual 
heat removal (RHR) cooldown time does not adversely impact safe 
shutdown and meets Standard Review Plan (SRP) requirements. The 
analysis results for peak accident pressure (Pa remain 
below the containment design pressure limit. Revised Technical 
Specification limits, such as SG minimum water levels, have been 
analytically determined based on the new SG design features to 
preserve analysis assumptions. Since the proposed change has no 
effect on accident initiators, the probability of design basis 
events is not impacted by these analyses.
    The leak rate assumed in accident analyses is conservative and 
independent of the value of Pa. Since the peak calculated 
accident pressure remains below the design value, the containment 
function as a barrier to the release of radioactive material is not 
adversely affected. Dose consequences have been analyzed with 
respect to the proposed change and applicable acceptance criteria 
continue to be met.
    HNP [Harris Nuclear Plant] structures, systems and components, 
as well as the RSGs, are designed to operate at the original 
Tavg value of 588.8 deg.F. The comprehensive engineering 
review performed to support SG replacement includes evaluation and 
analysis results for transients and design basis accidents that meet 
applicable acceptance criteria. The proposed DNB [departure from 
nucleate boiling] parameter values ensure the core and reactor 
coolant system do not exceed their safety limits. The trip setpoints 
initiate safeguards actions that mitigate postulated accident 
consequences. Since plant systems will function as designed and 
performance requirements are verified to be acceptable with the 
proposed changes, there are no impacts to accident initiators or 
precursors. Therefore, these changes do not involve a significant 
increase in the probability of an accident previously evaluated.
    The RSGs are designed to minimize potential for the types of 
tube degradation experienced with the OSGs [original steam 
generators]. The changes to reactor core safety limits, RTS/ESFAS 
[reactor trip system/engineered safety feature actuation system] 
values and Tavg limit support transient and accident 
analysis assumptions relating to fuel clad failure. Since SG tube 
and fuel clad integrity will continue to be adequately maintained, 
these barriers to release of radioactive material will perform their 
required function. All dose consequences continue to meet acceptance 
criteria. As a result, the changes do not involve a significant 
increase in consequences of previously evaluated accidents.
    The proposed change reduces Technical Specification limits on 
reactor coolant activity to ensure that the dose consequences of the 
proposed new SGTR [steam generator tube rupture] analysis meet 
guidelines. The Technical Specification limits are not accident 
initiators and the proposed change provides additional conservatism 
in the limiting initial condition for RCS activity level. The change 
in these limits functions only to mitigate consequences of RCS 
liquid and steam release events and does not impact methods for 
plant operation. Since there are no adverse impacts to initiators of 
accidents previously evaluated, there is no significant increase in 
the probability of such accidents.
    The proposed change to the basis for thyroid dose conversion 
factors is an accepted update to that used in the analysis of 
record. Dose consequences for all analyses, including SGTR, remain 
within allowable guidelines. Although there are increases to 
previously reported dose consequences, the increase is not 
significant since the results remain under the SRP acceptance 
criteria.
    Reducing the required time at hot standby to maintain adequate 
CST [condensate storage tank] minimum volume ensures accident 
mitigation capability and is not involved with accident initiation. 
Conservative reactor trip setpoints during power operation with 
inoperable main steam line safety valves ensure accident analysis 
assumptions for maximum secondary system pressure. The proposed 
setpoint changes are also unrelated to initiation of previously 
evaluated accidents.
    The removal of F* plugging criteria is consistent with NUREG-
0452, ``Standard Technical Specifications.'' The remaining Technical 
Specification plugging criterion is based on a minimum wall 
thickness due to wastage as determined by ASME [American Society of 
Mechanical Engineers] Section XI. The proposed change is 
conservative because operation with F* degraded tubes is eliminated.
    The potential for a tube rupture is expected to be acceptably 
low based on the qualification analysis and testing for the RSGs. 
The proposed program for periodic in-service inspection monitors the 
integrity of the SG tubing to provide reasonable assurance that 
there is sufficient time to take proper and timely corrective action 
if any tube degradation is detected. The deleted OSG tube inspection 
requirements are not applicable to the RSG due to design 
differences.
    SG tube inspections are not accident initiators and the function 
of the inspection program is to identify and eliminate, when 
appropriate, accident precursor conditions. Because of this, the 
probability of postulated accidents (e.g., SG tube rupture) is not 
increased.
    Since the proposed changes are either conservative (reduced RTS 
setpoints, elimination of F* tube plugging criteria) or related to 
mitigation of postulated accidents (hot standby time prior to 
cooldown), there is no significant increase in dose consequences.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not adversely impact methods of plant 
operation and new methods of plant systems operation are not 
created. The impacts of SG replacement on NSSS [nuclear steam supply 
system] and other plant systems, including the change in RCS volume 
and other design differences, have been evaluated and design and 
performance requirements continue to be met. Accordingly, no new 
accident scenarios, failure mechanisms or limiting single failures 
are introduced.
    Changing Pa to reflect the results of current 
analyses while maintaining its value below the containment design 
pressure will not introduce significant or adverse changes to the 
plant design basis that would create a new or different kind of 
accident.
    Since HNP structures, systems and components are designed for 
operation at the proposed Tavg, returning to this value 
following SG replacement does not introduce a new method of plant 
operation. The proposed reactor core safety limits, RTS/ESFAS and 
Tavg limit changes do not introduce new failure 
mechanisms or limiting single failures since the values have been 
shown to preserve transient and accident analysis assumptions. The 
acceptable evaluation results demonstrate that the changes do not 
adversely affect or challenge the performance or integrity of safety 
related systems. Hence, no new accident scenarios are created.
    The proposed change to specific activity limits does not modify 
plant systems,

[[Page 65340]]

structures, or components or impact methods of plant operation. The 
SGTR analysis that provides the basis for the change demonstrates 
adequate margin to overfill based on revised emergency operating 
procedures [EOPs]. Operator actions and response times have been 
demonstrated to be acceptable and to satisfy analysis assumptions in 
simulator exercises. The operator actions do not represent different 
types of actions from those currently included in the EOPs, and are 
not subject to errors that would create new failure mechanisms.
    No new types of failures and no new or different accident 
initiators are created. As a result, the change will not introduce 
significant or adverse changes to the plant design basis that could 
lead to the creation of a new or different kind of accident.
    The revised thermal power restrictions imposed by [the] 
Technical Specification[s] when there are inoperable main steam line 
safety valves are conservative and consistent with the Technical 
Specification Bases methodology. The reduction in hot standby time 
to six hours continues to satisfy requirements for safe shutdown. No 
new failure mechanisms or limiting single failures are introduced 
and there are no adverse effects or challenges to the performance or 
integrity of safety related systems.
    Removing application of F* repair criteria and sleeving 
methodologies from [the] Technical Specification[s] with the RSGs 
does not introduce significant or adverse changes to the plant 
design basis that could lead to a new or different kind of accident 
being created. These requirements were needed to address design and 
material conditions of the original HNP SGs that are not applicable 
to the RSGs. This change does not alter the objective of SG 
surveillance activities--maintaining the structural integrity of 
this portion of the reactor coolant system. The surveillance 
activities are performed during outages and the proposed 
surveillance program is consistent with regulatory guidance. No new 
failures are created.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Analyses supporting the change reflect the RSG design, including 
changes to primary and secondary volume and minimum water level for 
adequate heat sink. Mass and energy and containment response 
analyses model ESF [engineered safety features] systems and safety 
analysis limits. The analysis results demonstrate that applicable 
acceptance criteria, including DNB, RCS pressure, peak cladding 
temperature, containment pressure and temperature, and dose, 
continue to be met.
    The calculated containment peak accident pressures result in a 
value of Pa that remains below the containment design 
pressure. The margin of safety during design basis events is 
preserved if Pa does not exceed the design pressure. In 
addition, performance of the containment leakage rate testing 
program ensures the validity of the assumptions used in the accident 
analyses.
    All evaluations and analyses performed to support SG replacement 
reflect the proposed values. The results demonstrate that applicable 
acceptance criteria (including DNB, RCS pressure and temperature, 
LOCA [loss-of-coolant accident] peak clad temperature, containment 
pressure and temperature, and dose) continue to be met. Calculations 
confirm that the proposed RTS/ESFAS values provide margin to the 
safety analysis limits, thereby ensuring that the margin of safety 
inherent in the safety analysis limit itself is preserved.
    The specific activity limit has no effect on safety limits or 
limiting safety system settings. The only impact of the specific 
activity limits is on dose consequences for certain design basis 
accidents such as SGTR. The proposed change is in the conservative 
direction for dose. The SGTR analysis is the most limiting of these 
accident dose evaluations. The proposed new analysis demonstrates 
that the SGTR consequences remain within allowable SRP guideline 
limits, which establish the margin of safety. Since the analysis 
results remain below the SRP guidelines, the margin of safety is not 
reduced.
    The thermal power restrictions imposed by [the] Technical 
Specification[s] when there are inoperable main steam line safety 
valves [ensure] sufficient relieving capacity during all postulated 
transients, thereby maintaining margin of safety for secondary 
system pressure.
    The proposed in-service inspection program for the RSGs monitors 
the integrity of the SG tubing and is consistent with NUREG-0452. 
Since the F* repair criteria and sleeving methods specifically 
addressed OSG degradation issues, their removal does not involve a 
reduction in a margin of safety. Analyses demonstrate that the RSG 
tubing retains adequate structural and leakage integrity using the 
proposed plugging criteria during normal, transient, and postulated 
accident conditions. The RSG tubing design and proposed plugging 
criteria are consistent with applicable ASME requirements and do not 
allow for operation with indications identified by F* criteria.
    The proposed plugging limit maintains the margin of safety by 
providing for leakage detection and shutdown in the event of an 
unexpected tube leak while minimizing the potential for excessive 
leakage or tube burst in the event of [an] MSLB [main steam line 
break] or LOCA.
    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: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Richard P. Correia.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: September 20, 2000.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 5.5.7.d to decrease the maximum 
allowed pressure drops across control room emergency filtration (CREF) 
make-up and recirculation train filters and charcoal adsorbers. 
Specifically, the pressure drops would be changed from ``6.0 inches WG 
[water gauge]'' to ``3.0 inches WG'' for the CREF makeup train, and 
from ``8.0 inches WG'' to ``4.2 inches WG'' for the CREF recirculation 
train. Additionally, the words ``(CREF only)'' would be removed to 
clarify that standby gas treatment system (SGTS) prefilter is included 
in the Ventilation Filter Testing Program (VFTP).
    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 revise the pressure drop acceptance 
criteria of TS Section 5.5.7.d from ``6.0 inches WG'' to ``3.0 
inches WG'' for the CREF makeup train, and from ``8.0 inches WG'' to 
``4.2 inches WG'' for the CREF recirculation train. The change in 
pressure drop reflects the impact of reduced fan speed on system 
characteristics. The removal of ``(CREF only)'' is editorial and 
clarifies the inclusion of the SGTS prefilters in the VFTP. These 
changes assure that the Surveillance procedure appropriately 
demonstrate[s] the ability of the CREF trains to perform its 
specified function.
    No changes in either system design or operating strategies will 
be made as a result of these changes. Thus, no opportunity exists to 
increase the probability or consequences of a previously analyzed 
accident. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of a 
previously evaluated accident.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a change to the plant 
operation. As a result, the proposed change does not affect any of 
the parameters or conditions that could contribute to the initiation 
of any accidents. This change involves the pressure drop across the 
filters and adsorbers of the CREF make-up and recirculation trains. 
The

