[Federal Register Volume 65, Number 164 (Wednesday, August 23, 2000)]
[Notices]
[Pages 51346-51369]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-21340]


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NUCLEAR REGULATORY COMMISSION


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

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 July 31, 2000, through August 11, 2000. The 
last biweekly notice was published on August 9, 2000 (65 FR 48744).

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 Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By September 22, 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

[[Page 51347]]

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, the 
Gelman Building, 2120 L Street, NW., Washington, DC, 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, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

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

    Date of amendment request: July 14, 2000
    Description of amendment request: The proposed amendment would 
slightly reduce the required minimum reactor cavity water level.
    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 any accident previously 
evaluated.
    The proposed change to the Technical Specifications involves the 
minimum or reference reactor cavity water level requirement 
(relative to the reactor pressure vessel [RPV] flange) during 
refueling operations. Reactor cavity water level can affect the 
consequences of events that may be postulated to occur during 
shutdown conditions (including fuel handling operations), namely a 
fuel handling accident, loss of normal decay heat removal 
capability, or inadvertent reactor draindown. Such events, however, 
are caused by equipment failures or human errors. The proposed 
change has no impact on such failures or errors, particularly their 
probability of occurrence. Therefore, the proposed change will not 
significantly increase the probability of a fuel handling accident, 
loss of decay heat removal, or inadvertent reactor draindown.
    With regard to impact on the consequences of postulated events/
accidents, the effect of the change on the consequences of a fuel 
handling accident is minimal. The accident producing the largest 
number of failed irradiated fuel rods is the drop of an irradiated 
fuel assembly onto the reactor core when the reactor vessel head is 
removed (Reference USAR 15.7.4.1.1). Since this event takes place 
only in the containment and the release associated with this event 
must be transferred from the containment atmosphere to the secondary 
containment, the accident which produces the most severe 
radiological release is a drop of channeled fuel onto unchanneled 
spent fuel in the fuel storage racks in the fuel building i.e. 
directly within the secondary containment. The proposed change has 
no impact on a fuel handling accident in the fuel building. A drop 
of a fuel

[[Page 51348]]

bundle on the RPV flange may involve a release of fission products 
from the dropped fuel bundle, but such a release would be less 
severe as it would involve much less fuel damage (notwithstanding 
potentially less pool depth), compared to the drop of a fuel bundle 
onto the reactor core. It has therefore been determined that 
lowering the minimum water level from 23 feet (ft) to 22 ft, 8 
inches has no significant effect on the consequences of a fuel 
handling accident.
    With respect to a loss of normal decay heat removal capability, 
or an inadvertent reactor draindown, the change reduces slightly the 
volume of water required for decay heat removal capability and 
reactor coolant inventory to mitigate a draindown event. Since the 
volume change has an insignificant effect on the reactor/pool 
volume's total available decay heat removal capability (as a backup 
in the event of a loss of normal decay heat removal capability) and 
has a negligible effect on the operator's ability to mitigate a 
draindown event, lowering the minimum specified water level from 23 
feet to 22 ft, 8 inches will not increase the consequences of such 
events.
    Based on the above, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident.
    (2) The proposed change would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change to the Technical Specifications involves a 
slight change to the minimum required/reference reactor cavity water 
level during refueling operations. No new modes of operation or the 
utilization of equipment are involved. No new accident initiators 
are introduced as a result of allowing a lower minimum/reference 
water level. Therefore, this change does not involve a new or 
different kind of accident from any accident previously evaluated.
    (3) The proposed change does not involve a significant reduction 
in the margin of safety.
    The margin of safety involved with this change involves the 
consequences that could result from the release of radioactive 
material from damaged fuel following a fuel handling accident, loss 
of decay heat removal, or inadvertent reactor draindown. The 
consequences of a dropped fuel bundle in the upper containment pool 
are insignificantly affected by allowing a slightly lower reactor 
cavity water level, as such an event would remain bounded by a 
dropped fuel bundle in the fuel building. Allowing a slightly lower 
required minimum reactor cavity water level during refueling 
operations would also have an insignificant effect on the volume of 
water available for decay heat removal capability, or to mitigate a 
draindown event. Therefore, the changes will not result in 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: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW, Washington, DC 20036-5869.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: July 27, 2000.
    Description of amendment request: The proposed amendment revises 
the Safety Limit Minimum Critical Power Ratio (SLMCPR) in the Technical 
Specifications (TSs) and makes some administrative changes associated 
with the revised SLMCPR to the TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The SLMCPR, which is determined by using NRC approved methods, 
ensures that during normal operation and/or anticipated operational 
occurrences greater than 99.9% of all fuel rods in the core avoid 
the onset of transition boiling. (The operating limit for MCPR is 
determined by adding the change in Critical Power Ratio for 
anticipated operational occurrences to the SLMCPR. For limiting 
faults such as a loss of coolant accident, SLMCPR does not apply.) 
Although the SLMCPR is established to minimize the potential for 
fuel damage in response to anticipated operational occurrences, it 
has no impact on the cause of such occurrences. That is, 
establishment of the SLMCPR has no impact on the equipment failures 
or events that can lead to such occurrences. Therefore, the proposed 
change does not involve an increase in the probability of an 
accident.
    The derivation of the cycle-specific SLMCPRs for incorporation 
into the TS has been performed using the methodology discussed in 
``General Electric Standard Application for Reactor Fuel,'' NEDE-
24011-P-A-14 (GESTAR-II), June 2000. Amendment 25, which describes 
the methodology for determining the SLMCPR, was incorporated into 
GESTAR-II in June 2000. GESTAR-II, Amendment 25 was approved by the 
NRC as of a March 11, 1999 safety evaluation report.
    The basis of the MCPR safety limit is to ensure that greater 
than 99.9% of all fuel rods in the reactor core avoid the onset of 
transition boiling if the limit is not violated. The proposed SLMCPR 
preserves the existing margin to transition boiling and fuel damage 
in the event of a postulated transient/accident. The fuel licensing 
acceptance criteria for the SLMCPR calculation apply to the next 
operating cycle at CPS (Cycle 8) in the same manner as they have 
applied previously. The new core design for two-loop and single-loop 
operation that includes GE14 fuel, is in compliance withAmendment 22 
to ``General Electric Standard Application for Reactor Fuel,'' NEDE-
24011-P-A-14 and U.S. Supplement, NEDE-24011-P-A-14-US, June 2000 
(GESTAR-II) which provides the NRC approved fuel licensing criteria. 
Since the basis of the MCPR safety limit remains unchanged, the 
probability of fuel damage and the potential consequences of 
anticipated operational occurrences is not increased. Therefore, the 
proposed TS changes do not involve an increase in the probability or 
consequences of an accident previously evaluated.
    In addition to the proposed change to the single-loop SLMCPR, 
the Note preceding TS 2.1.1.2 previously incorporated as part of 
License Amendment 113 is being proposed to be deleted. The Note 
associated with TS 2.1.1.2 was originally included to ensure that 
the SLMCPRs values were only applicable for the identified cycle 
(Cycle 7). Since that time, Amendment 25 to NEDE-24011-P-A-14 has 
been approved by the NRC, and new SLMCPRs have been calculated for 
the forthcoming fuel cycle, so this Note is no longer necessary. The 
Note was for information only and has no impact on the design or 
operation of the reactor. The proposed deletion of the Note is an 
administrative change that does not involve an increase in the 
probability or consequences of an accident previously evaluated.
    The analysis contained in TS 5.6.5, ``Core Operating Limits 
Report (COLR),'' Paragraph b., is proposed to be updated to remove 
the references to the three letters that were submitted to the NRC 
to support Cycle 7 and which are not applicable to subsequent 
operating cycles, and to retain the reference to the ongoing 
standard non-cycle specific analysis approved by the NRC (i.e., 
GESTAR). This is an administrative change to ensure that the 
references contained in the CPS TS are accurate and consistent with 
other licensing documents. Therefore, this change does not involve 
an increase in the probability or consequences of an accident 
previously evaluated.
    Based on the above, the proposed changes to the TS do not 
involve an increase in the probability or consequences of an 
accident.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The new SLMCPR limit for CPS nuclear fuel, including GE-14 fuel, 
has been determined using NRC approved methods. Use of the NRC-
approved methodology preserves the basis for the MCPR safety limit 
which ensures that during normal operation and during an anticipated 
operational occurrence greater than 99.9% of all fuel rods in the 
core avoid the onset of transition boiling. For other accidents such 
as a loss of coolant accident, the SLMCPR does not apply. The 
proposed change does not involve any new modes of operation, 
modifications to plant equipment, and any setpoint changes. As a 
result, the proposed change does not involve a new or different kind 
of accident from any accident previously evaluated.

[[Page 51349]]

    With regard to the previously described changes concerning the 
Note associated with TS 2.1.1.2 and references in TS 5.6.5, 
Paragraph b, these changes are administrative in nature. As such, 
these changes do not create the possibility of a new or different 
kind of accident from any that were previously evaluated.
    Based on the above, the proposed changes to the TS do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.

    The SLMCPRs ensure that greater than 99.9% of all fuel rods in 
the core will avoid the onset of transition boiling if the limit is 
not violated when all uncertainties are considered, thereby 
preserving the fuel cladding integrity. In addition, appropriate 
MCPR Operating Limits will continue to be enforced by procedures 
such that in the event of a transient, there will be adequate margin 
to the SLMCPR. The MCPR Operating Limits are based on the SLMCPR and 
NRC approved methods in GESTAR-II. Therefore, the proposed change to 
the single-loop SLMCPR will not involve a reduction in the margin of 
safety previously approved by the NRC.
    Additionally, the proposed changes that remove the note 
preceding TS 2.1.1.2 and the removal of outdated references in TS 
5.6.5, Paragraph b, are administrative changes that will not reduce 
the margin of safety previously approved by the NRC.
    Based on the above, the proposed changes to the TS do not 
involve any 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: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW, Washington, DC 20036-5869.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: April 1, 1999, as supplemented June 14, 
and July 27, 2000, (the April 1, 1999 application was submitted by GPU 
Nuclear, Inc., but has subsequently been adopted by AmerGen Energy 
Company, LLC). The June 14 and July 27, 2000, supplements did not 
supercede the original April 1, 1999 application in its entirety. The 
April 1, 1999, application was noticed in the Federal Register on July 
28, 1999 (64 FR 40906).
    Description of amendment request: The June 14 and July 27, 2000, 
supplements revised the original application to change the Technical 
Specification (TS) limiting conditions for operation (LCOs) and the 
surveillance requirements related to the core flood tanks to be more 
consistent with the Standard Technical Specifications for B&W [Babcock 
& Wilcox] Plants (NUREG-1430 Rev. 1) than the proposed TS changes of 
the original application. This included the addition of a new 
surveillance Table 4.1-5. The supplements also revised TS 3.3.1.3.b and 
c related to the sodium hydroxide tank limits by moving them to a new 
TS 3.3.2.1. The proposed change to TS 4.5.3.1.b.2 has been revised to 
reflect the issuance of Amendment No. 212 on June 21, 1999, which had 
previously changed that TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated. The proposed amendment makes administrative corrections, 
adds conditions to the limiting conditions for operation [LCOs], 
revises selected time clocks and surveillance requirements 
consistent with NUREG 1430, and adds a time clock to a unique LCO. 
These changes have no effect upon the plant design or operation. The 
reliability of systems and components relied upon to prevent or 
mitigate the consequences of accidents previously evaluated is not 
degraded by the proposed changes. Therefore, operation in accordance 
with the proposed amendment does not involve a significant increase 
in the probability of occurrence or consequences of an accident 
previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any previously evaluated, because no new 
accident initiators would be created.
    (3) Operation of the facility in accordance with the proposed 
amendment will not involve a significant reduction in a margin of 
safety because no changes to plant operating limits or limiting 
safety system settings are proposed.

    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.77
    Attorney for licensee: Edward J. Cullen, Jr., Esq., PECO Energy 
Company, 2301 Market Street, S23-1, Philadelphia, PA 19103.
    NRC Section Chief: Marsha Gamberoni.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina, and Docket 
Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, 
Mecklenburg County, North Carolina

    Date of amendment request: June 29, 2000, as supplemented by letter 
dated July 27, 2000.
    Description of amendment request: The amendments would revise 
McGuire Nuclear Station, Units 1 and 2, and Catawba Nuclear Station, 
Units 1 and 2 Technical Specification 5.6.5 Core Operating Limits 
Report, and the Bases of Sections 3.2.1 Heat Flux Hot Channel Factor 
FQ(X,Y,Z), 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor 
FDH(X,Y), 3.2.4 Quadrant Power Tilt Ratio, 3.5.1 
Accumulators, and 3.5.2 ECCS-Operating.
    These changes are being proposed to incorporate the Westinghouse 
Best-Estimate Large Break Loss of Coolant Analysis Methodology into the 
licensing basis for McGuire and Catawba units.
    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:

    Pursuant to 10 CFR 50.92, Duke Energy Corporation has made the 
determination that this license amendment involves no significant 
hazards considerations by applying the standards established by NRC 
regulations in 10 CFR 50.92(c). This ensures that operation of the 
facility in accordance with the proposed amendment would not:
    Involve a significant increase in the probability or consequence 
of an accident previously evaluated?
    No. The proposed changes involve use of the Best-Estimate Large 
Break Loss of Coolant Accident (LOCA) Analysis Methodology and 
implementation of associated Technical Specifications changes. The 
plant conditions assumed in the analysis are bounded by the design 
conditions for all of the equipment in the plant. Therefore, there 
will be no increase in the probability of a LOCA. Additionally, the 
consequences of a LOCA are not being increased, since it has been 
demonstrated that the Emergency Core Cooling System performance 
conforms to the criteria contained in 10 CFR 50.46(b). No other 
accidents are potentially affected by this change.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Create the possibility of a new or different kind of accident 
from any accident previously evaluated?

[[Page 51350]]

    No. The proposed changes to the Technical Specifications are to 
support implementation of Best-Estimate Large Break LOCA Analysis 
Methodology. There are no new modes of plant operation being 
introduced. The plant parameters assumed in the analysis are within 
the design limits of the existing plant equipment.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    Involve a significant reduction in a margin of safety?
    No. The analytic technique used in the analysis realistically 
describes the expected behavior of the McGuire/Catawba reactor 
system during a postulated LOCA. Uncertainties were accounted for as 
required by 10 CFR 50.46. A sufficient number of LOCA cases with 
different break sizes, different locations, and other variations in 
properties were analyzed to provide assurance that the most severe 
cases are calculated. It has been shown by the analysis that there 
is a high level of probability that all criteria contained in 10 CFR 
50.46(b) are met.
    Therefore the proposed amendment does not involve a significant 
reduction in any margin of safety.
    Duke Energy Corporation has concluded, based on the above 
discussion, that there are no significant hazards considerations 
involved in this license amendment request.

    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: Catawba Nuclear Station, Ms. Lisa F. Vaughn, 
Legal Department (PB05E), Duke Energy Corporation, 422 South Church 
Street, Charlotte, North Carolina 28201-1006. McGuire Nuclear Station, 
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: February 29, 2000, supplemented by 
letter dated July 5, 2000.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications, Table 3.3.2-1, Engineered Safety 
Feature Actuation System Instrumentation, Function 6.f, Auxiliary 
Feedwater Pump Suction Pressure-Lo.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. Only the trip setpoint and allowable value for CA pump low 
suction pressure auto-realignment to RN System are being modified in 
the Technical Specifications to accurately document the valid 
analyzed values stated in the calculations. The proposed change is 
consistent with the current licensing basis for the McGuire Nuclear 
Station, the setpoint methodologies used to develop the trip 
setpoints, the McGuire Safety Analyses, and current station 
calibration procedures and practices. 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 which will 
involve a reduction in reliability or effectiveness of the CA 
System. Consequently, any previous evaluations associated with 
accidents will not be affected by these changes.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. Only the trip setpoint and allowable value for CA pump low 
suction pressure auto-realignment to RN System are being modified in 
the Technical Specifications to accurately document the valid 
analyzed values stated in the calculations. No changes are being 
made to actual plant hardware which will result in any new failure 
modes or new accident initiation mechanisms. Also, no changes are 
being made to the way the plant is being operated. The McGuire 
Nuclear Station will continue the current practice of using the 
valid trip setpoint values documented in the instrumentation 
procedure. Consequently, no new plant accidents will be created by 
these changes.
    3. Does this change involve a significant reduction in a margin 
of safety?
    No. Only the trip setpoint and allowable value for CA pump low 
suction pressure auto-realignment to RN System are being modified in 
the Technical Specifications to accurately document the valid 
analyzed values stated in the calculations. The methods used for 
analyzing the allowable value are endorsed by Duke Power's EDM 102, 
``Instrument Uncertainty Calculations''. 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. The impact of the proposed change will not challenge or 
exceed any safety limits or design limits during a design basis 
accident. Consequently, the integrity of the fission product 
barriers will still be maintained.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina
    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: June 16, 1999, as supplemented on May 4 
and July 10, 2000.
    Description of amendment request: NEDE-24011-P-A ``General Electric 
Standard Application for Reactor Fuel'' (GESTAR-II) is one of the 
approved analytical methods for performing the reload analysis as 
specified in Technical Specification (TS) 5.6.5.b.1. The proposed 
amendment incorporates 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. NEDO-21231 forms the current basis for the Pilgrim 
reactor core design process. The Nuclear Regulatory Commission (NRC) 
staff approved NEDO-21231 by a letter to General Electric dated January 
25, 1985. NEDO-21231 describes a revised method for developing control 
rod withdrawal sequences to mitigate the consequences of the control 
rod drop accident (CRDA) in the startup and low power operating ranges 
of 20% RTP and 280 cal/gram peak fuel enthalpy. The proposed 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 proposed TS changes also include changes to the control rod 
worth limits to resolve License Event Report (LER) 98-006-00, dated 
April 30, 1998, and its supplement LER 98-006-01, dated August 27, 
1988.
    The proposed changes are modeled after NUREG-1433, Rev. 1, BWR/4 
Standard Technical Specifications (STS) for incorporating the Pilgrim 
cycle-specific data for CRDA and BPWS for 20% RTP. The STS format is 
adopted based upon GESTAR II to reflect the Specifications, Actions, 
and BASES derived from NEDO-21231. The proposed TS changes consist of 
(i) administrative changes, (ii) more restrictive changes, and (iii) 
less restrictive changes to comply with TS 5.6.5.b.1 incorporating the 
current Pilgrim core design based upon the NRC

[[Page 51351]]

approved NEDO-21231 and NEDE-24011-P-A.
    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:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes do not adversely affect accident initiators or 
precursors nor alter the design, conditions, and configuration of the 
facility or the manner in which the plant is operated. The proposed 
changes do not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the acceptance limits 
assumed in the Updated Final Safety Analysis Report (UFSAR).
    This proposed change relocating the details of the methods for 
timing control rod drives from the Specifications to the BASES involves 
no technical changes to the Specifications. The requirement to verify 
scram times is incorporated into proposed SR 3.3.B.1.4; therefore, it 
does not eliminate any requirements, or impose a new or different 
treatment of the requirements. The BASES are subject to the Technical 
Specifications Bases Control Program contained in the Administrative 
Controls Section of the Technical Specifications. Since any changes to 
the BASES will be in accordance with these requirements, no increase 
(significant or insignificant) in the probability or consequences of an 
accident previously evaluated will be allowed without prior staff 
approval.
    The proposed changes provide more stringent requirements than those 
currently in the Technical Specifications. The more restrictive 
requirements will not alter the operation of process variables or SSCs 
as described in the safety analyses; therefore, they will not involve a 
significant increase in the probability of an accident occurring.
    The proposed changes will ensure compliance with ``NEDO-21231, 
``Banked Position Withdrawal Sequence (BPWS)''. The NEDO-21231 limits 
the maximum rod worth such that fuel enthalpy addition due to a control 
rod drop accident (CRDA) will not exceed 280 cal/gm, or require the 
plant to be placed in a condition where the LCOs do not apply sooner. 
In addition, changes are proposed to require entering a MODE in which 
the LCOs do not apply sooner than currently required. Therefore, the 
new requirements would decrease the consequences of an analyzed event.
    Elimination of the requirement to shut down if one rod is stuck due 
to potential collet finger failure is being made concurrently with 
another change that will require a reactor shutdown if more than one 
rod is stuck for any reason. This additional restriction ensures that 
the reactor will be shut down as soon as it is determined that more 
than one rod may fail to scram. This differs from the existing 
requirement that allows operation with multiple stuck rods that are not 
fully inserted, provided reactivity margin is met. The consequences of 
an accident previously evaluated are not increased because the failure 
of a single control rod to insert will not prevent the reactor from 
reaching a subcritical condition as long as shutdown margin 
requirements are met.
    The proposed SRs 4.3.B.1.1 and 4.3.B.1.2 only increase the interval 
between performance of a surveillance for about 10% to 20% of the 
control rods (those that are partially withdrawn). The purpose of the 
surveillance is to verify that rods can be inserted, thus verifying 
that rods are not stuck and scram capability is maintained. The 80% to 
90% of the control rods that are fully withdrawn will continue to be 
tested at the 7-day frequency and should a stuck control rod be found, 
all withdrawn control rods will have to be tested within 24 hours. This 
change does not affect any initiating events for accidents previously 
evaluated. In addition, this change is being implemented concurrently 
with more restrictive requirements governing continued operation with 
stuck and inoperable control rods, which ensure the mitigative features 
of the control rods are maintained. Therefore, the proposed changes do 
not involve a significant increase in the probability or consequences 
of an accident previously evaluated.
    The proposed change eliminates the requirement to verify 
discernible neutron instrument response to control rod motion, the 
first time a rod is withdrawn after refueling or maintenance. The 
probability of an accident is not increased because the proposed change 
will not involve any physical changes to plant SSCs, or the manner in 
which these SSCs are operated, maintained, modified, tested, or 
inspected. The consequences of an accident are not increased because 
the CRDA analysis assumes a single failure of the control rod drive 
system when a single control rod drops out of the core from the fully 
inserted position after being disconnected from its drive and after the 
drive has been retracted to the fully withdrawn position while reactor 
power is less than 20%. During startup and before exceeding 20% reactor 
power, a large percentage of the rods are fully withdrawn in the normal 
course of a startup. All fully withdrawn rods are subjected to 
verification of coupling by the overtravel test, which verifies that 
the accident mitigation feature of the control rods is maintained.
    The proposed change will allow either a second licensed operator or 
other qualified members of the technical staff to verify movement of 
control rods when the rod worth minimizer (RWM) is inoperable. The 
function of the RWM is to control adherence to the control rod 
withdrawal and insertion sequence. The use of a second licensed 
operator or other qualified members of the technical staff to perform 
these control rod movement verifications provides alternate means to 
accomplish the same function, thus, there is no change in the 
probability or consequences of an accident previously evaluated. Also, 
the proposed change will only require that the RWM sequence be verified 
when it is changed. The RWM does not monitor core thermal conditions, 
but simply enforces preprogrammed rod patterns as a backup intended to 
prevent reactor operator error in selecting or positioning control 
rods. Therefore, these changes will not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The more restrictive and new requirements will not alter the plant 
configuration (no new or different type of equipment will be installed 
nor is any equipment being removed) or change methods governing normal 
plant operation. The changes do impose different requirements; however, 
they are consistent with assumptions made in the safety analyses, 
therefore, the proposed changes will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed less restrictive change that increases the interval 
between performance of surveillance designed to verify that rods can be 
inserted for only 10% to 20% of the control rods (those that are 
partially withdrawn) not the manner in which the surveillance is 
performed does not impact reactivity

[[Page 51352]]

controls. The changes in reactivity are not SSCs; therefore, the 
proposed changes will not involve any physical changes to the plant or 
the manner in which the plant is operated; and, therefore, the proposed 
changes will 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 a margin of 
safety?
    The proposed change relocating the details of the methods for 
timing control rod drives to the Bases does not impact the safety 
margin. The requirement to verify scram times is incorporated into 
proposed SR 3.3.B.1.4; thereby, preserving the analytic assumptions for 
the accident analyses, which also preserves the current margin of 
safety. The requirements to be transposed from the Technical 
Specifications, and are not being modified by the proposed change. 
Thus, there will be no significant reduction in the margin of safety.
    Adding new requirements and making existing ones more restrictive 
does not involve a physical alteration of the plant (no new or 
different type of equipment will be installed nor is any equipment 
being removed), introduce any new tests, or change methods governing 
normal plant operation. The BPWS limits the maximum rod worth such that 
fuel enthalpy addition due to a CRDA will not exceed 280 cal/gm, the 
current bases for the TS limit. Therefore, the proposed change does not 
involve a reduction in the margin of safety.
    Elimination of the requirement to shut down if one rod is stuck due 
to potential collet finger failure will not decrease a margin of safety 
because this change is being made concurrently with another change that 
will require a reactor shutdown if more than one rod is stuck for any 
reason. This additional restriction ensures that the reactor will be 
shut down as soon as it is determined that more than one rod may fail 
to scram, which ensures that the reactor is shut down when assumptions 
used in the analysis of those accidents and transients that depend on a 
scram may no longer be met. The failure of a single control rod to 
insert will not prevent the reactor from reaching a subcritical 
condition as long as shutdown margin requirements are met. Therefore, 
the proposed change does not involve a significant reduction in margin 
of safety.
    The proposed increase in the interval from weekly to monthly for 
partially withdrawn control rods for performances of a surveillance may 
increase the time before a partially withdrawn control rod is 
discovered to be stuck. Changing the interval between surveillances 
does not affect the surveillance acceptance criteria, thus, the 
proposed change does not affect the analysis assumptions concerning the 
number of control rods that insert following a scram.
    The proposed change will allow control rod movement verification, 
by licensed operators or other qualified members of the technical staff 
(i.e., personnel trained in accordance with an approved training 
program) when the RWM is inoperable, and limit the use of this 
alternate method to once per 12 months. This change does not impact the 
margin of safety because the verification of rod sequence and thus the 
assumed reactivity insertion rates following a reactor trip are 
maintained.
    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: W. S. Stowe, Esquire, Entergy Nuclear 
Generation Company, 800 Boylston Street, 36th Floor, Boston, 
Massachusetts 02199
    NRC Section Chief: James W. Clifford

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

    Date of amendment request: September 17, 1999, as supplemented by 
letters dated June 29 and August 3, 2000. The September 17, 1999, 
application was originally noticed in the Federal Register on February 
23, 2000 (65 FR 9004).
    Description of amendment request: The proposed amendment would 
change the Arkansas Nuclear One, Unit 2 (ANO-2) heavy load handling 
requirements and transportation provisions to permit the movement of 
the original and replacement steam generators through the ANO-2 
containment construction opening during the steam generator replacement 
outage.
    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:

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

    During the 2R14 refueling outage/steam generator replacement 
outage, the OSGs [original steam generators] and the RSGs 
[replacement steam generators] will be moved between the new steam 
generator storage area / original steam generator storage facility 
and the runway beam support system (RBSS) / outside lift system 
(OLS). The RBSS/OLS is the structure used to rig the SGs [steam 
generators] in and out of the reactor containment building. In 
consideration of the magnitude of the loads being handled, the RBSS, 
OLS and transporters are of a robust, rugged design, proven by many 
prior steam generator replacements and other heavy load handling 
operations. However, due to the location of safety related 
underground structures, systems, and components (SSCs) in the 
vicinity of the RBSS/OLS and along the steam generator (SG) haul 
route, potential load handling accidents along the load paths must 
be considered for their effects on the SSCs. At ANO-2, the ground 
cover over several buried SSCs is not sufficient to be able to rule 
out the potential for a load drop to damage or cause failure of 
these SSCs. The functions of the SSCs in question are as support 
systems to the ANO-1 [Arkansas Nuclear One, Unit 1] and ANO-2 
emergency diesel generators and the ANO-1 service water system. The 
fire protection system, a non-safety related system, was also 
considered. Existing plant procedures adequately address the 
scenario in question for the fire protection system.
    The cause of a SG drop is assumed to be a non-mechanistic 
failure of the RBSS/OLS (or associated rigging), a failure of the SG 
transporter leveling hydraulics, or a seismically-induced failure of 
the loaded RBSS/OLS or SG transporter. The possibility of drops 
associated with other external events, such as tornadoes, high 
winds, and tornado missiles will be substantially minimized by 
procedures that prevent load handling under these weather 
conditions.
    With ANO-2 defueled, the impact on ANO-2 due to loss of the 
emergency diesel generators fuel oil transfer system will be 
minimal. Long term actions to provide makeup water to the spent fuel 
pool may be necessary, but no immediate actions are required.
    For ANO-1, a steam generator drop could render both diesel 
generators inoperable due to the loss of the fuel oil transfer 
system, and the emergency cooling pond inoperable due to the loss of 
the service water return line to the pond. Since ANO-1 is expected 
to be at full power operation, these conditions would require prompt 
action in accordance with technical specifications. Immediately 
following a drop from the OLS or from the transporter in the 
vicinity of the OLS, where damage to these systems is possible, ANO-
1 will begin a shutdown and cooldown to cold shutdown conditions. In 
conjunction with the unit shutdown, contingency measures will be 
taken to compensate for the loss of the normal fuel oil supply to 
the emergency diesel generators.
    The ability of ANO-1 to safely respond to analyzed events would 
be undiminished with the possible exception of the functions 
affected by the damaged equipment. With the compensatory measures to 
be established prior to the steam generator handling operations, and 
with the planned responses to a steam generator drop, the support 
system functions of the diesel generators and the service water 
system can be assumed to be

[[Page 51353]]

maintained following the drop. Therefore, the drop will not affect 
the consequences of any analyzed event.
    While the drop of a steam generator could cause damage to some 
safety related plant equipment, the failures of these components are 
not precursors to any analyzed accident. The drop of a steam 
generator will not have any other impact on plant equipment, and 
thus will not induce any analyzed plant transient. It will, however, 
result in a malfunction of equipment important to safety of a 
different type than any previously evaluated. Based on the 
compensatory measures and the low likelihood of the event during SG 
movement, this temporary condition is considered to be acceptable. 
On these bases, it is concluded that the proposed load handling 
operations will not significantly increase the probability or the 
consequences of accidents previously analyzed.

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

    As noted in the response to the first question above, the only 
potential for a new or different kind of accident associated with 
this change request arises from a drop of a steam generator which is 
assumed to cause the loss of emergency power support systems for 
ANO-1. The cause of a SG drop is assumed to be a non-mechanistic 
failure of the RBSS/OLS (or associated rigging), a failure of the SG 
transporter leveling hydraulics, or a seismically-induced failure of 
the loaded RBSS/OLS or SG transporter. In the absence of a seismic 
event, there is no initiator for any consequential events (e.g., 
loss of offsite power) other than those directly caused by impact of 
the SG. Given this scenario, the plant response to a SG drop event 
would be governed by the technical specifications and existing plant 
procedures.
    If a SG drop is seismically-induced, the simultaneous loss of 
normal offsite power sources is also assumed in this case since 
these sources are not seismically qualified. While this event is 
very unlikely due to the low frequency of earthquakes and the small 
amount of time that a steam generator will be in a position to cause 
damage, Entergy [Operations, Inc.] will provide contingency plans 
and compensatory measures to compensate for the loss of the normal 
fuel oil supply to the emergency diesel generators. Long term 
actions to provide makeup water to the spent fuel pool may be 
necessary, but no immediate actions are required.
    Availability of the redundant ANO-1 service water heat sink, the 
Dardanelle Reservoir, during a seismic event assures that an 
uninterrupted source of service water will be available to support 
shutdown cooling of ANO-1.
    The proposed load handling plans will not create the possibility 
of a new or different kind of accident from any previously 
evaluated.

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

    ANO-1 Technical Specification 3.7.1.C requires both EDGs 
[emergency diesel generators] to be operable when the reactor 
temperature is  200  deg.F. If this condition is not met, 
Limiting Condition for Operation 3.0.3 applies. It requires that 
within one hour, action shall be initiated to place the unit in an 
operating condition in which the specification does not apply by 
placing it, as applicable, in at least hot standby within the next 6 
hours, at least hot shutdown within the following 6 hours, and at 
least cold shutdown within the subsequent 24 hours. The bases for 
technical specification 3.7.1.C indicate that these operability 
requirements ensure that an adequate, reliable power source is 
available for all electrical equipment during startup, normal 
operation, safe shutdown, and handling of all emergency situations. 
The bases for EDG operation also require at least a seven day total 
diesel oil inventory during complete loss of electrical power 
conditions.
    The postulated loss of both trains of the ANO-1 EDG fuel oil 
transfer system due to a SG drop would require that ANO-1 be shut 
down. This situation could be considered to involve a reduction in 
the margin of safety, because a new common cause failure mechanism 
is being introduced by the movement of the SGs over the EDG fuel oil 
lines and transfer pump power cables. To restore the margin of 
safety and return the EDGs to functionality, temporary compensatory 
measures are being proposed.
    Based on the above discussions, with the implementation of the 
proposed compensatory measures and the low likelihood of such an 
event, the failures caused by a SG drop event will not involve a 
significant reduction in the margin of safety.

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

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

    Date of amendment request: August 18, 1999, as supplemented by 
letters dated June 29, July 19, and August 9, 2000. The August 18, 
1999, application was originally noticed in the Federal Register on 
February 23, 2000 (65 FR 9005).
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 4.4.5, ``Steam Generators,'' to 
note that the requirements for inservice inspection do not apply during 
the steam generator replacement outage (2R14), to delete inspection 
requirements associated with steam generator tube sleeving and repair 
limits, to revise the requirement for tube inspection to mean an 
inspection from tube end (cold leg side) to tube end (hot leg side), 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 would also be made.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

    The accidents of interest are a tube rupture, loss of coolant 
accident (LOCA) in combination with a safe shutdown earthquake and a 
steam line break in combination with a safe shutdown earthquake. A 
reduction in tube integrity could increase the possibility of a tube 
rupture accident and increase the consequences of a steam line break 
or LOCA. The tubing in the replacement steam generators is designed 
and evaluated consistent with the margins of safety specified in the 
ASME [American Society of Mechanical Engineers] Code [Boiler and 
Pressure Vessel Code], Section III. The program for periodic 
inservice inspection provides sufficient time to take proper and 
timely corrective action if tube degradation is present. The ASME 
[Code], Section XI basis for the 40% through wall plugging limit is 
applicable to the replacement steam generators just as it was to the 
original steam generators. As a result there is no reduction in tube 
integrity for the replacement steam generators.
    Addition of a ``Note'' to clarify that inservice inspection is 
not required during the steam generator replacement outage is an 
administrative change that provides clarification regarding 
inservice inspection requirements. The change in reporting 
requirements is also an administrative change. The requirements for 
inservice inspection or the plugging limit for the tubes are not 
altered by these administrative changes. Additionally, changes were 
made to the bases to remove potentially misleading information. 
Bases changes are considered to be administrative in nature.
    Elimination of the repair option and the associated references 
to repair of the original steam generator tubes is an administrative 
adjustment since the sleeve design is not applicable to the 
replacement steam generators. The elimination of the repair option 
does not alter the requirements for inservice inspection or reduce 
the plugging limit for the tubes.
    A preservice eddy current inspection will be performed onsite 
prior to installation of the replacement steam generators. The

[[Page 51354]]

orientation of the replacement steam generators during the eddy 
current exam will not impact the results. The hydrostatic test 
required by the ASME Code, Section III for the replacement steam 
generators is to be performed in the manufacturing facility and not 
as part of a reactor coolant system hydrostatic test.
    The post-repair leakage test required by the ASME Code, Section 
XI for an operating plant is performed at a much lower pressure. No 
evolutions subsequent to the replacement steam generator hydrostatic 
test are expected to occur that will change the condition of the 
tubes prior to operation. This change does not alter the requirement 
to perform a preservice inspection. As a result, an inservice 
inspection is not required during the steam generator replacement 
outage.
    The requested ANO-2 [Arkansas Nuclear One, Unit 2] Technical 
Specification changes do not alter the requirements for tube 
integrity or tube plugging limits. The change to the definition of 
tube inspection is a conservative change; therefore, this change 
does not involve a significant increase in the probability or 
consequences of any accident previously evaluated.

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

    The proposed changes do not affect the design or function of any 
other safety-related component. There is no mechanism to create a 
new or different kind of accident for the replacement steam 
generators by eliminating repair criteria or by clarifying the 
applicable preservice and inservice inspection requirements because 
a baseline of tube conditions is established and plugging limits are 
maintained to ensure that defective tubes are removed from service.
    The requested ANO-2 Technical Specification changes do not alter 
the requirements for tube integrity or tube plugging limits. The 
change to the definition of tube inspection is a conservative 
change; therefore, this change does not [create the possibility of a 
new or different kind of accident from any previously evaluated].

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

    The tubing in the replacement steam generators is designed and 
evaluated consistent with the margins of safety specified in the 
ASME Code, Section III. The program for periodic inservice 
inspection provides sufficient time to take proper and timely 
corrective action to preserve the design margin if tube degradation 
is present.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, Entergy Operations 
[, Inc.] has determined that the requested change does not involve a 
significant hazards consideration.

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

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

    Date of amendment request: July 19, 2000.
    Description of amendment request: The amendment will extend the 
applicability of the current reactor coolant system (RCS) pressure/
temperature limits and maximum allowed RCS heatup and cooldown rates to 
21.7 effective full power years (EFPY) of operation. The associated low 
temperature overpressure protection (LTOP) temperature limits, which 
are based on the pressure/temperature limits, will also be extended to 
21.7 EFPY of operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The pressure-temperature (P/T) limit curves in the Technical 
Specifications are conservatively generated in accordance with the 
fracture toughness requirements of 10 CFR 50 Appendix G as 
supplemented by the ASME Code Section XI, Appendix G 
recommendations. The adjusted reference temperature (ART) values are 
based on the Regulatory Guide 1.99, Revision 2 shift prediction and 
attenuation formula and have been validated by a credible reactor 
vessel surveillance program. There are no changes to the limit 
curve, only a change in the period of applicability based on more 
recent fluence predictions. Based on the current fluence 
projections, analysis has demonstrated that the current P/T limit 
curves will remain conservative for up to 21.7 EFPY.
    In conjunction with extending the effectiveness of the existing 
P/T limit curves, the low temperature overpressure protection (LTOP) 
analysis for 15 EFPY is also extended. The LTOP analysis confirms 
that the current setpoints for the power-operated relief valves 
(PORV) will provide the appropriate overpressure protection at low 
RCS temperatures. Because the P/T limit curves have not changed, the 
existing LTOP values have not changed, this includes the PORV 
setpoints.
    The P/T limit curves and LTOP analysis have not changed; 
therefore, the proposed amendment does not represent a change in the 
configuration or operation of the plant. The results of the existing 
LTOP analysis have not changed, and the limiting pressures for given 
temperatures will not be exceeded for the postulated transients. 
Therefore, assurance is provided that reactor vessel integrity will 
be maintained. Thus, the proposed amendment does not involve an 
increase in the probability or consequences of accidents previously 
evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The requirements for P/T limit curves and LTOP have been in 
place since the beginning of plant operation. The only changes in 
these curves are the extension of the period of applicability 
(EFPY), which is based on new fluence data and the operating time 
(EFPY) required to reach the same limiting fluence used for the 
current 15 EFPY P/T curves. Since there is no change in the 
configuration or operation of the facility as a result of the 
proposed amendment, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    Analysis has demonstrated that the fracture toughness 
requirements of 10 CFR 50 Appendix G are satisfied and that 
conservative operating restrictions are maintained for the purpose 
of low temperature overpressure protection. The P/T limit curves 
will provide assurance that the RCS pressure boundary will behave in 
ductile manner and that the probability of a rapidly propagating 
fracture is minimized. Therefore, operation in accordance with the 
proposed amendment would not involve a significant reduction in a 
margin of safety.

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

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

    Date of amendment request: March 7, 2000, as supplemented on April 
21, 2000 and June 14, 2000.
    Description of amendment request: The proposed amendment would 
revise the surveillance requirements from once per refueling interval 
for each excess flow check valve (EFCV) to testing a representative 
sample of EFCVs once per 24 months.