[[Page 65341]]

surveillance requirements for performing the actual test have not 
changed. The VFTP tests are performed such that the pressure drop 
across the combined HEPA [high efficiency particulate adsorber] 
filters, the prefilters, and the charcoal adsorbers is less than the 
value specified in the acceptance criteria in accordance with 
Regulatory Guide 1.52, Revision 2, and ANSI/ASME N510-1980 at the 
system flowrate of plus or minus 10 percent. The change involves 
restrictive changes in test acceptance criteria and test methodology 
and no change in system operation.
    No new accident scenarios are created by the lower value of 
filter and adsorber pressure drop. No safety-related equipment or 
safety functions are altered as a result of this change. 
Additionally, the removal of ``(CREF only)'' is editorial and 
clarifies the inclusion of the SGTS prefilters in the VFTP. 
Therefore, the changes do not create the possibility of a new or 
different kind of accident or malfunction from those previously 
analyzed.
    3. The change does not involve a significant reduction in the 
margin of safety.
    The proposed change in combination with existing restrictions 
within the TS provides assurance that there are no credible 
mechanisms to prevent the CCHVAC [control center heating, 
ventilation, and air conditioning] system from performing its 
specified function. The maximum allowable differential pressure is 
more restrictive and corresponds to the reduced fan speed across the 
HEPA filters, prefilters, and charcoal adsorbers as a result of this 
change. There will be no changes in system operating strategies 
because of the inclusion of the make-up moisture separator/prefilter 
and test acceptance criteria for the CREF trains. Additionally, the 
removal of ``(CREF only)'' is editorial and clarifies the inclusion 
of the SGTS prefilters in the VFTP. 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: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: June 13, 2000.
    Description of amendment request: The proposed amendments would 
make administrative changes to the McGuire Nuclear Station, Units 1 and 
2, Facility Operating Licenses (FOLs). Specifically, they would: (1) 
Delete existing License Conditions which have been met by completed 
licensee's actions or are imposed by other regulatory requirements, or 
(2) make other miscellaneous changes to the McGuire Nuclear Station, 
Units 1 and 2, FOLs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will the change involve a significant increase in the 
probability or consequence of an accident previously evaluated?
    No. The proposed amendment to the FOLs are either 
administrative, eliminate duplication of other regulatory 
requirements or delete License Conditions fully met by Duke Energy 
Corporation. No actual plant equipment, operating practices, or 
accident analyses are affected by this proposed amendment. 
Therefore, the proposed amendment has no impact on the possibility 
of any type of accident: new, different, or previously evaluated.
    2. Will the change create the possibility of a new or different 
kind of accident from any previously evaluated?
    No. The proposed amendment to the FOLs [are] either 
administrative, eliminate duplication of other regulatory 
requirements or delete License Conditions fully met by Duke Energy 
Corporation. No actual plant equipment, operating practices, or 
accident analyses are affected by this proposed amendment and no 
failure modes not bounded by previously evaluated accidents are 
created. Therefore, the proposed amendment has no impact on the 
possibility of any type of accident: new, different, or previously 
evaluated.
    3. Will the change involve a significant reduction in a margin 
of safety?
    No. Margin of safety is associated with confidence in the 
ability of the fission product barriers (i.e., fuel and fuel 
cladding, reactor coolant system pressure boundary, and containment 
structure) to limit the level of radiation dose to the public. The 
proposed amendment to the FOLs are either administrative, eliminate 
duplication of other regulatory requirements or delete License 
Conditions fully met by Duke Energy Corporation. Therefore, no 
reduction in any existing margin of safety is involved.

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

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

    Date of amendment request: August 22, 2000.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications and associated Bases to limit the 
amount of time a Refueling Water Storage Tank level channel can be in a 
trip condition.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability of consequences of an accident previously evaluated?
    No. The proposed change modifies the allowed outage time that a 
channel of the Refueling Water Storage Tank (RWST) can be in the 
tripped condition from an indefinite period of time to a more 
conservative maximum of 48 hours. The Engineered Safety Features 
Actuation System (ESFAS) is an accident mitigating system, and not 
an accident initiator. Therefore, the proposed change will have no 
impact on any accident probabilities. Accident consequences will not 
be affected, as no changes are being made to the plant involving a 
reduction in reliability or effectiveness of the Emergency Core 
Cooling System (ECCS). Consequently, any previous evaluations 
associated with accidents will not be affected by this change.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed change does not modify the design or 
configuration of the plant. The proposed change provides a more 
conservative time limit for a channel to be in the tripped 
condition. No physical changes are being made to plant systems, 
structures or components nor will the proposed change reduce the 
ability of any of the safety related equipment required for accident 
mitigation. Consequently, this change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    No. Margin of safety is related to the confidence in the ability 
of the fission product barriers to perform their design functions 
during and following accident conditions. These barriers include the 
fuel cladding, the reactor coolant system, and the containment 
system. The proposed change provides a more restrictive time limit 
for a channel of the RWST to be in a tripped condition than is 
currently allowed by the ITS. The performance of the fission product 
barriers will not be degraded by the proposed changes. Consequently, 
plant safety analyses will not be affected.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three

[[Page 65342]]

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: September 1, 2000.
    Description of amendment request: The proposed amendment would make 
a change to Pilgrim Technical Specification Table 4.6-3. The proposed 
change substitutes ``21 (approx)'' under the column ``Effective Full 
Power Years (EFPY)'' for the current ``18 (approx).'' This proposed 
change would defer the second reactor vessel surveillance capsule pull 
to allow Pilgrim to pursue participation in the Boiling Water Reactor 
Integrated Surveillance Program Plan.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is 
presented below:
    The proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Pressure-temperature (P/T) limits (Pilgrim Technical Specifications 
Figures 3.6.1, 3.6.2, and 3.6.3) are imposed on the reactor coolant 
system to ensure that adequate safety margins against nonductile or 
rapidly propagating failure exist during normal operation, anticipated 
operational occurrences, and system hydrostatic tests. The P/T limits 
are related to the nil-ductility reference temperature, RTndt, as 
described in American Society of Mechanical Engineers (ASME) Section 
III, Appendix G. Changes in the fracture toughness properties of 
reactor pressure vessel (RPV) beltline materials, resulting from the 
neutron irradiation and the thermal environment, are monitored by a 
surveillance program in compliance with the requirements of 10 CFR Part 
50, Appendix H. The effect of neutron fluence on the shift in the nil-
ductility reference temperature of pressure vessel steel is predicted 
by methods given in Regulatory Guide (RG) 1.99, Revision 2.
    The proposed change is a revision of the second surveillance 
capsule withdrawal time, identified in Technical Specification Table 
4.6-3, from approximately 18 effective full power years (EFPY) to 
approximately 21 EFPY. This change will not affect P/T limits as given 
in Pilgrim Technical Specifications Figures 3.6.1, 3.6.2, and 3.6.3. 
This change is not related to any accidents previously evaluated. This 
change will not affect any plant safety limits or limiting conditions 
of operation. The proposed change will not affect reactor pressure 
vessel performance because no physical changes are involved and Pilgrim 
vessel P/T limits will remain in accordance with RG 1.99, Revision 2 
requirements. The proposed change will not cause the reactor pressure 
vessel or interfacing systems to be operated outside of their design or 
testing limits. Also, the proposed change will not alter any 
assumptions previously made in evaluating the radiological consequences 
of accidents.
    Therefore, the probability or consequences of accidents previously 
evaluated will not be increased by the proposed change.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change revises the second RPV material surveillance 
capsule withdrawal time in Pilgrim Technical Specification Table 4.6-3 
from approximately 18 EFPY to approximately 21 EFPY. This proposed 
change does not involve a modification of the design of plant 
structures, systems, or components (SSCs). The proposed change will not 
impact the manner in which the plant is operated as plant operating and 
testing procedures will not be affected by the change. The proposed 
change will not degrade the reliability of SSCs important to safety 
because equipment protection features will not be deleted or modified, 
equipment redundancy or independence will not be reduced, supporting 
system performance will not be downgraded, the frequency of operation 
of equipment important to safety will not be increased, and increased 
or more severe testing of equipment important to safety will not be 
imposed. No new accident types or failure modes will be introduced as a 
result of the proposed change. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from that 
previously evaluated.
    The proposed changes do not involve a significant reduction in a 
margin of safety.
    Operation in accordance with the existing P-T curves ensures 
reactor vessel and cooling system integrity. The capsule pull is a 
surveillance technique that provides data for modification of the 
curves. The margins will not change without new data from the capsule 
pull. The NRC-approved methods used to develop the temperatures 
associated with the current curves reside in Pilgrim's TSs and have 
been determined to be conservative. The proposed change will not affect 
any safety limits, limiting safety system settings, or limiting 
conditions of operation. The proposed change does not represent a 
change in initial conditions, or in a system response time, or in any 
other parameter affecting the course of an accident analysis supporting 
the Bases of any Technical Specification. Therefore, the proposed 
changes do not involve a significant reduction in any margins of 
safety.
    Based on the staff's analysis, 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: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts 02360-5599
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: August 7, 2000.
    Description of amendment request: The proposed changes would revise 
Technical Specification (TS) Bases 3/4.3.1 and 3/4.3.2 to clarify the 
actions that must be performed when Steam and Feedwater Rupture Control 
System (SFRCS) components and SFRCS-actuated components are inoperable. 
Specifically, the proposed changes would provide guidance on which TS 
actions are applicable for SFRCS-actuated components. Also, the 
proposed changes would introduce new TS 3/4.7.1.8 which would provide 
appropriate requirements for the Main Feedwater Control Valves and the 
Startup Feedwater Control Valves. Additionally, the proposed changes 
would introduce new TS 3/4.7.1.9 which would provide requirements for 
the Turbine Stop Valves.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 65343]]

As required by 10 CFR 50.91(a), the licensees have provided their 
analysis of the issue of no significant hazards consideration, which is 
presented below:

    The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the 
proposed changes and determined that a significant hazards 
consideration does not exist because operation of the Davis-Besse 
Nuclear Power Station, Unit No. 1, in accordance with these changes 
would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no such accidents are affected by 
the proposed changes. The amendment application proposes to revise 
Technical Specification (TS) Bases 3/4.3.1 and 3/4.3.2, Reactor 
Protection System and Safety System Instrumentation, to clarify Steam 
and Feedwater Rupture Control System (SFRCS) requirements; to add new 
TS 3/4.7.1.8, Main Feedwater Control Valves (MFCVs) and Startup 
Feedwater Control Valves (SFCVs); to add new TS 3/4.7.1.9, Turbine Stop 
Valves (TSVs); and to make associated changes to the Bases and TS 
Index.
    The proposed change would provide clear guidance for the actions 
required when SFRCS logic and SFRCS-actuated components are inoperable. 
The new TSs for MFCVs, SFCVs, and TSVs provide appropriate operational 
requirements for these SFRCS-actuated components. The proposed change 
will not change any system hardware or testing requirements. Initiating 
conditions and assumptions remain as previously analyzed for accidents 
in the DBNPS Updated Safety Analysis Report (USAR). The proposed 
changes to the TS Bases and Index are consistent with the changes 
described above.
    The proposed changes are consistent with the intent of NUREG-1430, 
``Standard Technical Specifications--Babcock and Wilcox Plants,'' 
Revision 1, April 1995. The proposed change does not involve a change 
to any plant hardware and does not affect the probability of any 
equipment malfunction or accident-initiating event.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the source term, containment 
isolation, or radiological releases are not affected by the proposed 
changes. The proposed changes ensure that excessive unavailability of 
SFRCS-actuated equipment which is used to mitigate accident 
consequences does not occur. The reliability of the plant hardware and 
the ability of the SFRCS-actuated equipment to perform its safety 
function are not affected. Existing system and component redundancy is 
not affected by the proposed changes. The existing system and component 
operation is not affected by the proposed changes, and the assumptions 
used in evaluating the radiological consequences in the DBNPS USAR are 
not invalidated. Therefore, for each postulated accident the 
consequences remain bounded by the consequences from the previously 
evaluated accidents.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because these proposed 
changes do not involve any physical changes to systems or components, 
nor do they alter the manner in which the systems or components are 
operated.
    3. Not involve a significant reduction in a margin of safety 
because, for the proposed changes, there are no new or significant 
changes to the initial conditions contributing to accident severity or 
consequences. Accordingly, there are no significant reductions in a 
margin of safety.
    On the basis of the above, the DBNPS has determined that the 
License Amendment Request does not involve a significant hazards 
consideration. As this License Amendment Request concerns a proposed 
change to the Technical Specifications that must be reviewed by the 
Nuclear Regulatory Commission, this License Amendment Request does not 
constitute an unreviewed safety question.

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

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

    Date of amendment request: September 19, 2000.
    Description of amendment request: The proposed amendment revises 
the Standby Liquid Control (SLC) boron solution requirements in TS 
Figure 3.1.7-1 to ensure a minimum boron concentration of 660 parts per 
million (ppm) in the reactor.
    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 current TS SRs ensure acceptable SLC boron solution volume 
and concentration values which produce a minimum boron concentration 
of 600 ppm in the reactor. This proposed amendment revises the boron 
solution requirements to ensure a minimum boron concentration of 660 
ppm in the reactor. A minimum boron concentration of 660 ppm in the 
reactor is sufficient to initiate and maintain reactor 
subcriticality as the nuclear system cools. The DAEC is currently 
pursuing the use of GE-14 fuel, power uprate to 1912 MWth and 
extended cycle length. Increasing the minimum boron concentration to 
660 ppm in the reactor will provide adequate shutdown margin for 
these future core designs. A General Electric analysis, using the 
approved methods described in Revision 14 of General Electric 
Standard Application for Reactor Fuel (GESTAR II), NEDE 24011-P-A, 
provides the basis for the increase in minimum boron concentration. 
This analysis assumes DAEC operation with an equilibrium core of GE-
14 fuel, operating at 1912 MWth with 24 month cycles and bounds all 
planned future core designs.
    The concentration requirement of the SLC boron solution will 
continue to comply with the requirements of 10 CFR 50.62(c)(4).
    The proposed editorial change to the title of TS Figure 3.1.7-1 
provides clarity only and does not alter performance of TS SR 
3.1.7.1 and SR 3.1.7.5.
    Therefore, this proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Ensuring a minimum boron concentration of 660 ppm in the reactor 
will result in an increase in the acceptable minimum volume of boron 
solution in the SLC tank. The tank can accept the increased volume 
of boron solution. The tank low and high level alarms will be 
altered; these alarms alert the operator to take action in the event 
boron solution volume is decreasing or increasing, respectively. 
Increasing the volume in the SLC tank and increases to the low and 
high level alarm setpoints will have no effect on operator actions 
to maintain system operability. Revising the SLC boron solution 
requirements to ensure a minimum boron concentration of 660 ppm in 
the reactor cannot create a new or different kind of accident.
    The proposed editorial change to the title of TS Figure 3.1.7-1 
provides consistent wording between the Figure, TS SR 3.1.7.1 and 
the TS BASES to avoid potential inconsistencies when determining the 
tank volume acceptability.
    Therefore this proposed amendment will not create the 
possibility of a new or different

[[Page 65344]]

kind of accident from any accident previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    DAEC TS SRs ensure SLC boron solution volume and concentration 
values which produce a minimum boron concentration of 600 ppm in the 
reactor. This proposed amendment revises SLC boron solution 
requirements to ensure a minimum boron concentration of 660 ppm in 
the reactor. A boron concentration of 660 ppm in the reactor is 
sufficient to initiate and maintain reactor subcriticality as the 
nuclear system cools. The DAEC is currently pursuing the use of GE-
14 fuel, power uprate to 1912 MWth and extended cycle length. A 
General Electric (GE) analysis, using the approved methods described 
in Revision 14 of General Electric Standard Application for Reactor 
Fuel (GESTAR II), NEDE 24011-P-A, provides the basis for the 
increase in minimum boron concentration. GE's analysis demonstrated 
that at a minimum boron concentration of 600 ppm in the reactor, the 
SLC shutdown margin was marginal with respect to the TS requirement 
for shutdown margin. At a minimum boron concentration of 660 ppm in 
the reactor, sufficient margin is maintained such that TS Section 
3.1.1 will be met under all anticipated conditions. The GE analysis 
assumes DAEC operation with an equilibrium core of GE-14 fuel, 
operating at 1912 MWth with 24 month cycles and bounds all planned 
future core designs.
    The concentration requirement of the SLC boron solution will 
continue to comply with the requirements of 10 CFR 50.62(c)(4).
    The SLC system is designed to inject a quantity of boron that 
produces a minimum boron concentration of 600 ppm in the reactor. An 
additional 25% of that quantity of boron is also injected to 
compensate for imperfect mixing, leakage and volume in other small 
piping connected to the reactor. This margin will be maintained such 
that an additional 25% of the quantity of boron required to achieve 
a minimum boron concentration of 660 ppm in the reactor will also be 
injected. Therefore, this proposed amendment will not involve a 
significant reduction in a margin of safety.
    The proposed editorial change to the title of TS Figure 3.1.7-1 
provides consistent wording between the Figure, TS SR 3.1.7.1 and 
the TS BASES to avoid potential inconsistencies when determining 
tank volume acceptability.

    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.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point 
Nuclear Station Unit No. 1, Oswego County, New York

    Date of amendment request: September 26, 2000.
    Description of amendment request: The licensee proposed to amend 
the unit's Technical Specifications (TSs), Sections 3.3.0, 3.3.2 thru 
3.3.5 and 3.3.7 to allow performance of reactor vessel hydrostatic or 
leakage tests, and scram time and excess flow check valve tests, when 
the reactor system temperature is above 215 degrees Fahrenheit and the 
reactor is not critical, without having to maintain primary containment 
integrity. These proposed changes are consistent with the guidelines of 
NUREG-1433, Revision 1, ``Standard Technical Specifications, General 
Electric Plants, BWR/4.'' Concurrently, the licensee proposed to amend 
TS Sections 3.4.0 thru 3.4.5 to require reactor building (secondary 
containment) integrity when the reactor system temperature is above 215 
degrees Fahrenheit. This will ensure that, should accidents occur 
during the above tests, the reactor building will perform its function 
to contain radiological releases. The licensee also proposed to impose 
a dose-equivalent iodine level of 1.5 microcuries per gram (as opposed 
to 9.47 microcuries per gram allowed during plant operation) during the 
above conditions to limit radiological consequences of possible 
accidents to limits bound by the previously analyzed main steam line 
break accident.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    The operation of the unit in accordance with the proposed amendment 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed TS changes will result in not requiring containment 
integrity, but requiring reactor building integrity (secondary 
containment) during performance of the specified tests. These are 
changes in configuration (no hardware design change is involved) to 
permit testing of various systems or components under different 
conditions. The changed conditions are not considered precursor of 
accidents. Accordingly, the probability of an accident previously 
evaluated is not increased.
    Accidents could occur during a test. As part of the proposed TS 
change, the licensee proposed to limit the dose-equivalent iodine level 
to ensure that the consequences of possible accidents will continue to 
be bounded by a previously analyzed accident, the main steam line break 
accident. Thus, there will not be any increase in the consequences of 
an accident previously evaluated.
    The operation of the unit in accordance with the proposed amendment 
will not create the possibility of a new or different kind of accident 
from any accident previously evaluated. No hardware design change is 
involved with the proposed TS changes. Other than the changes in the 
configuration of the containment, the reactor building, and the 
permissible dose-equivalent iodine level, there are no other procedural 
changes. Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any previously 
evaluated.
    The operation of the unit in accordance with the proposed amendment 
will not involve a significant reduction in a margin of safety.
    The proposed TS changes will not adversely affect the performance 
characteristics and intended functions of systems and components 
covered by the above-listed TS sections. The tests, not affected by the 
proposed amendment, will ensure the ability of the subject systems and 
components to perform their intended function. Therefore, the proposed 
changes do not involve a significant reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Marsha Gamberoni.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: August 25, 2000.
    Description of amendment request: Millstone Unit No. 1 (MP1) is 
being decommissioned. To support this activity, several modifications 
are

[[Page 65345]]

required to modify/eliminate MP1 systems that support the operation of 
structures, systems, and components that are shared or common to 
Millstone Unit No. 2 (MP2) and Millstone Unit No. 3 (MP3). One of the 
separation projects entails the replacement of the existing MP1 to MP2 
4160-volt cross-tie with a new MP3 to MP2 4160-volt cross-tie. 
Northeast Nuclear Energy Company (NNECO) has proposed a one time 
extension to the allowed outage time (AOT) for Action a.2 of Technical 
Specification (TS) 3.8.1.1, ``Electrical Power Systems--A. C. Sources--
Operating'' in order to complete this modification. The proposed change 
would extend the AOT from 72 hours to 14 days, provided the MP3 station 
blackout (SBO) diesel generator (DG) is available to supply MP2; 
otherwise the AOT would only be extended to 7 days.
    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 offsite circuits supply power to equipment required to 
support the safe shutdown and post-accident operations of MP2. The 
preferred off-site power supply is from the 345 kV [kilovolt] 
switchyard, through the reserve station service transformer (RSST). 
The alternate (delayed) source of offsite power is the 4.16 kV 
cross-tie from MP1 via bus 14H.
    To ensure that the probability of a complete loss of offsite 
power is not significantly increased, the MP1 4.16 kV cross-tie will 
only be removed from service when the weather conditions and 
forecast are favorable. Additionally, during the time that the 
alternate offsite source is inoperable, actions will be taken to 
protect the operable offsite circuit (i.e., no work will be 
conducted that could challenge the operability of the offsite 
circuit).
    Although the offsite circuits provide power to components that 
help mitigate the consequences of accidents previously evaluated, 
the extension in the AOT does not affect any of the assumptions used 
in the deterministic evaluations of these accidents. Thus, this 
change will not increase the consequences of any accident previously 
analyzed.
    Based on the above, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change is an extension to a TS AOT. It does not 
alter the physical design, configuration, or method of operation of 
the plant. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any 
previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    During the implementation of the modification to provide MP2 
with a 4.16 kV cross-tie with MP3, NNECO will:
    a. Appropriately consider the 7 day and 14 day weather forecasts 
prior to removing the MP2 4.16 kV cross-tie with MP1 from service to 
minimize the potential for loss of offsite power due to severe 
weather or salt spray.
    b. Protect the equipment redundant to the systems removed from 
service or whose power supply is affected by the modification. This 
includes limiting work on the 345 kV lines, the switchyard, the 
RSST, the diesel generators, the service water system, the high 
pressure safety injection system, and the reactor building closed 
cooling water system. This restriction will ensure that MP2 will 
remain capable of mitigating any potential design basis accident 
during the implementation of the modification.
    c. Within 7 days of entering Action a. of TS 3.8.1.1, establish 
the capability to supply MP2 with power from the MP3 SBO DG via 
operator actions within one hour of an event resulting in a loss of 
the remaining offsite source of power. The capability to utilize the 
MP3 SBO DG is a contingency measure (i.e., will be able to serve as 
a temporary diesel).
    Additionally, NNECO has evaluated the dominant sequences 
affecting plant risk using probabilistic safety analysis techniques. 
The analysis determined that the Delta Core Damage Probability 
associated with the extended allowed outage time was small.
    There will be no significant reduction in a margin of safety 
because the increased risk is acceptable, focus on maintaining the 
operability of the redundant equipment and systems will be 
increased, the probable weather conditions will be appropriately 
considered, and the capability to supply MP2 with power from the MP3 
SBO DG via operator action will be established within 7 days of 
entering Action a. of TS 3.8.1.1. The capability to utilize the MP3 
SBO DG is a contingency measure (i.e., will be able to serve as a 
temporary diesel).