[[Page 51355]]

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

    1. Will operation of the facility in accordance with the 
proposed amendment involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    This change will not alter the physical design of the plant. The 
proposed Amendment would modify the testing of excess flow check 
valves (EFCV) from each valve being tested once per refueling 
interval to testing a representative sample of EFCVs once per 24 
months (the length of a refueling interval). The EFCVs installed at 
Oyster Creek are extremely reliable. Oyster Creek records 
demonstrate that there has never been a failure of an EFCV to 
isolate in the thirty-year history of Oyster Creek.
    A GE [General Electric] Topical Report evaluated the reliability 
of EFCVs installed at Oyster Creek and other plants. Oyster Creek 
and three other facilities have installed Chemquip excess flow check 
valves. Chemquip EFCVs were shown in the Topical Report to have a 
failure rate of 1.78E-7, which was the lowest of the valve 
manufacturers included in the evaluation. The current Oyster Creek 
accident analysis does not take credit for any flow restriction 
provided by EFCVs although the valve design does restrict flow. 
Therefore, changing the surveillance requirements for the EFCVs does 
not involve a significant increase in the probability of an 
accident.
    EFCVs limit the reactor coolant release following the failure of 
an instrument line, valve or component on an instrument line. The 
valves isolate at a given flow and are periodically functionally 
tested to ensure proper isolation with resulting minimal flow. The 
radiological consequences of an instrument line break have been 
evaluated at Oyster Creek. That evaluation does not take credit for 
the excess flow check valve when assessing the radiological 
consequences of the accident. The analysis was submitted to the NRC 
and was approved in NUREG 1382 ``Safety Evaluation Report related to 
the full term operating license for Oyster Creek Nuclear Generating 
Station.''
    This change will not increase the consequences of an instrument 
line break or any postulated accident.
    2. Will operation of the facility in accordance with the 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    Operation of the facility in accordance with the proposed 
amendment would modify the testing frequency of EFCVs. This change 
does not add components or make any other physical change to the 
plant. The valves will be tested in the same manner as they are now 
although less frequently. EFCVs are located exclusively in 
instrument lines and the failure of an instrument line is currently 
analyzed in the FSAR [final safety analysis report]. The plant is 
not being physically changed, and the consequences of a valve 
failing to isolate are within the FSAR analyzed event. Therefore, 
this change does not create the possibility of a new or different 
accident not previously analyzed.
    3. Will operation of the facility in accordance with the 
proposed amendment involve a significant reduction in a margin of 
safety?
    The proposed Amendment would modify the testing frequency of 
excess flow check valves (EFCVs) which are located in instrument 
lines. The only function of EFCVs is to limit the reactor coolant 
release following the failure of an instrument line, valve or 
component on an instrument line. The current Oyster Creek accident 
analysis does not take credit for any flow restriction provided by 
EFCVs, although the valve design does restrict flow. The proposed 
change does not alter the plant design in any manner. Furthermore, 
the instrument line break analysis assumptions also remain 
unchanged. Therefore, there is no impact on the current procedures 
or accident analysis. As a result, operating the plant in accordance 
with the proposed Amendment does not involve a significant reduction 
in a margin of safety.

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

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

    Date of amendment requests: April 6, 2000
    Description of amendment requests: The proposed amendments would 
approve an unreviewed safety question allowing a change to the Updated 
Final Safety Analysis to allow a change to the analysis methodology 
used in the High Energy Line Break (HELB) program to incorporate the 
recommendations of NUREG/CR-2913.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    The proposed changes do not impact the design of these high-
energy lines such that previously analyzed [structures, systems and 
components] SSCs would now be more likely to fail. The changes will 
not modify high-energy lines to reduce their design capability of 
maintaining pressure boundary integrity during normal operating and 
accident conditions. The use of the NUREG/CR-2913 methodology to 
more accurately define the dynamic effects from high-energy line 
breaks and cracks does not affect the probability of any analyzed 
piping break or critical crack events. The use of the NUREG/CR-2913 
methodology does not affect high-energy line break or crack 
initiators or precursors. The [steam generator blowdown] SGBD and 
[chemical, volume and control system] CVCS letdown piping will be 
modified and analyzed, as required, to ensure that the piping 
stresses remain below the threshold for postulation of a critical 
crack or break. Also, the effects of breaks or critical cracks 
outside of the break exclusion zones have been reviewed and 
determined to not have an adverse impact on the piping within the 
exclusion zone. The modified SGBD piping in the normal flash tank 
room will be analyzed to ensure that the application of the 
[Standard Review Plan] SRP for postulating cracks based on piping 
stresses is acceptable. Therefore, incorporating these new 
methodologies does not affect equipment malfunction probability, nor 
does it affect or create new accident initiators or precursors. 
Additionally, the NRC expected the results of revisions to SRP 
Section 3.6.2 requirements to yield more efficient regulatory 
practices, improve plant piping systems design, increase plant 
reliability, and decrease occupational radiation exposure associated 
with inspections and repairs.
    The proposed changes permit relaxation of protective 
requirements that may represent a potential increase in the 
consequences of an accident. However, the proposed changes are 
consistent with the current regulatory guidelines for HELB 
evaluations and continue to ensure that protection of SSCs required 
for accident mitigation is maintained. The NUREG/CR-2913 methodology 
for determining the effects of jet flow from HELB events shows that 
SSCs outside the distance of ten piping diameters

[[Page 51356]]

from the break or critical crack are undamaged. The SRP allowances 
for break and crack exclusions embody the understanding that the 
probability of breaks or critical cracks in piping systems that 
satisfy the stress criteria is extremely low. For those areas 
addressed by the methodology changes, protection is not required 
while still providing reasonable assurance that there is no undue 
risk to the health and safety of the public. Therefore, protection 
of SSCs required for accident mitigation is assured by use of these 
well-defined design methodologies. Thus, there will be no reduction 
in the capability of those SSCs in limiting the consequences of 
previously evaluated accidents. Malfunctions caused by HELBs and 
critical cracks have been previously analyzed in the Updated Final 
Safety Analysis Report (UFSAR). Thus, no additional radiological 
source terms are generated, and the consequences of an accident 
previously evaluated in the UFSAR will not be increased.
    Therefore, the probability of occurrence or the consequences of 
accidents previously evaluated are not significantly increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not impact the design of these high-
energy lines such that previously unanalyzed breaks would now occur. 
The change to incorporate the NUREG/CR-2913 methodology does not 
introduce any new malfunctions; it more accurately defines the 
effects from the high-energy line breaks and cracks for use in the 
HELB program.
    Regarding the incorporation of the SRP break exclusion zones, 
the break exclusion stress thresholds provide assurance that the 
piping is capable of withstanding the design loadings without the 
possibility of developing a through wall crack or break. The piping 
will be modified and completely analyzed to ensure that the piping 
stresses are below the threshold for break exclusion. The effects of 
breaks outside of the break exclusion zones have been reviewed and 
determined to not have an adverse impact on the piping within the 
exclusion zone. The modified SGBD piping in the normal flash tank 
room will be analyzed to ensure that the application of the SRP for 
postulating cracks based on piping stresses is acceptable. The 
proposed changes do not result in modification to high-energy lines 
that would reduce their design capabilities to maintain pressure 
boundary integrity during normal operating and accident conditions. 
Therefore, use of the new design methodologies does not affect or 
create new accident initiators or precursors or create the 
possibility of a new or different kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The approval of the license amendment will not result in any 
modifications to high-energy lines that would reduce their design 
capabilities to maintain pressure boundary integrity during normal 
operating and accident conditions. By using these new design 
methodologies, protection of SSCs required for accident mitigation 
is assured.
    The NUREG/CR-2913 methodology better defines the extent of 
impingement loads from the postulated high-energy line breaks and 
cracks. Use of the NUREG/CR-2913 methodology establishes that 
unprotected components located more than ten diameters from a pipe 
break or crack in piping containing fluids within the assumptions of 
NUREG/CR-2913 are without further analysis assumed undamaged by a 
jet. This conclusion has been reviewed and accepted by the NRC as 
providing adequate safety margin for high-energy piping. Protection 
of SSCs required for accident mitigation will continue to be assured 
by use of the NUREG/CR-2913 methodology if modifications to those 
SSCs are implemented in the future.
    The use of the SRP break exclusion zones incorporates industry 
lessons learned and ensures that an adequate safety margin is 
maintained. The SGBD and CVCS letdown piping will be analyzed after 
modifications are performed in accordance with the original piping 
design code to ensure that the piping stresses are below the SRP 
threshold for break exclusion. Also, the effects of breaks outside 
of the break exclusion zones have been reviewed and determined to 
not have an adverse impact on the piping within the exclusion zone. 
The modified SGBD piping in the normal flash tank room will be 
analyzed to ensure that the application of the SRP for postulating 
cracks based on piping stresses is acceptable. Therefore, the 
capability of those SSCs to limit the offsite dose consequences of 
previously evaluated accidents to levels below the approved 
acceptance limits will continue to be assured.
    The SRP presents the most definitive basis available for 
specifying the NRC's design criteria and design guidelines for an 
acceptable level of safety. The SRP guidelines resulted from many 
years of experience gained by the NRC in establishing and using 
regulatory requirements in the safety evaluation of nuclear 
facilities. The implementation of the design guidelines contained in 
MEB 3-1 assures that adequate protection is provided and a 
consistent level of safety is maintained. In addition, some 
regulatory requirements developed over the years as part of the 
licensing process have resulted in additional safety margins that 
overlap the safety margins provided by the criteria of MEB 3-1. 
Consequently, use of these new design methodologies instead of the 
previous licensing basis requirements cannot significantly reduce 
the existing margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    In summary, based upon the above evaluation, I&M has concluded 
that the proposed changes involve no significant hazards 
consideration.

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

    Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
    Date of amendment requests: June 12, 2000
    Description of amendment requests: The proposed amendments would 
allow the licensee to use the methodology and the alternative source 
term (AST) contained in 10 CFR 50.67 as described in NUREG-1465, 
``Accident Source Terms for Light-water Nuclear Power Plants'' to show 
compliance with 10 CFR Part 50, Appendix A, Criteria 19.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    The proposed change to implement the AST involves changes to the 
methodologies and acceptance criterion associated with the control 
room dose analysis. The actual sequence and progression of accidents 
are not changed. However, the regulatory assumptions regarding the 
analytical treatment of the accidents are affected by the change. 
The use of an AST alone cannot increase the probability of an 
accident or the core damage frequency. The proposed change to use 
the AST does not make any changes to equipment, procedures, or 
processes that increase the likelihood of an accident. It does not 
affect any accident initiators or precursors. The methodology is 
used to determine consequences of an accident and has no impact on 
their likelihood of occurrence. Therefore, this proposed change does 
not involve a significant increase in the probability of an accident 
previously evaluated.
    The current acceptance criterion specify the dose to personnel 
in terms of ``rem whole body'' or equivalent for the duration of the 
accident, where the dose derived using the AST is given in rem 
[total effective dose equivalent] TEDE, as described in 10 CFR 
50.67. TEDE includes internal and external exposure; whole body 
includes external exposure only. The current acceptance criterion 
focuses on doses to the thyroid and the whole body. It is based on 
the assumption that the major contributor to dose will be 
radioiodine. Although this may be appropriate with the Technical 
Information Document (TID)-14844, ``Calculation of Distance Factors 
for Power and Test Reactor Sites'', source term implemented by RGs 
1.4, it may not be true for a source term based on

[[Page 51357]]

a more complete understanding of accident sequences and 
phenomenology. The AST includes a larger number of radionuclides 
than did the TID-14844 source term as implemented in regulatory 
guidance. The whole body and thyroid dose criteria considered the 
noble gases and iodine contributors as the limiting factors. The 
acceptance criteria of 5 rem TEDE and 5 rem whole body are not 
equivalent, so they cannot be compared directly. I&M has reanalyzed 
the loss of coolant accident (LOCA) and non-LOCA events to determine 
the limiting condition for control room dose using the AST. The 
calculated dose for all the analyzed events meets the acceptance 
criterion for GDC-19 as described in 10 CFR 50.67. Therefore, the 
consequences are not significantly increased.
    The [control room emergency ventilation system] CREVS is 
designed to mitigate the consequences of an accident. It is not 
assumed to operate in the pressurization mode until after an 
accident has occurred. The system itself has no impact on the 
initiation of any evaluated accidents. Therefore, the changes to the 
CREVS requirements do not increase the probability of an accident 
previously evaluated.
    The proposed changes to the CREVS requirements do not affect the 
ability to maintain a control room pressure boundary. The changes 
ensure that the control room will be pressurized following an 
accident where the CREVS is required to operate to minimize 
unfiltered inleakage. The proposed 24-hour allowed outage time and 
subsequent shutdown action are consistent with the requirements for 
an inoperable filter unit and are reasonable due to the low 
probability of the initiation of an accident requiring actuation of 
the CREVS occurring when the pressure boundary is inoperable. 
Control room dose is significantly increased with increased 
unfiltered inleakage. Specifying the test condition in the 
surveillance allows increases in unfiltered inleakage to be 
identified and evaluated. Preserving the control room pressure 
boundary provides assurance that the consequences of an accident 
previously evaluated are not significantly increased.
    The proposed applicability and action requirements during the 
movement of irradiated fuel assemblies ensure the CREVS is operable 
for the protection of control room personnel in the event of a fuel 
handling accident.
    The proposed changes to the Limiting Conditions for Operation 
(LCO) address redundant dampers that are being installed. Adding the 
new equipment to the LCO and action requirements ensures all 
components associated with the CREVS are operable or action is taken 
to restore them. The proposed changes do not affect equipment design 
or operation. Therefore, the consequences of accidents previously 
evaluated are not increased.
    The proposed changes for the charcoal testing method affect 
activities in the laboratory only and have no impact on plant 
operation. Sampling and testing charcoal will not initiate an 
accident. The charcoal adsorbers are used to mitigate the 
consequences of an accident and are not operated until after an 
accident has occurred. Therefore, the probability of an accident 
previously evaluated is not affected. Charcoal testing verifies the 
ability of the charcoal adsorbers to function as assumed following 
an accident. The new method for testing the CREVS samples provides 
more accurate and reproducible laboratory results. These results 
provide assurance that the charcoal adsorbers will meet the assumed 
radioiodine removal efficiency following an accident. Therefore, the 
consequences of accidents previously evaluated are not increased.
    The [high-efficiency particular air] HEPA filter/charcoal 
adsorber units in the CREVS, [engineered safety features ventilation 
system] ESFVS, and [storage pool ventilation system] SPVS are 
designed to mitigate the consequences of an accident. They are not 
assumed to operate until after an accident has occurred. The 
adsorber units have no impact on the initiation of any evaluated 
accidents. Therefore, the proposed change to reduce the differential 
pressure does not increase the probability of an accident previously 
evaluated. The proposed change to surveillance requirements to 
reduce the allowable pressure drop across the HEPA filter/charcoal 
adsorber unit ensures the system flow rates can be maintained so 
that the system performs as designed. The change ensures that filter 
units are replaced before airflow is restricted. This allows the 
required area to be pressurized so that unfiltered inleakage remains 
within the amount assumed in the accident analysis. Therefore, the 
proposed revision to reduce the allowable pressure drop requirement 
does not significantly increase the consequences of an accident 
previously evaluated.
    The remaining changes are administrative in nature. The proposed 
editorial changes involve reformatting of the individual T/S pages 
to standardize page appearance and readability and do not alter any 
requirements. The proposed change to separate the CREVS functions 
into individual specifications does not affect the system 
operability requirements or make any changes in how the equipment is 
operated. The separation of the two functions does not affect the 
ability of the CREVS to cool or pressurize the control room 
envelope. The proposed change to incorporate the new laboratory 
testing standard for charcoal adsorbers in the ESFVS and SPVS is 
administrative because the test conditions are consistent with the 
standard referenced in the T/S. These changes are administrative in 
nature and do not affect the probability or consequences of 
accidents previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The use of an AST alone cannot create the possibility of a new 
or different kind of accident. The proposed change to use the AST 
does not make any changes to equipment, procedures, or processes. 
The AST does not create any new accident initiators or precursors. 
It is merely a method used to predict radionuclides released 
following an accident. Therefore, this proposed change does not 
increase the possibility of a new or different kind of accident than 
previously evaluated.
    The CREVS is designed to mitigate the consequences of an 
accident. It is not assumed to operate until after an accident has 
occurred. The proposed LCO requirement to maintain the control room 
envelope/pressure boundary operable and expand the area to include 
the control room heating ventilation and air conditioning equipment 
room and plant process computer room does not affect system design 
or operation. The area defined as the control room envelope includes 
all of the areas that communicate with the control room. A tracer 
gas test confirmed that the defined control room envelope can be 
pressurized to greater than or equal to 1/16 inch of water gauge, as 
assumed in the accident analysis. The proposed surveillance 
requirement to specify a makeup airflow rate of less than or equal 
to 1000 cubic feet per minute (cfm) allows periodic verification 
that the assumed unfiltered inleakage is within the assumptions of 
the accident analysis. The new requirement provides added assurance 
that the pressure boundary is maintained operable. The proposed 
changes to the CREVS requirements do not introduce any new plant 
equipment or new methods of operating the equipment. No new failure 
mechanisms are introduced.
    The proposed change to incorporate the new testing requirements 
of ASTM D3803-1989 is administrative in nature. It affects 
activities in the laboratory only and has no impact on plant 
operation. The change does not affect the method for obtaining the 
charcoal sample. It does not cause any of the ventilation equipment 
to be operated in a new or different manner.
    The change to reduce the allowable pressure drop across the 
pressurization filter train to 4 inches water gauge ensures system 
performance is consistent with design. The revised value is more 
restrictive and provides assurance that the affected components of 
the filter unit are replaced before airflow is reduced to the extent 
that it affects the pressurization capability of the CREVS, ESFVS, 
and SPVS. No new failure mechanism is created.
    The remaining changes are administrative in nature. The proposed 
editorial changes involve reformatting of the individual T/S pages 
to standardize page appearance and readability and do not alter any 
requirements. The proposed change to separate the CREVS functions 
into individual specifications does not affect the system 
operability requirements or make any changes in how the equipment is 
operated. The separation of the two functions does not affect the 
ability of the CREVS to cool or pressurize the control room 
envelope. The proposed change to incorporate the new laboratory 
testing standard for charcoal adsorbers in the ESFVS and SPVS is 
administrative because the test conditions are consistent with the 
standard referenced in the T/S. These changes are administrative in 
nature and do not create the possibility of a new or different kind 
of accident from any previously evaluated.
    Therefore, it is concluded that the proposed changes do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.

[[Page 51358]]

    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change to implement the AST for the revised 
analysis incorporates the guidance for application of the AST 
provided in NUREG-1465 and draft RG-1081. The change involves the 
use of new terminology for the acceptance criterion expressed as 5 
rem TEDE. The term TEDE is defined in 10 CFR 20 as the sum of the 
deep-dose equivalent (for external exposures) and the committed 
effective dose equivalent (for internal exposures). The acceptance 
criteria of 5 rem TEDE and 5 rem whole body are not equivalent. The 
NRC has revised the current GDC-19 whole body dose criterion with a 
criterion in terms of rem TEDE for the duration of the accident in 
10 CFR 50.67 for the licensee that seeks to revise its current 
radiological source term with an AST.
    The NRC recognizes that an analysis using the AST may represent 
a reduction in the margin of safety for some applications. The 
margin of safety is typically defined as the difference between the 
calculated parameters (offsite and control room dose) and the 
associated regulatory or safety limit. Implementing the AST in 
accordance with draft RG-1081 and 10 CFR 50.67 revises the 
acceptance criterion (regulatory limit) contained in GDC-19 to 5 rem 
TEDE. The calculated control room dose is below the new acceptance 
criterion. In 10 CFR 50.67, the rule considers the 5 rem whole body, 
or its equivalent to any part of the body is accounted for in the 
definition of TEDE and by the 5 rem TEDE annual limit. Therefore, 
revising the control room dose analysis using the new terminology 
for the AST does not involve a significant reduction in a margin of 
safety.
    The margin of safety associated with the CREVS T/S is to 
maintain control room dose within the limits of GDC-19. The proposed 
changes to the CREVS requirements ensure that accident analysis 
assumptions are preserved so that the dose limit is met. The 
proposed change for control room envelope/pressure boundary provides 
assurance that positive pressure is maintained in the envelope and 
that unfiltered inleakage is bounded by the accident assumption. 
Adding a test requirement for filtered makeup airflow also supports 
this requirement. The proposed change to expand the applicability 
requirements and actions provides assurance that the CREVS is 
operable during times when an accident could occur that may affect 
the control room environment. The proposed changes that reflect 
addition of the dampers provides assurance that the control room 
pressure boundary will be isolated and the envelope will be 
pressurized when CREVS is actuated following an accident. The 
proposed change to reduce the allowable pressure drop across the 
HEPA filter/charcoal adsorber units provides assurance that the 
CREVS, ESFVS, and SPVS provide the required airflow. This allows 
areas to be pressurized as required.
    The proposed change to incorporate the testing standards 
recommended for the charcoal adsorbers in GL 99-02 provides 
assurance that the charcoal adsorbers will remove radioiodine as 
assumed in the accident analysis. Additional margin is gained by 
applying a safety factor to the iodine removal efficiency assumed in 
the accident analysis. This safety factor applies to CREVS , ESFVS, 
and SPVS. The T/S have also been revised to reflect the iodine 
removal efficiency assumed in the accident analysis. The acceptance 
criterion reflects the analysis assumption and the safety factor.
    The remaining changes are administrative in nature. The proposed 
editorial changes involve reformatting of the individual T/S pages 
to standardize page appearance and readability and do not alter any 
requirements. The proposed change to separate the CREVS functions 
into individual specifications does not affect the system 
operability requirements or make any changes in how the equipment is 
operated. The separation of the two functions does not affect the 
ability of the CREVS to cool or pressurize the control room 
envelope. The proposed change to incorporate the new laboratory 
testing standard for charcoal adsorbers in the ESFVS and SPVS is 
administrative because the test conditions are consistent with the 
standard referenced in the T/S. These changes are administrative in 
nature and do not involve a significant reduction in the margin of 
safety.
    The proposed changes support the control room dose calculations 
that demonstrate that the GDC-19 requirement will be met. Therefore, 
these changes do not involve a significant reduction in the margin 
of safety.
    In summary, based upon the above evaluation, I&M has concluded 
that these changes involve no significant hazards consideration.

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

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

    Date of amendment request: November 30, 1999, as supplemented June 
28, 2000
    Description of amendment request: The licensee proposed to amend 
the unit's Technical Specifications (TS), Section 3.7.2, ``Control Room 
Envelope Filtration (CREF) System'' and Section 5.5.7, ``Ventilation 
Filter Testing Program (VFTP),'' to require testing consistent with 
American Society for Testing and Materials (ASTM) Standard D3803-1989, 
in lieu of the current D3803-1979. This application for amendment is a 
response to the NRC's Generic Letter (GL) 99-02, ``Laboratory Testing 
of Nuclear-Grade Activated Charcoal.'' The licensee's November 30, 
1999, application proposed to amend only the TS that was then in 
effect; the TS was fully overhauled in style and format by Amendment 
No. 91. In anticipation of such overhaul of the TS, the staff declined 
to proceed with review of the November 30, 1999, application. The 
licensee's June 28, 2000, application proposes to amend the TS in its 
current form (i.e., as revised by Amendment No. 91 to the Improved 
Technical Specification format). The licensee's two submittals differ 
only in form and style; the proposed TS requirements and supporting 
analyses in these submittals are identical.
    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 in its November 30, 1999, application. The NRC staff has 
reviewed the licensee's analysis against the standard of 10 CFR 
50.92(c). The NRC staff's review is presented below:

    1. 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 change will require testing the Standby Gas 
Treatment (SGT) System and CREF System charcoal filters in 
accordance with ASTM D3803-1989 versus the current ASTM D3803-1979. 
Neither the SGT nor CREF system is an initiator or precursor to an 
accident previously evaluated; both systems perform mitigative 
functions in response to an accident. Failure of either system would 
result in the inability to perform its mitigative function but no 
failure would increase the probability of an accident. Accordingly, 
changing the test methodology of the charcoal filters will not 
affect any accident precursors. Therefore, the probability of an 
accident previously evaluated is not increased.
    The SGT system is designed to limit the release of radioactive 
gases to the environment within the guidelines of 10 CFR 100 for 
analyzed accidents. The CREF system is designed to limit doses to 
control room operators to less than the values allowed by General 
Design Criterion 19. Both systems contain charcoal filters which 
require laboratory carbon sample analysis be performed in accordance 
with RG 1.52 as required by TS. Charcoal filter samples are tested 
to determine whether the filter adsorber efficiency is greater than 
that assumed in the design basis accident analysis. The proposed TS 
changes to test the charcoal material in accordance with

[[Page 51359]]

ASTM D3803-1989 (versus ASTM D3803-1979) will assure the ability of 
the subject systems to perform their intended function. As long as 
these systems perform their intended functions, there will not be 
any increase in the consequences of an accident previously 
evaluated.
    2. 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.
    The proposed TS change will require testing the SGT and CREF 
charcoal filters in accordance with ASTM D3803-1989 versus ASTM 
D3802-1979. This change will not involve placing these systems in 
new configurations or operating the systems in a different manner 
that could result in a new or different kind of accident. Testing in 
accordance with the ASTM D3803-1989 standard will assure the ability 
of the subject systems to perform their intended function by 
providing a more realistic prediction of the capability of the 
charcoal filters. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. 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 of the SGT or CREF System nor will it 
affect the ability of these systems to perform their intended 
functions. Charcoal filter samples are tested to determine whether 
the filter adsorber efficiency is greater than that assumed in the 
design basis accident analysis. The proposed TS changes to test the 
charcoal material in accordance with ASTM D3803-1989 (versus ASTM 
D3803-1979) will assure the ability of the subject systems to 
perform their intended function by providing a more realistic 
prediction of the capability of the charcoal filters. Also, the 
proposed changes are consistent with the changes recommended in NRC 
GL 99-02. 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. 50-423, Millstone 
Nuclear Power Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: April 19, 2000
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TS) Definition 1.7, ``CONTAINMENT 
INTEGRITY''; Sections 3/4.6.1.1, ``Containment Systems, Primary 
Containment, CONTAINMENT INTEGRITY''; 3/4.6.1.2, ``Containment Systems, 
Primary Containment, Containment Leakage''; 3/4.6.1.3, ``Containment 
Systems, Primary Containment, Containment Air Locks''; 3/4.6.1.6, 
``Containment Systems, Primary Containment, Containment Structural 
Integrity''; 3/4.6.6.3, ``Containment Systems, Secondary Containment 
Structural Integrity''; and 6.8, ``Procedures and Programs.'' The use 
of this option requires the implementation of a program based on 
Regulatory Guide 1.163, ``Performance-Based Containment Leak-Test 
Program,'' and modification of the Technical Specifications to reflect 
this program. The proposed Technical Specifications changes will 
implement a performance-based Containment Leakage Testing Program in 
accordance with 10 CFR Part 50, Appendix J, Option B as a substitute 
for the requirements of 10 CFR Part 50, Appendix J, Option A. The Bases 
for these Technical Specifications will be modified to address the 
proposed changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The changes involved in this license amendment request revise 
the testing criteria for the containment penetrations. The revised 
criteria will be based on the guidance in Regulatory Guide 1.163, 
``Performance-Based Containment Leak-Test Program.'' This guidance 
allows for the use of relaxed testing frequencies for containment 
penetrations that have performed satisfactorily on a historical 
basis.
    The Containment Leakage Rate Testing Program considers the type 
of service, the design of the penetration, and the safety impact of 
the penetration in determining the testing interval of each 
penetration. The [Nuclear Regulatory Commission] Staff has reviewed 
the potential impact of performance-based testing frequencies for 
containment penetrations during the development of the Option B 
regulation. The NRC Staff review is documented in NUREG-1493 
``Performance-Based Containment Leak-Test Program.'' The review 
concluded that reducing the frequency of Type A tests (Integrated 
Leak Rate Tests) from three per ten years to one per ten years leads 
to an imperceptible increase in risk. EPRI Research Project Report 
TR-104285, ``Risk Impact Assessment of Revised Containment Leak Rate 
Testing Intervals,'' also concluded that a relaxation of the test 
intervals for Type B and C penetrations results in a negligible 
increase in total plant risk.
    The use of Option B will allow the extension of testing 
intervals with a minimal impact on the radiological release rates 
since most penetration leakage is continually well below the 
specified limits. In the accident risk evaluation, the NRC Staff 
noted that the accident risk is relatively insensitive to the 
containment leakage rate because the accident risk is dominated by 
accident sequences that result in failure of or bypass of the 
containment. The containment leak rate and component performance 
history at Millstone Unit No. 3 are consistent with the conclusions 
reached in NUREG-1493.
    Therefore, the proposed license amendment adopting a 
performance-based approach for verification of leakage rates for 
isolation valves, containment penetrations, and the containment 
overall will continue to meet the regulatory goal of providing an 
essentially leak-tight containment boundary, and will provide an 
equivalent level of safety as the current requirements.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously analyzed.
    Changes to the Administrative section describe the containment 
testing program only and cannot increase the probability or 
consequences of an accident previously analyzed.
    (2) Create the possibility of a new or different kind of 
accident from any previously analyzed.
    The proposed license amendment does not change the operation of 
the plant. The proposed changes do not involve any physical or 
operational changes to structures, systems or components. No new 
failure mechanisms beyond those already considered in the current 
plant safety analyses are introduced. Since there is no change to 
the equipment or the operation of the plant, there is no possibility 
of creating a new or different kind of accident than previously 
analyzed. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously analyzed.
    Changes to the Administrative section describe the containment 
testing program only and cannot create a different accident from any 
previously analyzed.
    3. Involve a significant reduction in the margin of safety.
    During the development of 10 CFR Part 50, Appendix J, Option B, 
the NRC Staff determined the reduction in safety associated with the 
implementation of the performance-