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

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

    Date of amendment request: August 25, 2000.
    Description of amendment request: Millstone Unit No. 1 (MP1) is 
being decommissioned. To support this activity, several modifications 
are required to modify/eliminate MP1 systems that support the operation 
of structures, systems, and components that are shared or common to 
Millstone Unit No. 2 (MP2) and Millstone Unit No. 3 (MP3). One of the 
separation projects entails the replacement of the existing MP1 to MP2 
4160-volt cross-tie with a new MP3 to MP2 4160-volt cross-tie. 
Northeast Nuclear Energy Company (licensee) has evaluated this proposed 
new cross-tie utilizing the criteria of 10 CFR 50.59. The modification 
involves four unreviewed safety questions (USQs). One USQ pertains to 
MP2 and three USQs pertain to MP3. The licensee is requesting approval 
of the USQs.
    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:

Basis for No Significant Hazards Consideration--Millstone Unit No. 
2 Unreviewed Safety Question

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The test that will be performed to verify the ability of the 
cross-tie to supply approximately 3 MVA of power from MP3 to MP2 
will require parallel operation of the MP3 SBO [station blackout] DG 
[diesel generator] with the MP2 4160 Volt system. To ensure that the 
MP3 SBO DG and associated connection will be isolated immediately 
should any abnormal event or failure occur at MP2, a special Class 
1E instantaneous overcurrent relaying scheme will be used. This will 
provide additional assurance that if a fault occurs in the MP3 SBO 
DG or the associated cross-tie cabling, it will be isolated 
immediately and should not affect operability of the MP2 safety 
related 4160 Volt system. In addition, if a

[[Page 65346]]

loss of normal power occurs at MP2 during the test, the undervoltage 
condition will be detected by the credited undervoltage scheme, and 
MP2 will respond as designed to the loss of power.
    The performance of this test does not significantly alter the 
manner in which the plant is operated. There will be no adverse 
effect on plant operation or accident mitigation equipment. The 
response of the plant and the operators following a design basis 
event will not be significantly different. Therefore, the proposed 
activity does 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 installation of the Class 1E instantaneous overcurrent relay 
scheme to be used during performance of this test will provide 
additional assurance that the MP2 safety related/busses will remain 
operable during the test to parallel the MP3 SBO DG to the MP2 4160 
Volt system. The failure of the temporary overcurrent protection 
scheme would have the same effect as failure of the existing 
overcurrent protection scheme. Therefore, the proposed activity does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The installation of the Class 1E instantaneous overcurrent relay 
scheme to be used during performance of this test will provide 
additional assurance that the MP2 safety related busses will remain 
operable during the test to parallel the MP3 SBO DG to the MP2 4160 
Volt system. This will provide additional assurance that if a fault 
occurs in the MP3 SBO DG or the associated cross-tie cabling, it 
will be isolated immediately and should not affect operability of 
the MP2 safety related 4160 Volt system. In addition, if a loss of 
normal power occurs at MP2 during the test, the undervoltage 
condition will be detected by the credited undervoltage scheme, and 
MP2 will respond as designed to the loss of power.
    The proposed activity will have no adverse effect on plant 
operation or equipment important to safety. The plant response to 
the design basis accidents will not change and the accident 
mitigation equipment will continue to function as assumed in the 
design basis accident analysis. Therefore, the proposed activity 
does not involve a significant reduction in a margin of safety.

Basis for No Significant Hazards Consideration--Millstone Unit No. 
3 Unreviewed Safety Questions

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.

MP3 USQ 1

    The MP3 to MP2 4160 Volt cross-tie will impose up to an 
additional 3 MVA load on MP3. As a result, the minimum acceptable 
switchyard voltage for MP3 will increase from 334 kV to 337 kV. The 
analysis performed to support the new design shows that, with a 
minimum voltage of 337 kV in the switchyard, MP3 will remain 
connected to offsite power. The minimum voltage protection assures 
that acceptable starting and running voltages are present for the 
safety systems that may be required to operate should a loss of 
coolant accident (limiting design basis event) occur at MP3 while 
MP2 is being supplied with the safe shutdown load of 3 MVA from 
either the MP3 RSST [reserve station service transformer] or the MP3 
NSST [normal station service transformer]. Thus, the MP3 offsite 
connection will not be tripped when offsite power is available.
    The extra loading imposed by the MP2 safe shutdown loads on the 
MP3 power supplies is within the MP3 capability to simultaneously 
supply the worst case MP3 normal and accident loads. The ability of 
MP3 to mitigate design basis accidents and events is not adversely 
affected by the new design.
    This activity does not significantly alter the manner in which 
the plant is operated. There will be no adverse effect on plant 
operation or accident mitigation equipment. Also, the response of 
the plant and the operators following a design basis event will not 
be significantly different. Therefore, this activity does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.

MP3 USQ2

    The proposed design change will allow the MP3 RSST (or the MP3 
NSST, if the MP3 RSST is unavailable) to provide the alternate 
offsite source for MP2 to meet GDC [General Gesign Criterion] 17 
requirements. If the MP3 RSST is not available, MP2 will need to 
credit the MP3 NSST as the alternate offsite power supply for GDC 17 
compliance. The connection for the MP3 NSST is between breakers 13T 
and 14T. This connection point only provides 1 breaker separation 
(13T) between the MP2 RSST and the MP3 NSST. If breaker 13T is 
closed, this arrangement does not provide adequate separation 
between the two offsite sources as required by GDC 17. Opening 
breaker 13T when MP3 is shutdown will provide the required 
separation for MP2, but it will reduce the reliability of the 
offsite supply for MP3. Therefore, opening the 13T breaker to allow 
MP2 to meet GDC 17 requirements increases the probability of a loss 
of offsite power at MP3 when shutdown.
    A loss of offsite power at MP3 due to a fault in the offsite 
distribution network is a low probability event. When combined with 
the expected frequency of removing the MP3 RSST from service of once 
per MP3 refueling outage, the probability of a loss of offsite power 
at MP3 when shutdown as a result of breaker 13T being open is low. 
In addition, the MP3 shutdown risk program will evaluate the impact 
of removing the MP3 RSST from service, and plan accordingly. This 
will ensure the MP3 RSST will be removed from service when the 
shutdown risk is determined to be acceptably low. Also, at least one 
MP3 EDG will be available, as required by Technical Specifications, 
to supply power to necessary loads if offsite power is lost.
    This activity does not significantly alter the manner in which 
the plant is operated. There will be no adverse effect on accident 
mitigation equipment.
    Also, the response of the plant and the operators following a 
design basis event will not be significantly different. Therefore, 
this activity does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

MP3 USQ 3

    The connection of an additional 3 MVA load for MP2 onto the MP3 
electrical distribution system will increase the short circuit fault 
levels on the MP3 electrical distribution system. If worst case 
conditions are established, the additional contribution from the MP2 
loads would increase the fault level to above the switchgear 
ratings. The worst case short circuit conditions occur with MP3 at 
full power, the switchyard at its maximum voltage, an MP3 EDG 
paralleled to the bus for surveillance testing, and the MP3 to MP2 
cross-tie supplying 3 MVA of power to MP2.
    Although the fault levels could exceed the MP3 switchgear and 
circuit breaker ratings, it is not expected that the worst case 
conditions will be frequently established. For the majority of 
scenarios, this situation can be avoided because the new design 
allows the 3 MVA of power for MP2 to be supplied by either MP3 bus 
34A or 34B. Procedural guidance will be provided to select the MP3 
bus (34A or 34B) to supply power to MP2 that powers the opposite 
train from the train associated with the MP3 EDG to be tested. Even 
if the worst case plant configuration cannot be avoided, there is a 
low probability of event occurrence given the low probability of a 
bus fault coincident with the short time that the MP3 EDG would be 
paralleled to the system. In addition, the event would only affect 
one MP3 train of safety related equipment. The other train will be 
available to function as assumed to mitigate any accidents that may 
occur.
    The use of the MP3 to MP2 cross-tie does not significantly alter 
the manner in which the plant is operated. Considering the low 
likelihood of a fault occurring concurrently with worst case 
conditions, and that only one MP3 train of safety related equipment 
should be affected, sufficient accident mitigation equipment will be 
available to mitigate a design basis accident. In addition, the 
response of the plant and the operators following a design basis 
event will not be significantly different. Therefore, this activity 
does 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.

MP3 USQ 1

    The ability to supply 3 MVA of power from MP3 to MP2 will reduce 
the margin between the minimum switchyard voltage and the minimum 
acceptable switchyard voltage. However, the extra loading imposed by 
the MP2 safe shutdown loads on the MP3 power supplies is within the 
MP3 capability to simultaneously supply the worst case MP3 normal 
and accident loads. As a result, the response of the plant and the 
operators following an accident will not be

[[Page 65347]]

significantly different. There will be no adverse effect on accident 
mitigation equipment and no new failure modes will be introduced. 
Therefore, the activity will not create the possibility of a new or 
different kind of accident from any previously evaluated.

MP3 USQ 2

    Maintaining breaker 13T and its associated disconnect switches 
open to ensure separation and independence of the two MP2 offsite 
sources is a change to the normal configuration of MP3. The new 
configuration involves the opening of a breaker. It does not involve 
any other physical changes to the plant or the operating 
methodology. This activity does reduce the diversity of offsite 
power supplies for the MP3 NSST, which increases the potential for a 
loss of offsite power at MP3. However, the plant response to a loss 
of offsite power because breaker 13T is open will not be 
significantly different from any previously analyzed loss of offsite 
power. Therefore, opening breaker 13T and its associated disconnect 
switches will not create the possibility of a new or different kind 
of accident from any previously evaluated.