[[Page 51360]]

based testing program. The results of this review are documented in 
NUREG-1493. The review concluded that reducing the frequency of Type 
A tests (Integrated Leak Rate Tests) from three per ten years to one 
per ten years leads to an imperceptible increase in risk. The use of 
Option B will allow the extension of testing intervals with a 
minimal impact on the radiological release rates since most 
penetration leakage is continually well below the specified limits. 
In the accident risk evaluation, the NRC Staff noted that the 
accident risk is relatively insensitive to the containment leakage 
rate because the accident risk is dominated by accident sequences 
that result in failure of or bypass of the containment. The use of a 
performance-based testing program will continue to provide assurance 
that the accident analysis assumptions remain bounding. Therefore, 
these changes do not involve a significant reduction in the margin 
of safety.
    Changes to the Administrative section describe the containment 
testing program only and cannot reduce the margin of safety.

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

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

    Date of amendment request: April 19, 2000
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 3.7.1.5, ``Plant Systems--Main 
Steam Line Isolation Valves.'' Specifically, the change will remove the 
requirement to perform partial stroke testing of the main steam line 
isolation valves during power operation, modify the TS wording for 
clarity, combine two surveillance requirements into one, and modify the 
associated Bases for consistency.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to Technical Specification 3.7.1.5 will not 
affect the operability requirements for the MSIVs [Main Steam 
Isolation Valves] during plant operation in Modes 1 through 4. If a 
MSIV is not operable, restoration of operability is still required, 
or the valve will be closed. Once closed, the MSIV is performing the 
accident mitigation function.
    The addition of a footnote to allow performance of the full 
valve stroke surveillance requirement when the MSIVs are closed to 
comply with action requirements will allow testing to be performed 
that may be necessary to demonstrate MSIV operability. Since proper 
operation of the MSIVs would be expected when utilizing this 
provision, and this test is used to confirm valve operability, the 
MSIVs should function properly to mitigate an accident.
    The proposed change to remove the requirement to perform partial 
stroke testing of the MSIVs when the plant is in Modes 1 or 2 will 
eliminate a high risk activity that is not necessary to ensure the 
ability of the MSIVs to perform their safety function. Recent valve 
design changes and improvements to the MSIV solenoid valves have 
increased reliability in proper main valve operation. Additionally, 
redundant solenoid valve design precludes a single failure from 
affecting the ability of the main valve to close within the required 
time. Thus, the full stroke test is sufficient to ensure 
operability.
    The proposed changes will have no adverse effect on plant 
operation, or the availability or operation of any accident 
mitigation equipment. The plant response to the design basis 
accidents will not change. In addition, the proposed changes can not 
cause an accident. Therefore, there will be no significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes will not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. The changes do not alter the way 
any structure, system, or component functions and do not adversely 
alter the manner in which the plant is operated. The proposed 
changes do not introduce any new failure modes. The proposed changes 
will reduce the likelihood of a transient by eliminating a high risk 
surveillance. Also, the response of the plant and the operators 
following these accidents is unaffected by the change. Therefore, 
the proposed changes 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.
    The proposed changes modify the LCO [Limiting Condition for 
Operation], applicability, action requirements, and surveillance 
requirements of Technical Specification 3.7.1.5. These changes have 
no adverse effect on equipment important to safety. This equipment 
will continue to function as assumed in the design basis accident 
analyses. The proposed changes will not result in any plant 
configuration changes. There will be no adverse effect on plant 
operation or accident mitigation equipment. The plant response to 
design basis accidents will not change. Therefore, there will be no 
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.

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

    Date of amendment request: April 19, 2000.
    Description of amendment request: The proposed amendment would 
modify Technical Specification Sections 3.8.4.1, ``Electrical Power 
Systems--Containment Penetration Conductor Overcurrent Protective 
Devices''; 3.8.4.2.1, ``Electrical Power Systems--Motor-Operated Valves 
Thermal Overload Protection''; and 3.8.4.2.2, ``Electrical Power 
Systems--Motor-Operated Valves Thermal Overload Protection Not 
Bypassed. The proposed changes will relocate the requirements for 
containment penetration conductor overcurrent and motor-operated valve 
thermal overload protective devices from the Technical Specifications 
to the Technical Requirements Manual (TRM). The Bases for these 
Technical Specifications will be modified to address the proposed 
changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to relocate the requirements for 
containment penetration conductor overcurrent and motor-operated 
valve thermal overload protective devices from Technical 
Specifications to the TRM will have no adverse effect on plant 
operation, or the availability or operation of any accident 
mitigation equipment. The plant response to the design basis 
accidents will not change. Operation of the containment penetration 
conductor overcurrent and motor-operated valve thermal overload 
protective devices are not accident initiators and cannot cause an

[[Page 51361]]

accident. Whether the requirements for the containment penetration 
conductor overcurrent and motor-operated valve thermal overload 
protective devices are located in Technical Specifications or the 
TRM will have no effect on the probability or consequences of any 
accident previously evaluated. Therefore, there will be no 
significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed changes to relocate the requirements from Technical 
Specifications to the TRM will not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. The proposed changes will not 
introduce any new failure modes that could result in a new accident. 
Also, the response of the plant and the operators following the 
design basis accidents is unaffected by the changes. Therefore, the 
proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Involve a significant reduction in a margin of safety.
    The proposed changes will relocate the requirements for 
containment penetration conductor overcurrent and motor-operated 
valve thermal overload protective devices from Technical 
Specifications to the TRM. Any future changes to the relocated 
requirements will be in accordance with 10 CFR 50.59 and approved 
station procedures. The proposed changes will have no adverse effect 
on plant operation, or the availability or operation of any accident 
mitigation equipment. The plant response to the design basis 
accidents will not change. In addition, the relocated requirements 
do not meet any of the 10 CFR 50.36c(2)(ii) criteria on items for 
which Technical Specifications must be established. Therefore, there 
will be no significant reduction in the margin of safety.

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

Northern States Power Company, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota.

    Date of amendment request: July 18, 2000.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications to add operability requirements for 
the No. 12 residual heat removal service water (RHRSW) pump.
    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 12 RHRSW Pump is not an accident (fire) initiator. During a 
fire in the Control Room or Cable Spreading Room, the ASDS 
[alternate shutdown system] panel provides alternate shutdown 
capability. The proposed amendment provides operability requirements 
to ensure 12 RHRSW Pump is available when alternate shutdown is 
required so that safe shutdown can be achieved and maintained in 
accordance with existing procedures. The proposed operability 
requirements are consistent with previous ASDS requirements for 12 
RHRSW Pump and other equipment required for alternate shutdown. Dose 
to the public and the Control Room operators are not affected by the 
proposed change.
    The proposed Technical Specification change does not introduce 
new equipment operating modes, nor does the proposed change alter 
existing system relationships. The proposed amendment does not 
introduce new failure modes.
    Therefore, the proposed amendment will not significantly 
increase the probability or the 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 
analyzed.
    The proposed Technical Specification change does not introduce 
new equipment operating modes, nor does the proposed change alter 
existing system relationships. The proposed amendment does not 
introduce new failure modes.
    Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The proposed amendment is within current Technical Specification 
requirements for other equipment required for alternate shutdown and 
ensures that 12 RHRSW Pump will be available for alternate shutdown 
when required. The allowed ASDS outage time for 12 RHRSW Pump is 
consistent with that allowed for other alternate shutdown equipment. 
The proposed amendment maintains margins of safety. Therefore, the 
proposed amendment will not involve a significant reduction in the 
margin of safety.

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

Northern States Power Company, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota.

    Date of amendment request: July 20, 2000.
    Description of amendment request: The proposed amendment would (1) 
revise the Technical Specifications (TSs) to include the automatic 
reactor water cleanup (RWCU) system isolation feature, (2) restore the 
dose equivalent iodine-131 (DEI) limit to 2 microcuries per gram, (3) 
change the RWCU reactor water level automatic isolation signal from Low 
to Low-Low reactor water level, add TSs for the high pressure coolant 
injection (HPCI) and reactor core isolation cooling (RCIC) low steam 
line pressure isolation instrumentation, (4) delete the HPCI 150,000 
lb/hr low range high flow isolation instrumentation and add a time 
delay to the 300,000 lb/hr upper range high flow isolation 
instrumentation, and (5) change the suppression chamber water allowable 
water level from volume units to level units.
    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 proposed new limiting conditions for operation and 
surveillance requirements for the RWCU high system flow and room 
temperature signals and the increase in the allowable reactor 
coolant DEI are administrative in nature and do not involve an 
increase in the probability or consequences of a previously 
evaluated accident. The RWCU automatic isolation instrumentation 
will reduce the consequences of a break in the RWCU System allowing 
the reactor coolant DEI to be increased, consistent with the value 
assumed in the Monticello USAR [updated safety analysis report] for 
the design basis MSLB [main steam line break] safety analysis.
    Changing the reactor water level RWCU automatic isolation 
setpoint to Low Low Reactor Water Level will not increase the 
probability or consequences of a break in the RWCU system because 
new high flow and high area temperature instrumentation provides an 
improved capability for isolating

[[Page 51362]]

a RWCU break independent of changes in reactor water level.
    Changes to the HPCI steam supply line automatic isolation 
instrumentation will not increase the probability or consequences of 
a break in the HPCI steam line. Elimination of the 150,000 lb/hr 
isolation signal will improve the reliability of the HPCI system. 
The remaining 300,000 lb/hr delay high flow isolation signal 
provides more than adequate protection for a steam line break in 
this system. The consequences of a break in the HPCI remain bounded 
by the MSLB safety analysis.
    Adding a description of the Group 3 logic and recirc 
[recirculation] sample valves isolation in the Bases; adding LCOs 
[limiting conditions for operation] and Surveillance Requirements 
for the HPCI and RCIC low steam line pressure isolation logic; and 
changing the method of describing the allowable suppression pool 
water inventory from volume to level are all administrative changes 
than [sic] cannot adversely affect the consequences of any evaluated 
accident.
    The proposed changes do not present the opportunity for a new 
release path for radioactive material.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    No system, structure, or component (SSC) described in the USAR 
as important to safety is adversely affected by these changes. No 
new type of credible event could be identified which would be 
created by the proposed Technical Specification changes. Nothing was 
identified in these changes which could create the possibility for a 
new or different kind of accident.
    The RWCU line break accident was previously analyzed. However, 
non-conservative values for mass and energy release were used. New, 
more conservative, calculations prompted the prudent installation of 
an automatic RWCU break isolation system at Monticello. The RWCU 
line break outside of containment is once again bounded by 
previously analyzed design basis MSLB accident.
    Changes to the HPCI steam line high flow instrumentation will 
improve the reliability of the HPCI system, while continuing to 
provide a high degree of protection for a break in the HPCI steam 
supply line.
    Addition of LCOs and Surveillance Requirements for the HPCI and 
RCIC low steam line pressure isolation and changing the way in which 
suppression pool level is specified in the Technical Specifications 
are administrative changes that cannot result in a new or different 
kind of accident than any previously analyzed.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The margin of safety for the RWCU line break can be expressed in 
terms of the offsite and control room doses which could result from 
this event. The previous RWCU line break analysis relied on operator 
action to isolate the line break after an assumed delay of 10 
minutes. The new RWCU automatic isolation instrumentation initiates 
isolation of a RWCU line break in less than 27 seconds. The new 
isolation logic limits the potential offsite radiological 
consequences from a RWCU line break to a fraction of the bounding 
MSLB accident. The proposed Technical Specification would 
incorporate LCOs and Surveillance Requirements for the new isolation 
logic and change the reactor water level setpoint to a value which 
helps prevent unnecessary isolation of the system following reactor 
scrams. Margins of safety are improved by these changes.
    Proposed changes to the HPCI steam line high flow 
instrumentation will improve the reliability of the HPCI system and 
increase existing margins of safety for accidents in which HPCI 
operation is credited. The margin of safety for a HPCI steam line 
break can also be expressed in terms of the offsite and control room 
doses which could result from this event. The modified HPCI 
isolation logic limits the potential offsite radiological 
consequences from a HPCI steam line break to a fraction of the 
bounding MSLB accident.
    Adding a description of the Group 3 recirc sample isolation to 
the Bases, adding LCOs and Surveillance Requirements for the HPCI 
and RCIC low steam line pressure isolation, and changing the method 
of describing the allowable suppression pool water inventory are 
administrative changes. No significant changes in plant equipment or 
plant operation will occur and no equipment important to safety is 
affected as a result of these changes. No margin of safety is 
therefore affected.

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

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

    Date of amendment request: July 27, 2000
    Description of amendment request: The amendment would delete a note 
to Technical Specification Section 12.1.A to allow the Safety Limit 
Minimum Critical Power Ratio (SLMCPR) to be applicable beyond cycle 14. 
The amendment would also revise the reference to the General Electric 
Standard Application for Reactor Fuel (GESTAR) document in Section 
6.9.a.4 to incorporate the latest revision.
    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:

    Operation of the FitzPatrick plant in accordance with the 
proposed amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Deletion of a note stating that the SLMCPR remains applicable 
through Cycle 14 does not affect the initiation of any accident. 
Operation in accordance with the current SLMCPR ensures the 
consequences of previously analyzed accidents are not changed. 
Therefore, this proposed change 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 SLMCPR establishes a performance limit for the fuel. This 
limit remains unchanged. Deleting a note to reflect this is an 
administrative change and will not initiate any accident. Therefore, 
this proposed change 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.
    GE [General Electric] has performed an evaluation of the SLMCPR 
for Cycle 15 and found that the cycle specific value, based on 
current reload plans, is bounded by the generic value calculated for 
GE 12 fuel. The existing SLMCPR remains unchanged for Cycle 15 and 
the margin of safety for the prevention of onset of transition 
boiling is unchanged. Therefore, this proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: Marsha K. Gamberoni

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

    Date of amendment requests: July 20, 2000 (PCN-488, Supplement 1). 
This application supersedes the licensee's application of August 11, 
1999.
    Description of amendment requests: The U.S. Nuclear Regulatory 
Commission (the Commission) has granted the request of Southern 
California Edison Company to withdraw

[[Page 51363]]

its August 11, 1999, application for proposed amendments. The 
Commission had previously issued a Notice of Consideration of Issuance 
of Amendments published in the Federal Register on September 8, 1999 
(64 FR 48866). However, by letter dated July 20, 2000, the licensee 
withdrew the proposed change. TAC Nos. MA6282 and MA6283 used for the 
review of the August 11, 1999, application have been closed.
    As submitted by the licensee on July 20, 2000, the proposed 
amendments would modify the Technical Specifications for the San Onofre 
Nuclear Generating Station (SONGS) Units 2 and 3 to revise Surveillance 
Requirement (SR) 3.3.7.3 by providing allowable values in place of 
analytical limits for certain degraded voltage parameters, and by 
deleting unnecessary parameter limits in cases where plant safety is 
not affected. The proposed change would also delete redundant SR 
3.3.7.4.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the proposed amendment involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    No. Proposed Change Number (PCN)-488, Supplement 1, revises the 
Technical Specification (TS) Surveillance Requirement (SR) 
acceptance criteria of the Loss of Voltage Signal (LOVS), Degraded 
Grid Voltage with Safety Injection Actuation Signal (DGVSS), and 
Sustained Degraded Voltage Signal (SDVS) relay circuits. These 
circuits are not accident initiators.
    PCN-488 Supplement 1 revises the TS SR acceptance requirements 
to make them more limiting than the present requirements. Because 
the revised acceptance criteria are more limiting than the present 
requirements, the consequences of accidents analyzed in the Updated 
Final Safety Analysis Report (UFSAR) are not increased. PCN-488 
Supplement 1 also revises the TS SR acceptance requirements to 
delete or revise upper and lower bounds in cases where the deleted 
bound provides no safety benefit. Deleting or revising bounds having 
no safety significance does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    PCN-488 Supplement 1 deletes redundant SR 3.3.7.4, which is not 
in NUREG-1432, Standard Technical Specifications, Combustion 
Engineering Plants. Deleting a redundant requirement does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    Consequently, the proposed amendment does not result in an 
increase in the probability or consequences of accidents evaluated 
in the UFSAR.
    (2) Does this amendment request create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. PCN-488 Supplement 1 revises the TS SR acceptance criteria 
of the LOVS, DGVSS, and SDVS relay circuits, which are not accident 
initiators, and deletes a redundant SR. PCN-488 Supplement 1 does 
not introduce any revision in the hardware configuration of the 
protective circuitry for LOVS, DGVSS or SDVS. The measurement 
required by the deleted, redundant surveillance is required 
elsewhere in the TS. For these reasons, PCN-488 Supplement 1 does 
not create the possibility of any new or different kind of accident 
from any previously evaluated.
    (3) Does this amendment request involve a significant reduction 
in a margin of safety?
    No. PCN-488 Supplement 1 provides allowable values for the 
acceptance criteria for the TS SR for LOVS, DGVSS and SDVS. As such, 
the revised values are more limiting than the current values, which 
represent design limits. Therefore, PCN-488 Supplement 1 does not 
involve a significant reduction in a margin of safety.
    PCN-488 Supplement 1 also revises the TS SR acceptance 
requirements to delete or revise upper and lower bounds in cases 
where the deleted bound provides no safety benefit. Deleting or 
revising bounds having no safety significance does not involve a 
significant reduction in a margin of safety.
    PCN-488 Supplement 1 additionally deletes a redundant SR. 
Because the deleted surveillance is required elsewhere in the TS, 
this action does not involve a significant reduction in a margin of 
safety.
    For these reasons, PCN-488 Supplement 1 does not involve a 
significant reduction in a margin of safety.

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

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: August 10, 2000.
    Brief description of amendments: The proposed change would revise 
Technical Specification 5.6.5 entitled CORE OPERATING LIMITS REPORT. 
TXU Electric proposes to revise the Large Break Loss of Coolant 
Accident methodology used at Comanche Peak Steam Electric Station, 
Units 1 and 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 proposed change involves an administrative change only. 
Designation of the Revised Large Break Loss of Coolant Accident 
analysis methodology, described in ERX-2000-002-P, as the approved 
Large Break Loss of Coolant Accident analysis methodology is 
required to maintain the accuracy of the Technical Specification 
5.6.5 (Core Operating Limits Report) and to maintain consistency 
with the resolution of issues as prescribed in 10 CFR 50.46.
    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 proposed change involves an administrative change only. 
Technical Specification 5.6.5, Item 15, is being changed to 
reference the revised Large Break Loss of Coolant Accident analysis 
methodology currently under NRC review. No actual plant equipment 
will be affected by the proposed change. An analysis for Unit 1, 
Cycle 8, is imbedded in the referenced Topical Report, from which it 
is concluded that no failure modes, not bounded by previously 
evaluated accidents, will be created.
    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
    Margin of safety is associated with the 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. This 
request involves an administrative change (subject to NRC approval 
of the revised Large Break Loss of Coolant Accident Analysis 
methodology) only to incorporate the revised Large Break Loss of 
Coolant Accident analysis methodology into the allowable analysis 
methodologies specified in Technical Specification 5.6.5.
    No actual plant equipment will be affected by the proposed 
change. The compliance of the revised methodology with the 
requirements of 10 CFR 50.46 and Appendix K will be addressed 
through the NRC staff's review of the topical report. Therefore, it 
is concluded that the use of the proposed methodology will not 
degrade the confidence in the ability of the fission product 
barriers to limit the level of radiation dose to the public.


[[Page 51364]]


    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

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

    Date of application request: July 21, 2000 (ULNRC-04285)
    Description of amendment request: The proposed amendment would 
revise Limiting Condition for Operation (LCO) 3.9.4, ``Containment 
Penetrations,'' of the Callaway Technical Specifications (TS) to allow 
containment penetrations with direct access to the outside atmosphere 
to be open under administrative controls during refueling operations, 
by adding a note to the LCO that states ``containment penetration flow 
path(s) providing direct access from the containment atmosphere to the 
outside atmosphere may be unisolated under administrative controls.'' 
In addition, there would be a format and editorial correction to TS 
3.8.3, ``Diesel Fuel Oil, Lube Oil, and Start Air,'' to correct an 
error in the conversion to the improved TS issued May 28, 1999, in 
Amendment No. 133. There are also revisions to the TS Bases for the 
proposed changes to LCO 3.9.4.
    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 status of the penetration flow paths during refueling 
operations has no [effect] on the probability of the occurrence of 
any accident previously evaluated. The proposed revision does not 
alter any plant equipment or operating practices in such a manner 
that the probability of an accident is increased. Since the 
consequences of a FHA [fuel handling accident] inside containment 
with open penetration flow paths are bounded by the current analysis 
described in the FSAR [Callaway Final Safety Analysis Report] and 
the probability of an accident is not affected by the status of the 
penetration flow paths, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes to correct editorial/format errors involve 
corrections to the technical specifications that are associated with 
the original conversion application and supplements or the certified 
copy of the Improved Technical Specifications. As such, these 
changes are considered as administrative changes and do not modify, 
add, delete, or relocate any technical requirements in the technical 
specifications.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The open containment penetration flow paths are not accident 
initiators and do not represent a significant change in the 
configuration of the plant. The proposed allowance to open the 
containment penetrations during refueling operations will not 
adversely affect plant safety functions or equipment operating 
practices such that a new or different accident could be created.
    The proposed changes to correct editorial/format errors involve 
corrections to the technical specifications that are associated with 
the original conversion application and supplements or the certified 
copy of the improved Technical Specifications. As such, these 
changes are considered as administrative changes and do not modify, 
add, delete, or relocate any technical requirements [in] the 
technical specifications.
    Therefore, the proposed revision will not create a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Technical Specification LCO 3.9.4 closure requirements for 
containment penetrations ensure that the consequences of a 
postulated FHA inside containment during core alterations or 
irradiated fuel handling activities are minimized. The LCO 
establishes containment closure requirements, which limit the 
potential escape paths for fission products by ensuring that there 
is at least one integral barrier to the release of radioactive 
material. The proposed change to allow the containment penetration 
flow paths to be open during refueling operations under 
administrative controls does not significantly affect the expected 
dose consequences of a FHA because the limiting FHA is not changed. 
The proposed administrative controls provide assurance that prompt 
closure of the penetration flow paths will be accomplished in the 
event of a FHA inside containment thus minimizing the transmission 
of radioactive material from the containment to the outside 
environment. Under the proposed TS change, the provisions to 
promptly isolate open penetration flow paths provide assurance that 
the offsite dose consequences of a FHA inside containment will be 
minimized.
    The proposed changes to correct editorial/format errors involve 
corrections to the technical specifications that are associated with 
the original conversion application and supplements or the certified 
copy of the Improved Technical Specifications. As such, these 
changes are considered as administrative changes and do not modify, 
add, delete, or relocate any technical requirements in the technical 
specifications.
    Therefore, the proposed changes to the Technical Specifications 
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: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Section Chief: Stephen Dembek

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

    Date of amendment request: March 10, 2000
    Description of amendment request: The proposed amendments would 
implement a Pressure and Temperature Limits Report (PTLR) concurrent 
with the implementation of the Improved Standard Technical 
Specifications. NRC Generic Letter 96-03 provides guidance for 
licensees allowing relocation of the reactor coolant system pressure 
temperature limit curves and low temperature overpressure protection 
system limits from the Technical Specifications to a PTLR.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    The proposed changes relocate the pressure-temperature limits 
and low temperature overpressure protection limits from the 
Technical Specifications to a Pressure Temperature Limits Report 
(PTLR). The proposed changes also provide revised pressure-
temperature limits and revised low temperature overpressure 
protection limits. Appropriate design and safety limits are retained 
in the Specifications, thereby meeting the requirements of 10 CFR 
50.36. Specific, approved methodologies used to determine and 
evaluate the parameter requirements are added to the Specifications 
and a reporting requirement is added to

[[Page 51365]]

ensure the NRC is apprised of all changes. Operation of the PBNP 
will continue to meet all design and safety analysis requirements 
because approved methodologies are required to be used to evaluate 
and change parameters, and appropriate safety and design limits 
maintained in the Technical Changes.
    Therefore, neither the probability nor consequences of an 
accident previously evaluated can be increased.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendment does not create a new or different kind of 
accident from any accident previously evaluated.
    Operation of PBNP, in accordance with the proposed changes, will 
continue to meet all design and safety limits. Appropriate design 
and safety limits continue to be controlled within the Technical 
Specifications as they are presently. These changes will not result 
in a change to the design and safety limits under which PBNP 
operation has been determined to be acceptable. These changes cannot 
result in a new or different kind of accident from any accident 
previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendment does not result in a significant reduction in 
a margin of safety.
    Appropriate safety limits continue to be controlled by the 
Specifications. Changes to the relocated pressure-temperature and 
low temperature overpressure protection limits will be accomplished 
using NRC approved methodologies, thereby ensuring operation will 
continue within the bounds of the existing safety analyses including 
all applicable margins of safety. Therefore, operation in accordance 
with the proposed changes cannot result in a reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Section Chief: Claudia M. Craig

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.