MP3 USQ 3

    The connection of an additional 3 MVA load for MP2 onto the MP3 
electrical distribution system will increase the short circuit fault 
levels on the MP3 electrical distribution system. In the worst case, 
the short circuit fault could lead to a loss of one MP3 train of 
safety related equipment. However, the safety related equipment in 
the other train would remain capable of mitigating the event. The 
loss of a train of safety-related equipment is an analyzed event. 
Therefore, this activity will not introduce the possibility of a new 
or different kind of accident than any previously evaluated.
    3. Involve a significant reduction in a margin of safety.

MP3 USQ 1

    The MP3 to MP2 4160 Volt cross-tie will impose up to an 
additional 3 MVA load on MP3. As a result, the minimum acceptable 
switchyard voltage for MP3 will increase from 334 kV to 337 kV. The 
margin between the minimum switchyard voltage (345 kV) and the 
minimum acceptable switchyard voltage will decrease from 11 kV to 8 
kV. This is an improvement for MP2. However, it is a reduction in 
the operating margin for MP3.
    The analysis performed to support the new design shows that, 
with a minimum voltage of 337 kV in the switchyard, MP3 will remain 
connected to offsite power. The minimum voltage protection assures 
that acceptable starting and running voltages are present for the 
safety systems that may be required to operate should a loss of 
coolant accident (limiting design basis event) occur at MP3 while 
MP2 is being supplied with the safe shutdown load of 3 MVA from 
either the MP3 RSST or the MP3 NSST. Thus, the MP3 offsite 
connection will not be tripped when offsite power is available.
    The extra loading imposed by the MP2 safe shutdown loads on the 
MP3 power supplies is within the MP3 capability to simultaneously 
supply the worst case MP3 normal and accident loads. As a result, 
the response of the plant and the operators following an accident 
will not be significantly different. There will be no adverse effect 
on plant operation or equipment important to safety. The plant 
response to the design basis accidents will not change and the 
accident mitigation equipment will continue to function as assumed 
in the design basis accident analysis. Therefore, this activity does 
not involve a significant reduction in a margin of safety.

MP3 USQ 2

    The proposed design change will allow the MP3 RSST (or the MP3 
NSST, if the MP3 RSST is unavailable) to provide the alternate 
offsite source for MP2 to meet GDC 17 requirements. If the MP3 RSST 
is not available, MP2 will need to credit the MP3 NSST as the 
alternate offsite power supply for GDC 17 compliance. This will 
require the 13T breaker to be open to meet GDC 17 requirements. As a 
result, the probability of a loss of offsite power at MP3 will 
increase.
    A loss of offsite power at MP3 due to a fault in the offsite 
distribution network is a low probability event. When combined with 
the expected frequency of removing the MP3 RSST from service of once 
per MP3 refueling outage, the probability of a loss of offsite power 
at MP3 when shut down as a result of breaker 13T being open is low. 
In addition, the MP3 shutdown risk program will evaluate the impact 
of removing the MP3 RSST from service, and plan accordingly. This 
will ensure the MP3 RSST will be removed from service when the 
shutdown risk is determined to be acceptably low. Also, at least one 
MP3 EDG will be available, as required by Technical Specifications, 
to supply power to necessary loads if offsite power is lost.
    The plant response to the design basis accidents will not change 
with breaker 13T open, and the accident mitigation equipment will 
continue to function as assumed in the design basis accident 
analysis. Therefore, this activity does not involve a significant 
reduction in a margin of safety.

MP3 USQ 3

    The connection of an additional 3 MVA load for MP2 onto the MP3 
electrical distribution system will increase the short circuit fault 
levels on the MP3 electrical distribution system. If worst case 
conditions are established, the additional contribution from the MP2 
loads would increase the fault level to above the switchgear 
ratings. The worst case short circuit conditions occur with MP3 at 
full power, the switchyard at its maximum voltage, an MP3 EDG 
paralleled to the bus for surveillance testing, and the MP3 to MP2 
cross-tie supplying 3 MVA of power to MP2.
    Although the fault levels could exceed the MP3 switchgear and 
circuit breaker ratings, it is not expected that the worst case 
conditions will be frequently established. For the majority of 
scenarios, this situation can be avoided because the new design 
allows the 3 MVA of power for MP2 to be supplied by either MP3 bus 
34A or 34B. Procedural guidance will be provided to select the MP3 
bus (34A or 34B) to supply power to MP2 that powers the opposite 
train from the train associated with the MP3 EDG to be tested. Even 
if the worst case plant configuration cannot be avoided, there is a 
low probability of event occurrence given the low probability of a 
bus fault coincident with the short time that the MP3 EDG would be 
paralleled to the system. In addition, the event would only affect 
one MP3 train of safety related equipment. The other train will be 
available to function as assumed to mitigate any accidents that may 
occur.
    The response of the plant and the operators following a design 
basis event will not be significantly different when using the MP3 
to MP2 4160 Volt cross-tie. The accident mitigation equipment will 
continue to function as assumed in the design basis accident 
analysis. Therefore, this activity does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: November 18, 1999, as supplemented 
August 7, 2000.
    Description of amendment request: The proposed amendment to the 
Kewaunee Nuclear Power Plant Technical Specifications requests approval 
to increase the allowable number of spent fuel assemblies stored in the 
spent fuel pools.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated. In the analysis of 
the safety issues concerning the expanded pool canal storage 
capacity, the following previously postulated accident scenarios 
have been considered:

a. A spent fuel assembly drop in the Spent Fuel Pool
b. Loss of Spent Fuel Pool cooling flow
c. A seismic event

    The probability that any of the accidents in the above list can 
occur is not significantly

[[Page 65348]]

increased by the modification. The probabilities of a seismic event 
or loss of SFP cooling flow are not influenced by the proposed 
changes. The probability of an accidental fuel assembly drop is 
primarily influenced by the methods used to lift and move the fuel. 
The method of handling fuel during the loading of the canal racks 
will be the same as current fuel handling methods, since the same 
equipment (i.e., Spent Fuel Handling Crane) and the same procedural 
guidance will be used. Since the methods used to move fuel during 
normal operations remain nearly the same as those used previously, 
there is no significant increase in the probability of an accident.
    Accordingly, the proposed modification does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The consequences of the previously postulated scenarios for an 
accidental drop of a fuel assembly in the SFP have been re-evaluated 
for the proposed change and were determined to be bounded by the 
existing analysis. The results show that the postulated accident of 
a fuel assembly striking the top of the storage racks will not 
distort the racks sufficiently to impair their functionality. The 
resulting structural damage to a falling assembly and/or a stored 
assembly has been determined to remain unchanged. The minimum 
subcriticality margin, Keff less than or equal to 0.95, 
will be maintained. The structural damage to the SFP structure, pool 
liner, and fuel assembly resulting from a fuel assembly drop 
striking the pool floor or another assembly located within the racks 
remains unchanged. The resulting structural damage to these items 
subsequent to this event is not influenced by the proposed changes.
    The consequences of a loss of SFP cooling have been evaluated 
and found to have no increase. The concern with this accident is a 
reduction of SFP water inventory from bulk pool boiling resulting in 
uncovering fuel assemblies. This situation would lead to fuel 
failure and subsequent significant increase in offsite dose. Loss of 
SFP cooling at Kewaunee is mitigated by ensuring that a sufficient 
time period exists between the loss of forced cooling and uncovering 
fuel. This period of time is compared against a reasonable period to 
re-establish cooling or supply an alternative water source (such as 
service water, makeup water, or fire protection system water). This 
evaluation included the determination of the time to boil. The time 
to boil represents the onset of loss of pool water inventory and is 
commonly used as a gage for establishing the comparison of 
consequences before and after a refueling project. The heat up rate 
in the SFP is a nearly linear function of the fuel decay heat load. 
The fuel decay heat load will increase subsequent to the proposed 
changes because of the increase in the number of stored assemblies. 
The heat up rate established for the limiting heat load conditions 
prior to the addition of canal racking was 10.8 deg.F per hour. This 
would result in the pool temperature increasing from the maximum 
design temperature of 150 deg.F to boiling in a period of 5.7 hours. 
The heat up rate established for the limiting heat load conditions 
subsequent to the proposed changes has been determined as 7.5 deg.F 
per hour. This would result in the pool temperature increasing from 
the maximum design temperature of 150 deg.F to boiling in a period 
of 8.4 hours.
    This time to boil comparison was made for limiting heat load 
conditions. However, the end of this period of time does not 
represent the onset of any significant increase in offsite doses. As 
stated above, this consequence would result after fuel is uncovered 
through unchecked boiling and the resulting water level drop of 
approximately 25 feet to the top of the fuel storage racks. This 
depth is conservative, since the top of active fuel is below this 
level. Subsequent to the proposed changes under limiting heat loads 
the maximum boil-off rate will decrease to 40.9 gpm. This will 
result in the time lapse between loss of SFP cooling and the 
uncovering of the racks to increase to 48.5 hours.
    As stated above, subsequent to racking the canal, the time to 
boil after loss of forced cooling in the most severe scenario was 
8.4 hours. The ensuing rate of evaporative loss would not result in 
the fuel being uncovered until after an additional 40.1 hours. The 
margin concerning time to boil will increase due to the proposed 
modification. This increased margin is due to the conservatism of 
the assumptions previously used to determine the heat load for this 
condition (i.e. the assumption that all fuel assemblies in the core 
were transferred to the pool instantaneously). Also, another factor 
is that the required incore hold time prior to fuel movement will 
increase from 100 hours to 148 hours. Therefore, the calculated time 
to boil in this most severe scenario will increase subsequent to the 
proposed modification. In the unlikely event that all pool cooling 
is lost, sufficient time will continue to be available subsequent to 
the proposed changes for the operators to provide alternate means of 
cooling (i.e., service water, makeup water, or fire protection 
system water) before fuel is uncovered. Therefore, the proposed 
changes represent no increase in the consequences of a loss of pool 
cooling.
    The consequences of a design basis seismic event are not 
increased. The consequences of this accident are evaluated on the 
basis of subsequent fuel damage or compromise of the fuel storage or 
building configurations leading to radiological or criticality 
concerns. The canal racks have been analyzed and found to properly 
function during seismic motion. The stored fuel has been determined 
to remain intact and the storage racks maintain the fuel and fixed 
poison configurations subsequent to a seismic event. The structural 
capability of the pool and liner will not be exceeded under the 
appropriate combinations of dead weight, thermal, and seismic loads. 
The SFP structure will remain intact during a seismic event and will 
continue to adequately support and protect the fuel racks, stored 
fuel, and pool coolant. Thus, the consequences of a seismic event 
are not increased.
    Therefore, it is concluded that the proposed changes do not 
significantly increase the probability or consequences of any 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    During and following project implementation, the spent fuel must 
be safely stored, meeting the requirements 10 CFR 20 and 10 CFR 100. 
The installation activities as well as the use of the canal racks 
must be evaluated for the possibility of creating a new or different 
kind of accident from any previously analyzed.
    This evaluation was done by reviewing the differences between 
the analyses and assumptions associated with the proposed storage 
configuration (existing racks plus canal racks) with those of the 
existing storage configuration. The accident scenarios associated 
with the existing racks were then reviewed to determine if the 
possibility of a new type of accident would be introduced by the 
implementation of the proposed modification. Also, the installation 
activities were evaluated.
    Due to the proposed changes, the following events were 
considered as the only events which might represent a new or 
different kind of accident:
    a. An accidental drop of a rack module into the pool during 
construction activity.
    b. Fuel assembly mispositioning accident in the canal.
    A construction accident resulting in a rack drop is an unlikely 
event. A rack-lifting rig will be used to lift and suspend the racks 
using the existing Fuel Handling Building crane. The crane, hoists 
and lifting rig have been or will be designed, tested and inspected 
to ensure that they will properly lift and transport the new racks. 
The postulated rack drop event is commonly referred to as a ``heavy 
load drop'' over the pools. Heavy loads will not be allowed to 
travel over any racks containing fuel assemblies. The area of 
concern represented by this event is that the pool structure will be 
compromised leading to loss of moderator/coolant. The question of a 
new or different type of event is answered by determining whether 
heavy load drops of this type have been considered previously. The 
last phase of the re-racking of the SFPs at KNP was completed in 
1987. The rack modules transported during installation of the pool 
re-racking project were significantly larger than those associated 
with the proposed canal rack project. Also, the KNP Technical 
Specifications state that ``Placement of additional fuel storage 
racks is permitted, however, these racks must not traverse directly 
above spent fuel stored in the pools''. All movements of heavy loads 
will be strictly controlled by procedure, will follow approved safe 
load paths, and will be performed by qualified personnel. Therefore, 
the rack drop does not represent a new or different kind of 
accident.
    Fuel assembly mispositioning in the canal is an unlikely event 
since assembly placement will be administratively controlled. The 
administrative controls will include positive, visual identification 
of each fuel assembly to be stored in the canal racks. The 
mispositioning event for the canal racks represents a change from 
the previously analyzed condition, since Kewaunee currently has no 
restriction on fuel storage.

[[Page 65349]]

However, the mispositioning event in the canal does not represent a 
new or different kind of accident. A fuel assembly mispositioning 
event has been previously analyzed for the existing racks and was 
found to be acceptable. The new mispositioning event for the canal 
racks was evaluated using similar techniques with similar acceptance 
criteria and was also shown to be acceptable. Therefore, since both 
events represent the mispositioning of a fuel assembly, the 
mispositioning of an assembly in the canal racks is not considered 
to represent a new or different kind of accident.
    Under the extremely unlikely conditions of a mispositioning 
event involving the placement of a fresh (un-burned) fuel assembly 
in the canal racks, the boron in the pool water will maintain the 
required subcriticality margin. Only a small fraction of the minimum 
pool boron concentration required by the KNPP Technical 
Specifications would be required for this condition. Under the 
postulated accident condition of a total loss of boron from the pool 
water, proper subcriticality margin would exist. The multiple 
accident scenario of the extremely unlikely mispositioning event 
combined with a postulated total loss of pool boron is a non-
credible condition. Therefore, although credit is taken for a small 
amount of boron in the pool water for one accident condition, a 
decrease of pool boron concentration is not a new or different kind 
of accident from any previously analyzed.
    The proposed change does not involve the modification of any 
equipment credited in the mitigation of the design basis accidents. 
The proposed change meets all applicable requirements for the safe 
storage of spent fuel at KNP. Therefore, the potential for a new or 
previously unanalyzed accident is not created by the implementation 
of this modification.
    The refueling canal does not have floor drains. Therefore, 
draining the refueling canal through a floor drain is not a 
postulated event.
    3. Involve a significant reduction in the margin of safety.
    The function of the SFP is to store the fuel assemblies in a 
subcritical and coolable configuration through all environmental and 
abnormal loadings, such as an earthquake or fuel assembly drop. The 
new rack design must meet all applicable requirements for safe 
storage and be functionally compatible with the SFP.
    WPSC has addressed the safety issues related to the expanded 
pool storage capacity in the following areas:

a. Material, mechanical and structural considerations
b. Nuclear criticality
c. Thermal-hydraulic and pool cooling

    The mechanical, material, and structural designs of the new 
racks have been reviewed in accordance with the applicable 
provisions of the NRC Guidance entitled, ``Review and Acceptance of 
Spent Fuel Storage and Handling Applications''. The rack materials 
used are compatible with the spent fuel assemblies and the SFP 
environment. The design of the new racks preserves the proper margin 
of safety during abnormal loads such as a dropped assembly and 
tensile loads from a stuck assembly. It has been shown that such 
loads will not invalidate the mechanical design and material 
selection to safely store fuel in a coolable and subcritical 
configuration.
    The methodology used in the criticality analysis of the expanded 
SFP meets the appropriate NRC guidelines and ANSI standards. The 
margin of safety for subcriticality is maintained by having the 
neutron multiplication factor equal to, or less than, 0.95 under all 
accident conditions, including uncertainties. This criterion is the 
same as that used previously to establish criticality safety 
evaluation acceptance and remains satisfied for all analyzed 
accidents. Therefore, the accepted margin of safety remains the 
same.
    The thermal-hydraulic and cooling evaluation of the pool 
demonstrated that the pool can be maintained below the specified 
thermal limits under all required design conditions. The pool 
temperature will not exceed 140 deg.F during the failure of a 
cooling pump under normal offload conditions. The pool temperature 
will remain below the maximum design temperature of 150 deg.F under 
maximum heat load conditions. The maximum local water temperature in 
the hot channel will remain below the boiling point. The fuel will 
not undergo any significant heat up after an accidental drop of a 
fuel assembly on top of the rack blocking the flow path. Following a 
loss of cooling to the pool sufficient time exists (48.5 hours for 
the limiting heat load) for the operators to intervene and line up 
alternate cooling paths and the means of inventory makeup before the 
onset of pool boiling. The thermal limit specified for the 
evaluations performed to support the proposed change is the same as 
was used in the previous evaluation. Therefore, the accepted margin 
of safety remains the same.
    Thus, it is concluded that the changes do not involve a 
significant reduction in the margin of safety.

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

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

    Date of amendment request: September 14, 2000.
    Description of amendment request: The proposed changes will delete 
Technical Specification (TS) Section 6.2.3 and revise TS Sections 6.3.1 
and 6.5.2.8 to remove the references to the Independent Safety 
Engineering Group (ISEG).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    South Carolina Electric & Gas Company (SCE&G) has evaluated the 
proposed changes to the VCSNS TS described above against the 
significant Hazards Criteria of 10 CFR 50.92 and has determined that 
the changes do not involve any significant hazard. The following is 
provided in support of this conclusion.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed amendment is a programmatic and administrative 
change to delete the ISEG. The requirement for an ISEG is located 
within the Administrative Section of TS (Section 6.0). The ISEG 
functions in an oversight role and, as such, performs no actions 
that would affect any precursors to any accident previously 
evaluated. This change does not physically alter safety-related 
systems, nor does it affect the way in which safety-related systems 
perform their functions. However, the independent oversight 
functions and qualification requirements stated in TS are being 
performed by qualified personnel in other departments of the VCSNS 
organization. This assures that the functions and qualification 
requirements delineated for the ISEG are being met. Because the 
design of the facility and system operating parameters are not being 
changed, and the oversight functions are performed by other 
departments, the proposed amendment does not involve an increase in 
the probability or consequences of any 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 proposed amendment to delete the ISEG is a programmatic and 
administrative change. There are no physical alterations to safety-
related systems and no changes to the functions of any safety-
related systems. The independent oversight functions assigned to the 
ISEG are addressed by other departments within the VCSNS 
organization. Personnel in those departments meet the qualifications 
required of the ISEG. This assures that the oversight functions and 
qualification requirements delineated for the ISEG are being met. 
Oversight functions that are performed do not in themselves lead to 
activities that are considered as accident precursors or initiators. 
Because this change does not alter the design of the facility and 
system operating parameters are not being changed, and the oversight 
functions and qualification requirements of the ISEG are being met 
by other departments, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does this change involve a significant reduction in margin of 
safety?
    The proposed amendment is a programmatic and administrative 
change to

[[Page 65350]]

delete the ISEG. This proposed change involves utilizing 
appropriately trained and qualified personnel in other departments 
to perform the oversight functions previously assigned to the ISEG. 
No physical alterations to safety-related systems and no changes to 
the functions of any safety-related systems are being made. The use 
of personnel in other departments to perform the oversight functions 
of the ISEG will provide assurance that plant operations continue to 
be conducted in a safe manner. These oversight functions will be 
performed by personnel that maintain their independence from line 
operations. Because the design of the facility and system operating 
parameters are not being changed, and the oversight functions of the 
ISEG are appropriately addressed by other departments and will 
continue to be performed, the proposed amendment 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: Thomas G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: September 6, 2000.
    Brief description of amendments: The proposed amendment would 
change Comanche Peak Steam Electric Station (CPSES), Units 1 and 2, 
Technical Specification (TS) 5.5.9, ``Steam Generator Tube Surveillance 
Program,'' to permit installation of a laser welded tube sleeve as an 
alternative to plugging defective steam generator tubes. TS 5.5.10, 
``Steam Generator Tube Inspection Report,'' would be revised to address 
reporting requirements for repaired tubes. Also, an editorial 
correction is proposed to Table 5.5-2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The tubesheet and/or tube support plate intersection laser 
welded sleeve configurations [were] designed and analyzed in 
accordance with the requirements of the ASME Code [American Society 
of Mechanical Engineers Boiler and Pressure Vessel Code]. Fatigue 
and stress analyses of the sleeved tube assemblies produced 
acceptable results. Additionally, mechanical testing for the full 
length tubesheet sleeves has shown that the structural strength of 
Alloy 690 sleeves under normal, faulted, and upset conditions is 
within acceptable limits. Leakage testing for these same \3/4\ inch 
tube sleeves has demonstrated that primary to secondary leakage is 
not expected during any plant conditions. Similar results are 
anticipated for the lower joints of elevated tubesheet sleeves. 
Confirmatory mechanical and leak testing will be conducted 
supporting the installation of elevated tubesheet sleeves at CPSES, 
Unit 1.
    The hypothetical consequences of failure of a sleeve would be 
bounded by the current steam generator tube rupture analysis 
included in the Comanche Peak Steam Electric Station (CPSES) Final 
Safety Analysis Report (FSAR). Due to the slight reduction diameter 
caused by the sleeve wall thickness, it is expected that primary 
coolant release rates would be slightly less than assumed for the 
steam generator tube rupture analysis (depending on the break 
location), and therefore, would result in lower total primary fluid 
mass release to the secondary system. Combinations of tubesheet 
sleeves and tube support plate sleeves would reduce the primary 
fluid flow through the sleeved tube assembly due to the series of 
diameter reductions the fluid would have to pass on its way to the 
break area. The overall effect would be reduced steam generator tube 
rupture release rates. The proposed Technical Specification change 
to support the installation of Alloy 690 laser welded sleeves does 
not adversely impact any other previously evaluated design basis 
accident or the results of (LOCA) [loss-of-coolant accident] and 
non-LOCA accident analyses for the current Technical Specification 
minimum RCS [reactor coolant system] flow rate.
    Conformance of the sleeve design with the applicable sections of 
the ASME Code and the successful completion of the leakage and 
mechanical tests (for the lower sleeve joint for the elevated 
tubesheet sleeves (ETS), support the conclusion that the 
installation of laser welded tube sleeves will not increase the 
probability or consequences of an accident previously evaluated. 
Depending upon the break location for a postulated steam generator 
tube rupture event, implementation of tube sleeving could act to 
reduce the radiological consequences to the public due to reduced 
flow rate through a sleeved tube compared [to] a non-sleeved tube 
based on the restriction afforded by the sleeve wall thickness.
    The editorial correction [to] Technical Specification (TS) Table 
5.5-2 is typographical in nature and does not require additional 
evaluation. Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Implementation of the laser welded sleeving (LWS) will not 
introduce significant or adverse changes to the plant design basis. 
Stress and fatigue analysis of the repair has shown the ASME Code 
minimum stress values are not exceeded. Implementation of laser 
welded sleeving restores the overall tube bundle structural and 
leakage integrity to a level consistent to that of the originally 
supplied tubing during all plant conditions. Any hypothetical 
accident as a result of potential tube or sleeve degradation in the 
repaired portion of the tube is bounded by the existing tube rupture 
accident analysis. Finally, through the results obtained from the 
extensive testing and qualification program, the possibility of a 
common-mode failure, such as multiple simultaneous steam generator 
tube failures, is not credible. Therefore, it is concluded that the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The editorial correction [to] TS Table 5.5-2 is typographical in 
nature and does not require additional evaluation. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The laser welded sleeving repair of degraded steam generated 
tubes as identified in References 1 and 2 was shown by analysis to 
restore the integrity of the tube bundle consistent with its 
original design basis condition. The safety factors used in the 
design of sleeves for the repair of degraded tubes are consistent 
with the safety factors in the ASME Boiler and Pressure Vessel Code 
used in steam generator design. The design of the full length 
tubesheet sleeve lower joints for the \3/4\ inch tube sleeves (roll-
first installation sequence) were verified by testing to preclude 
pullout and primary-to-secondary leakage during normal and 
postulated accident conditions. The qualification of the lower joint 
of the TSS [tube support sleeve], ETS and the full length tubesheet 
sleeves (FLTS) (roll-last installation sequence) will be confirmed 
at the time of the sleeving outage. Since the installed sleeve 
represents a portion of the pressure boundary, a baseline inspection 
of these areas is required prior to operation with the sleeves 
installed. The portions of the installed sleeves assembly which 
represent the reactor coolant pressure boundary can be monitored for 
the initiation and progression of sleeve/tube wall degradation, thus 
satisfying the recommendations of Regulatory Guide 1.83, Rev. 1. The 
portion of the tube bridged by the sleeve joints is effectively 
removed from the pressure boundary, and the sleeve then forms the 
new pressure boundary. The areas of the sleeved tube assembly which 
require inspection are defined in WCAP-13698, Rev. 3.
    EPRI [Electric Power Research Institute] qualified eddy current 
techniques will be used for the detection of tube degradation in \3/
4\ inch laser welded sleeved tubes. Alternate

[[Page 65351]]

inspection techniques may be used as they become available, as long 
as it can be demonstrated that the technique used provides the same 
degree or greater degree of inspection rigor.
    The effect of sleeving on the design transients and accident 
analyses were reviewed and found to remain valid up to the level of 
steam generator tube plugging consistent with the minimum reactor 
flow rate as specified in Technical Specification 3.4.1. Continued 
compliance with the RCS flow limits of Technical Specification 3.4.1 
is assured through precision flow measurements.
    Because all relevant safety analyses were reviewed and found to 
remain valid, and because the appropriate design margins are 
maintained through compliance with the relevant ASME Code 
requirements, it is concluded that the proposed change does not 
involve a significant reduction in a margin of safety.
    The editorial correction [to] TS Table 5.5-2 is typographical in 
nature and does not require additional evaluation. The confirming 
modifications to the reporting requirements of TS 5.6.10 are 
administrative only. Therefore, these 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: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

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

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

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The addition of this requirement to the Technical Specifications 
is administrative and has no impact on accident initiation or 
mitigation. Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The addition of this requirement to the Technical Specifications 
is administrative and cannot initiate an accident. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The addition of this requirement to the Technical Specifications 
is administrative and has no impact on accident initiation or 
mitigation. Therefore the proposed change does not involve a 
reduction in a margin of safety.

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

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

    Date of amendment request: September 22, 2000.
    Description of amendment request: The proposed changes would remove 
obsolete license conditions from the Operating Licenses (OLs) and 
implement associated changes to the Technical Specifications (TS) and 
Bases. The proposed changes include removal of license conditions 
associated with completed facility modifications (including the Steam 
Generator (SG) Repair Program, as well as support modifications related 
to Leak-Before-Break Technology); removal of superseded license 
conditions (addressing security); relocation of secondary water 
chemistry monitoring program requirements into the TS; removal of 
expired license conditions and TS (addressing service water piping 
restoration); and editorial 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:

    Virginia Electric and Power Company has reviewed the 
requirements of 10 CFR 50.92 as they relate to the proposed 
administrative change to the Operating Licenses, DPR-32 and DPR-37, 
and the associated Technical Specifications for Surry Units 1 and 2 
and determined that a significant hazards consideration is not 
involved. The proposed administrative change to the Surry Operating 
Licenses and associated Technical Specifications makes minor 
editorial corrections, relocates one license condition to Appendix A 
of the license, and removes outdated, superceded or otherwise non-
applicable license conditions and Technical Specifications 
requirements and provides a license document that is directly 
applicable to the current plant licensing and design bases. There is 
no safety significance associated with this proposed change since 
the change does not alter any currently applicable Operating License 
requirements. Accordingly, the current Surry licensing and design 
bases are unchanged. In support of this conclusion, the following 
evaluation is provided.
    Criterion 1--The proposed license amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change is administrative (and in part editorial) in 
nature and neither station operations nor design are affected by the 
change. The removal of license conditions and associated Technical 
Specifications regarding superceded (OL Sections 3.H and 3.L) or 
expired (OL Section 3.0, TS Table 3.7-2, and TS 3.14) requirements 
has no impact on plant operations since the requirements no longer 
have a legitimate means of being applied. The relocation within the 
Operating License of the requirement to have a secondary water 
chemistry monitoring program (OL Section 3.K) to new Section 6.4.P 
of the Technical Specifications does not alter the program or its 
implementation. The impact of the replacement SGs at Surry and the 
re-design of the SG and RCP [reactor coolant pump] supports on the 
previously evaluated accidents was performed, approved and 
documented by the issuance of the license amendments for OL Sections 
3. G and 3.M. Removal of these license conditions which refer to 
completed work and a design and licensing bases that is documented 
in the UFSAR [updated final safety analysis report] also does not 
alter station operation or the design of the affected components. 
The proposed change is within the current design and licensing bases 
of the facility. This change does not affect the initiators of 
analyzed events nor the assumed mitigation of accident or transient 
events. Analyzed events are initiated by the failure of plant 
structures, systems, or components. This change does not impact the 
condition or performance of these structures, systems or components. 
Consequences of analyzed events are the result of the plant being

[[Page 65352]]

operated within assumed parameters at the onset of any event, and 
the successful functioning of at least one train or division of the 
equipment credited with mitigating the event. These changes do not 
impact the capability of the credited equipment to perform, nor is 
there any change in the likelihood that credited equipment will fail 
to perform. As a result, the proposed change to the Surry Operating 
Licenses and Technical Specifications does not involve any increase 
in the probability or the consequences of any accident or 
malfunction of equipment important to safety previously evaluated.
    Criterion 2--The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change is administrative (and in part editorial) in 
nature. The license conditions and Technical Specifications that are 
being removed or relocated by this proposed change do not impact 
station operations or station equipment in any manner. The proposed 
change does not involve a physical alteration of the plant, nor a 
change in the methods used to respond to plant transients that has 
not been previously analyzed. No new or different equipment is being 
installed and no installed equipment is being removed or operated in 
a different manner. There is no alteration to the parameters within 
which the plant is normally operated or in the setpoints, which 
initiate protective or mitigative actions. Consequently, no new 
failure modes are introduced and the proposed change to the Surry 
Operating Licenses and Technical Specifications does not create the 
possibility of a new or different kind of accident or malfunction of 
equipment important to safety from any previously evaluated.
    Criterion 3--The proposed license amendment does not involve a 
significant reduction in a margin of safety.
    The proposed change is administrative (and in part editorial) in 
nature and neither station operations nor design are affected by the 
change. Margin of safety is established through the design of the 
plant structures, systems and components, the parameters within 
which the plant is operated, and the establishment of the setpoints 
for the actuation of equipment relied upon to respond to an event. 
Since station operations are not affected by the proposed change, 
the change does not impact the condition or performance of 
structures, systems or components relied upon for accident 
mitigation or any safety analysis assumptions. Therefore, the 
proposed change to the Surry Operating Licenses and Technical 
Specifications does not involve a reduction in any margin of safety 
described in the bases of the Technical Specifications.

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

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

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

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

    Date of amendment requests: September 22, 2000 (PCN-520).
    Brief Description of amendment requests: The proposed amendments 
would revise the San Onofre Units 2 and 3 technical specifications 
(TSs) applicable in shutdown MODES relating to positive reactivity 
additions. For a summary of specific proposed TS changes, see Tables 1 
and 2 of the licensee's application dated September 22, 2000. The 
licensee's proposal generally conforms to industry Technical 
Specification Task Force (TSTF) TSTF-286, Revision 2.
    Date of publication of individual notice in Federal Register: 
October 13, 2000 (65 FR 60984).
    Expiration date of individual notice: November 13, 2000.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendment: July 14, 2000.
    Brief description of amendment: The amendment slightly reduces the 
required minimum reactor cavity water level.
    Date of issuance: October 12, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 133.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 23, 2000 (65 FR 
51347).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 12, 2000.
    No significant hazards consideration comments received: No.

[[Page 65353]]

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

    Date of application for amendment: November 30, 1999, as 
supplemented August 11, and September 14, 2000.
    Brief description of amendment: The amendment revised the test 
standard for laboratory testing of activated charcoal to test in 
accordance with the ASTM (American Society for Testing and Materials) 
D3803-1989 standard in response to Generic Letter 99-02, ``Laboratory 
Testing of Nuclear-Grade Activated Charcoal.''
    Date of issuance: October 18, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 226.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 8, 2000 (65 FR 
12287).
    The August 11, and September 14, 2000, letters provided clarifying 
information that did not change the initial no significant hazards 
consideration determination and did not expand the scope of the 
original notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 18, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 31, 2000.
    Brief description of amendments: The amendments deleted the 
requirements related to the shorting links from Technical Specification 
(TS) Sections 3/4.3.1, ``Reactor Protection System Instrumentation;'' 
3/4.9.2, ``Refueling Operations Instrumentation;'' and 3/4.10.3, 
``Shutdown Margin Demonstrations;'' and increased the required signal-
to-noise ratio for the source range monitor in TS Sections 3/4.9.2 and 
3/4.3.7.6, ``Source Range Monitors.''
    Date of issuance: October 10, 2000.
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 142 and 128.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 9, 2000 (65 FR 
48745).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 10, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 1, 2000, as supplemented by 
letter dated August 11, 2000.
    Brief description of amendments: The amendments revise Technical 
Specification 3/4.8.1, ``A. C. Sources--Operating,'' to permit 
functional testing of the emergency diesel generators to be performed 
during power operation.
    Date of issuance: October 16, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 143 and 129.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 14, 2000 (65 FR 
37423).
    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 October 16, 2000.
    No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities 
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

    Date of application for amendments: December 30, 1999; as 
supplemented on August 3, 2000.
    Brief description of amendments: The amendments revised the 
technical specifications to: (1) Remove the Main Steamline Radiation 
Monitor (MSLRM) scram and main steamline isolation functions, and (2) 
add a new requirement for the MSLRM mechanical vacuum pump trip 
function.
    Date of issuance: October 13, 2000.
    Effective date: Immediately, to be implemented within 120 days.
    Amendment Nos.: 196 and 192.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 9, 2000 (65 FR 
48747). The August 3, 2000, supplement provided additional information 
that did not change the scope of the original proposed no significant 
hazards findings.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 13, 2000.
    No significant hazards consideration comments received: No.

Consolidated Edison Company of New York, Inc., Docket No. 50-003, 
Indian Point Nuclear Generating Station, Unit 1

    Date of amendment request: February 14, 2000.
    Brief description of amendment: The amendment change to Indian 
Point Nuclear Generating Station, Unit 1, revised Technical 
Specification Sections 2.10.2, 3.1.2, 3.2.1, 4.1.8.1, and 4.1.8.1.b. 
Specifically, Sections 3.1.2, 3.2.1, and 4.1.8.1.b, make organizational 
title changes that are administrative in nature and reflect a 
streamlining of the Consolidated Edison Company of New York, Inc.''s 
management structure, Section 4.1.8.1 is changed to reflect the 
renumbering of 10 CFR Part 20, and a footnote was moved from Section 
2.11 to Section 2.10.2 to improve the clarity of the Technical 
Specification since it pertains to text in subsection 2.10.2.4.
    Date of issuance: October 12, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No: 48.
    Facility Operating License No. DPR-5: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 5, 2000 (65 FR 
17912).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 12, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 22, 1999.
    Brief description of amendment: The amendment authorized a change 
to the Pilgrim Nuclear Power Station Updated Final Safety Analysis 
Report (UFSAR). The change involves the use of containment overpressure 
to ensure sufficient net positive suction head for the emergency core 
cooling system pumps following a loss-of-coolant accident.
    Date of issuance: October 6, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 day from the date of issuance.
    Amendment No.: 185.

[[Page 65354]]

    Facility Operating License No. DPR-35: Amendment revised the UFSAR.
    Date of initial notice in Federal Register: March 8, 2000 (65 FR 
12290).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 6, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 16, 1999, as supplemented 
on May 4 and July 10, 2000.
    Brief description of amendment: The requested changes incorporate 
Technical Specification (TS) changes to comply with the operating 
requirements derived from GE Report, NEDO-21231, ``Banked Position 
Withdrawal Sequence (BPWS)'', dated January 1977, as referenced in 
NEDE-24011-P-A. The TS changes incorporate Specifications and Actions 
based upon the plant-specific CRDA and BPWS for 20% rated thermal power 
(RTP) and 280 cal/gram peak fuel enthalpy. The TS changes also include 
changes to the control rod worth limits to resolve Licensee Event 
Report (LER) 98-006-00, dated April 30, 1998, and its supplemental LER 
98-006-01, dated August 27, 1988.
    Date of issuance: October 16, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 186.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 23, 2000 (65 FR 
51350).
    The May 4 and July 10, 2000, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the scope of the 
application as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 16, 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: August 18, 1999, as supplemented 
by letters dated June 29, July 19, and August 9, 2000.
    Brief description of amendment: The amendment revised Technical 
Specification 4.4.5, ``Steam Generators,'' to note that the 
requirements for inservice inspection do not apply during the steam 
generator replacement outage (2R14), to revise the requirement for tube 
inspection to mean an inspection from tube end (cold leg side) to tube 
end (hot leg side), to delete inspection requirements associated with 
steam generator tube sleeving and repair limits, to revise the 
preservice inspection requirements on when the hydrostatic test and the 
eddy current inspection of the tubes would be performed, and to revise 
the reporting frequency of the results of steam generator tube 
inspections to within 12 months following completion of the inservice 
inspection. Related changes to the Bases were also made.
    Date of issuance: October 4, 2000.
    Effective date: As of the date of issuance to be implemented prior 
to startup from the 2R14 refueling outage.
    Amendment No.: 223.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 23, 2000 (65 
FR 9005). The application was renoticed on August 23, 2000 (65 FR 
51353).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 4, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 1, 2000.
    Brief description of amendment: The amendment revised certain 18-
month surveillance requirements in the technical specifications by 
eliminating the condition that testing be conducted during shutdown, or 
during cold shutdown or refueling mode. The affected systems are 
emergency core cooling system, containment depressurization and cooling 
system, chemical addition system, and containment isolation valve 
system.
    Date of issuance: October 13, 2000.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment No: 118.
    Facility Operating License No. NPF-73. Amendment revised the 
technical specifications.
    Date of initial notice in Federal Register: September 12, 2000 (65 
FR 55056).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 13, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 1, 1999, and as 
supplemented by letter dated May 22, 2000.
    Brief description of amendment: Consistent with the guidance of 
Generic Letter 99-02, ``Laboratory Testing of Nuclear Grade Activated 
Charcoal,'' this amendment modifies existing Technical Specification 
5.5.7, ``Ventilation Filter Testing Program (VFTP),'' to reference ASTM 
D3803--1989, ``Standard Test Method for Nuclear-Grade Activated 
Carbon.'' The amendment also incorporates the suggested safety factor 
for charcoal filter efficiency regarding methyl iodide penetration.
    Date of issuance: October 12, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 117.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70088).
    The supplemental information contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 12, 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: May 30, 2000.
    Brief description of amendments: The amendments would make changes 
to several Technical Specifications to reflect implementation of the 
revised 10 CFR Part 20, ``Standards for Protection Against Radiation.''
    Date of issuance: October 10, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 245 and 226.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 9, 2000 (65 FR 
48752).

[[Page 65355]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 10, 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: April 28, 2000.
    Description of amendment request: The amendment implements the 
Relaxed Axial Offset Control (RAOC) operating strategy in support of 
the use of upgraded Westinghouse fuel with Intermediate Flow Mixers.
    Date of issuance: October 6, 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.: 76.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 31, 2000 (65 FR 
34747).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 6, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 19, 2000.
    Brief description of amendment: This amendment modifies Technical 
Specification (TS) 3.7.1.5, ``Plant System--Main Steam Line Isolation 
Valves.'' Specifically, the change removes the requirement to perform 
partial stroke testing of the main steam line isolation valves during 
power operation, modifies the TS wording for clarity, combines two 
surveillance requirements into one, and modifies the associated TS 
Bases for consistency.
    Date of issuance: October 19, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 185.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 23, 2000 (65 FR 
51360).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 19, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 15, 1999, as 
supplemented August 24, 2000.
    Brief description of amendments: The amendments removed TS Table 
3.6.3-1 ``Primary Containment Isolation Valves,'' and references to the 
TS Table from the TSs and relocated the information from the TS Table 
to the Technical Requirements Manual. In addition, an administrative 
change was made which deleted references to TS Tables 3.6.5.2.1-1 and 
3.6.5.2.2-1.
    Date of issuance: October 18, 2000.
    Effective date: As of date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 146 and 107.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 8, 2000 (65 FR 
12294).
    The August 24, 2000, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination or expand the scope of the original Federal 
Register Notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 18, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 3, 2000.
    Brief description of amendment: The amendment revises Technical 
Specification Table 3.2-7 by changing the reactor water level setpoint 
for the anticipated transient without scram, the recirculation pump 
trip function, and the alternate rod insertion function.
    Date of issuance: October 10, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 264.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 22, 2000 (65 FR 
15383).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 10, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 20, 1999, as 
supplemented February 4, 2000.
    Brief description of amendment: The amendment changes the Main 
Steam Isolation Valve closure scram trip level setting from 
10 percent to 15 percent valve closure.
    Date of issuance: October 10, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 265.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 9, 2000 (65 FR 
6410).
    The February 4, 2000, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination and did not expand the amendment beyond the 
scope of the original notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 10, 2000.
    No significant hazards consideration comments received: No.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: July 21, 2000
    Brief description of amendment: The amendment revises the Technical 
Specifications to remove the Control Room Emergency Air Treatment 
System Actuation Instrumentation operability during Modes 5 and 6 
except during core alterations and fuel movement based on the control 
room dose calculations.
    Date of issuance: October 10, 2000.
    Effective date: October 10, 2000.
    Amendment No.: 78.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 9, 2000 (65 FR 
48757).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 10, 2000.
    No significant hazards consideration comments received: No.

[[Page 65356]]

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: February 23, 2000, as 
supplemented July 7, 2000.
    Brief description of amendments: The amendments revise the 
licensing basis for San Onofre Nuclear Generating Station (SONGS) Units 
2 and 3 regarding the methodology for measuring the reactivity worth of 
control element assembly (CEA) groups during low power physics testing 
following a refueling. The amendments allow measuring the worth of 
approximately three-fourths of the full-length CEA groups each 
refueling cycle rather than the present methodology, which measures the 
worth of all full-length CEA groups each refueling cycle.
    Date of issuance: October 10, 2000.
    Effective date: October 10, 2000. Implementation includes 
incorporation of the changes into the Updated Final Safety Analysis 
Report (UFSAR) at the next update of the UFSAR in accordance with the 
schedule in 10 CFR 50.71(e).
    Amendment Nos.: Unit 2-173; Unit 3-164.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
authorized revision of the UFSAR Section 4.2.1.5.2, CEA Performance 
Testing.
    Date of initial notice in Federal Register: March 22, 2000 (65 FR 
15385).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 10, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 8, 2000.
    Brief description of amendment: The amendment revises the Callaway 
Technical Specifications (TS) to annotate the frequency for 
Surveillance Requirement (SR) 3.5.2.5 that verification of the 
automatic closure function of the residual heat removal pump suction 
Valve BNHV8812A shall be performed prior to startup from the first 
shutdown to Mode 5 (cold shutdown) occurring after September 8, 2000, 
but no later than June 1, 2001. The next refueling outage is scheduled 
for April 2001. This amendment defers the test of the automatic closure 
function until the next plant shut down to cold shutdown.
    Date of issuance: October 6, 2000.
    Effective date: October 6, 2000, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 140.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (65 FR 56943 dated September 20, 2000). The notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by October 20, 2000, but indicated that if the Commission makes 
a final no significant hazards consideration determination any such 
hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated 
October 6, 2000.
    Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

    Dated at Rockville, Maryland, this 25th day of October 2000.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
 Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-27938 Filed 10-31-00; 8:45 am]
BILLING CODE 7590-01-P