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: February 1, 2000, as supplemented on 
June 1 and July 13, 2000
    Description of amendment request: The proposed amendment would 
change the Technical Specification and Bases Sections associated with 
the requirements for the Reactor Coolant System (RCS) loops and 
Shutdown Cooling (SDC) System trains during all modes of plant 
operation. Many of the proposed changes are associated with the format 
and structure of the affected Technical Specifications and will not 
result in any technical changes to the current requirements. The 
proposed format changes will result in Technical Specifications that 
will be clear, concise, and easier for the control room operators to 
use. Some of the changes are proposed to achieve consistency with the 
Standard Technical Specifications for Combustion Engineering Plants in 
NUREG-1432, Rev. 1. The Bases for the Technical Specifications would 
also be revised to reflect the proposed changes.
    Date of publication of individual notice in Federal Register: July 
31, 2000 (65 FR 46748)
    Expiration date of individual notice: August 31, 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, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

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

    Date of application for amendment: June 19, 2000 (U-603378)
    Brief description of amendment: The amendment changes the leak rate 
test frequency for the primary containment feedwater penetrations 
sealed by the Feedwater Leakage Control System.
    Date of issuance: August 11, 2000
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 131
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 3, 2000 (65 FR 
41103)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 11, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 20, 1999, as supplemented 
February 18, April 19, and May 22, 2000.
    Brief description of amendment: The amendment revised the 
calibration frequency of the 4kV (kilovolt)

[[Page 51366]]

Engineered Safeguards Bus Undervoltage Relays (Diesel Start) (item 43.a 
of Table 4.1-1 of the Technical Specifications (TSs)) from a refueling 
interval to annually. The TS Bases have also been changed to reflect 
that the degraded voltage relay setpoint tolerance is being changed 
from an ``as left'' to an ``as found'' reading. Additionally, the 
amendment approves a revision to the Updated Final Safety Analysis 
Report (UFSAR) to allow for manual operator action for voltage 
protection rather than full automatic voltage protection. These changes 
are reflected in the revised UFSAR pages 8.2-3 and 8.2-5.
    The amendment also adds new TSs 3.7.2.a(ii) and 3.7.2.h to address 
voltage on the 230 kV grid as a precondition of criticality and to 
provide a time limit for when the 230 kV grid voltage is found to be 
insufficient to support loss-of-coolant accident electrical loading 
during power operation. Various minor editorial changes have also been 
made. The Bases have also been changed to reflect the addition of the 
two new TSs and to provide clarification of the components to which 
surveillance is applicable.
    Date of issuance: August 3, 2000
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 224
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 1, 1999 (64 FR 
67334) and June 2, 2000 (65 FR 35404).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 3, 2000.
    No significant hazards consideration comments received: No

    Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson 
Steam
    Electric Plant, Unit No. 2, Darlington County, South Carolina.
    Date of application for amendment: June 14, 2000, as supplemented 
July 14, 2000.
    Brief description of amendment: The amendment revises Technical 
Specification 5.6.5 to incorporate analytical methodologies that are 
used for the Core Operating Limits Report that have been accepted by 
the Nuclear Regulatory Commission for referencing in licensing in 
cycle-specific applications.
    Date of issuance: August 3, 2000.
    Effective date: August 3, 2000.
    Amendment No.: 188.
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 28, 2000 (65 FR 
39957).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 3, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: January 11, 2000.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) to increase allowable out-of-service 
times (AOTs) and surveillance test intervals (STIs) for selected 
actuation instrumentation. The amendments implement AOT/STI changes 
based on Topical Reports by General Electric Company and the Boiling 
Water Reactor Owners' Group which have previously been reviewed and 
approved by NRC.
    Date of issuance: August 2, 2000.
    Effective date: Immediately, to be implemented within 120 days.
    Amendment Nos.: 177 and 173.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 8, 2000 (65 FR 
12290).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 2, 2000.
    No significant hazards consideration comments received: No.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York.

    Date of application for amendment: November 18, 1999, incorporating 
supporting analyses provided by letter dated October 8, 1999, as 
supplemented by letters dated February 14, March 21, April 6, April 13, 
and May 11, 2000.
    Brief description of amendment: The proposed amendment would remove 
the requirement for charcoal filters and high efficiency particulate 
filters in the containment fan cooler system, revise the time 
requirement for subcriticality prior to core alterations from 174 hours 
to 100 hours, revise flow rate requirements for containment fan coolers 
and control room ventilation units to be consistent with the design 
basis, state that the control room ventilation system, in the post-
accident mode, will be operated with filtered intake of outside air, 
allow containment personnel access doors to be open during refueling 
operations, and allow an administrative substitution of ``monthly'' in 
place of ``every 31 days'' in various surveillance requirements.
    Date of issuance: July 27, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 211.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 20, 2000 (65 FR 
3256).
    The February 14, March 21, April 6, April 13, and May 11, 2000, 
submittals contained supplemental information that did not change the 
original no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 27, 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: November 29, 1999, as 
supplemented December 20, 1999.
    Brief description of amendment: The amendment added a license 
condition authorizing a one-time extension of the steam generator 
inspection interval to permit the next inspection to coincide with the 
next scheduled refueling outage.
    Date of issuance: August 4, 2000.
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment No: 112.
    Facility Operating License No. NPF-73. Amendment revised the 
License.
    Date of initial notice in Federal Register: April 5, 2000 (65 FR 
17915). The December 20, 1999, letter provided additional information 
that did not change the initial proposed no significant hazards 
consideration determination or expand the amendment beyond the scope of 
the initial notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 4, 2000.
    No significant hazards consideration comments received: No.


[[Page 51367]]



GPU Nuclear Corporation and Saxton Nuclear Experimental Corporation, 
Docket No. 50-146, Saxton Nuclear Experimental Facility (SNEF), Bedford 
County, Pennsylvania.

    Brief description of amendment: The amendment changes the Technical 
Specification organizational and administrative controls for the SNEF 
to reflect changes in GPU Nuclear following the sale of the Oyster 
Creek Nuclear Generating Station.
    Date of Issuance: August 10, 2000.
    Effective date: The license amendment is effective as of its date 
of issuance.
    Amendment No.: 16.
    Amended Facility License No. DPR-4: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 28, 2000 (65 FR 
39956). The Commission's related evaluation of the amendment is 
contained in a safety evaluation dated August 10, 2000.
    No significant hazards consideration comments received: No.

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey.

    Date of application for amendment: November 5, 1999, as 
supplemented by two letters dated April 6, 2000, and April 13, 2000.
    Brief description of amendment: These amendments conform the 
license to reflect the transfer of Operating License No. DPR-16 for the 
Oyster Creek Nuclear Generating Station.
    Date of Issuance: August 8, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 213.
    Facility Operating License No. DPR-16: Amendment revised the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 16, 1999 (64 
FR 70292).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated June 6, 2000.
    The April 6 and April 13, 2000, supplements did not change the 
initial proposed no significant hazards consideration determination and 
were within the scope of the initial application as originally noticed.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 6, 2000, as supplemented 
by letters dated May 25, 2000, July 18, 2000, and August 8, 2000.
    Brief description of amendment: The amendment reflects an 
administrative name change from GPU Nuclear Corporation to GPU Nuclear, 
Inc. Furthermore, the license amendment makes an editorial change to 
better describe TMI-2's use of site physical security, guard training 
and qualification, and safeguard contingency plans that are maintained 
by the Three Mile Island Nuclear Station, Unit 1, licensee, AmerGen 
Energy Company, LLC. In addition, the licensee requested that minor 
changes (mainly in titles) be made in Section 6.0 of the Technical 
Specifications to reflect the TMI-2 organizational and administrative 
controls that will exist following the sale of the Oyster Creek Nuclear 
Generating Station, which occurred August 8, 2000. The May supplement 
provided a response to a staff request for additional information, and 
the July and August supplements related to the requested effective date 
of the amendment. The supplements did not expand the scope of the 
application, as noticed in the Federal Register (65 FR 21484, dated 
April 21, 2000), or change the proposed no significant hazards 
consideration determination.
    Date of issuance: August 9, 2000.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment No.: 54.
    Facility Operating License No. DPR-73: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 21, 2000 (65 FR 
21484).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 9, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 24, 1999, as 
supplemented February 4 and March 17, 2000.
    Brief description of amendment: The amendment conforms the license 
to reflect the transfer of operating authority under Operating License 
No. DPR-49 to Nuclear Management Company, LLC, as approved by order of 
the Commission dated May 15, 2000.
    Date of issuance: August 7, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 232.
    Facility Operating License No. DPR-49: The amendment revised the 
Operating License.
    Date of initial notice in Federal Register: February 4, 2000 (65 FR 
5703).
    The February 4 and March 17, 2000, supplements were within the 
scope of the initial application as originally noticed. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated May 15, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 7, 1999
    Brief description of amendment: This amendment removes the action 
requirement to suspend all operations involving positive reactivity 
additions from Technical Specification (TS) 3.4.2.1, ``Reactor Coolant 
System--Safety Valves,'' TS 3.4.2.2, ``Reactor Coolant System--Safety 
Valves,'' and TS 3.7.6.1, ``Plant Systems--Control Room Emergency 
Ventilation System.'' The associated Bases have also been revised.
    Date of issuance: August 7, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 248.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4285).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 7, 2000.
    No significant hazards consideration comments received: No.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of application for amendment: November 24, 1999, as 
supplemented February 2, 2000.
    Brief description of amendment: The amendment conforms the license 
to reflect the transfer of operating authority under Operating License 
No. DPR-22 to Nuclear Management Company, LLC, as approved by order of 
the Commission dated May 15, 2000.
    Date of issuance: August 7, 2000
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 110.

[[Page 51368]]

    Facility Operating License No. DPR-22. Amendment revised the 
Operating License.
    Date of initial notice in Federal Register: February 15, 2000 (65 
FR 7574)
    The February 2, 2000, supplement was within the scope of the 
initial application as originally noticed. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
May 15, 2000.
    No significant hazards consideration comments received: No

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, and Docket No. 72-10, 
Prairie Island Independent Spent Fuel Storage Installation, Goodhue 
County, Minnesota

    Date of application for amendments: November 24, 1999, as 
supplemented February 2, 2000.
    Brief description of amendment: The amendments conform the licenses 
to reflect the transfer of operating authority under Operating License 
Nos. DPR-42 and DPR-60 and Materials License No. SNM-2506 to Nuclear 
Management Company, LLC, as approved by order of the Commission dated 
May 15, 2000.
    Date of issuance: August 7, 2000
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 153 and 144
    Facility Operating License Nos. DPR-42 and DPR-60 and Materials 
License No. SNM-2506: Amendments revised the Operating Licenses and 
Materials License.
    Date of initial notice in Federal Register: February 15, 2000 (65 
FR 7574).
    The February 2, 2000, supplement was within the scope of the 
initial application as originally noticed. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
May 15, 2000.
    No significant hazards consideration comments received: No

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

    Date of application for amendment: March 1, 1999, as supplemented 
October 1, and October 6, 1999, and June 6, 2000.
    Brief description of amendment: The amendment supports the 
installation of a digital Power Range Neutron Monitoring system and the 
incorporation of the long-term thermal-hydraulic stability solution 
hardware.
    Date of issuance: August 1, 2000
    Effective date: Effective as of date of issuance and shall be 
implemented prior to restart from the Peach Bottom Atomic Power 
Station, Unit 2, Fall 2000 refueling outage.
    Amendment No.: 232
    Facility Operating License No. DPR-44: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 2, 1999 (64 FR 
29711). The October 1, and October 6, 1999, and June 6, 2000 submittals 
provided clarifying information that did not expand the scope of the 
original Federal Register notice or change the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 1, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 1, 1999.
    Brief description of amendments: The amendments revise the minimum 
fuel oil level for the diesel generator day tanks in Surveillance 
Requirement 3.8.1.3 and revise the acceptable fuel oil level storage 
band in Required Action Statement B of Limiting Condition for Operation 
3.8.3.
    Date of issuance: July 27, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 221 and 162.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 17, 1999 (64 
FR 62715).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 27, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 30, 1999.
    Brief description of amendments: The amendments revised the 
Technical Specifications 5.2.2, ``Unit Staff'', to raise the level of 
the approval authority for deviations above the guidelines provided to 
minimize unit staff overtime. Specifically, the amendments change the 
level of overtime approval authority from ``department superintend'' to 
``department manager.''
    Date of issuance: August 10, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 113 and 91.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 9, 2000 (65 FR 
6410).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 10, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 18, 2000.
    Brief description of amendments: These amendments revise Technical 
Specification (TS) Section 5.3, ``Design Features--Reactor Core,'' and 
TS Section 6.9, ``Administrative Controls--Reporting Requirements'' to 
identify M5 alloy as a material used in the construction of fuel 
assemblies and to cite the topical report that describes the fuel.
    Date of issuance: July 31, 2000.
    Effective date: July 31, 2000.
    Amendment Nos.: 258 and 249.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TSs.
    Date of initial notice in Federal Register: April 5, 2000 (65 FR 
17920).
    The Commission's related evaluation of the amendment is contained 
in an Environmental Assessment dated April 10, 2000, and in a Safety 
Evaluation dated July 31, 2000.
    No significant hazards consideration comments received: No.


[[Page 51369]]



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

    Date of application for amendments: June 30, 1999, as supplemented 
June 16 and August 3, 2000.
    Brief description of amendments: Updates Technical Specification 
(TS) requirements, and appropriate TS Bases sections for reactor 
coolant system (RCS) leakage detection and RCS operational leakage 
specifications to be consistent with the Improved Westinghouse Standard 
TS (NUREG-1431).
    Date of issuance: August 4, 2000.
    Effective date: August 4, 2000.
    Amendment Nos.: 259 and 250.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TS.
    Date of initial notice in Federal Register: October 20, 1999 (64 FR 
56533).
    The June 16 and August 3, 2000, letter provided clarifying 
information and changes that did not change the initial proposed no 
significant hazards consideration determination or expand the 
application beyond the scope of the original notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 4, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 24, 1999, as 
supplemented January 31, 2000.
    Brief description of amendments: The amendments conform the 
licenses to reflect the transfer of operating authority under Operating 
License Nos. DPR-24 and DPR-27 to Nuclear Management Company, LLC, as 
approved by order of the Commission dated May 15, 2000.
    Date of issuance: August 7, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 197 and 202.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Operating Licenses.
    Date of initial notice in Federal Register: February 4, 2000 (65 FR 
5705).
    The January 31, 2000, supplement was within the scope of the 
initial application as originally noticed. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
May 15, 2000.
    No significant hazards consideration comments received: No.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin.

    Date of application for amendment: November 24, 1999, as 
supplemented December 7, 1999, and February 8, 2000.
    Brief description of amendment: The amendment conforms the license 
to reflect the transfer of operating authority under Operating License 
No. DPR-43 to Nuclear Management Company, LLC, as approved by order of 
the Commission dated May 15, 2000.
    Date of issuance: August 7, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 149.
    Facility Operating License No. DPR-43: Amendment revised the 
Operating License.
    Date of initial notice in Federal Register: February 4, 2000 (65 FR 
5706).
    The February 8, 2000, supplement was within the scope of the 
initial application as originally noticed. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
May 15, 2000.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 16th day of August 2000.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-21340 Filed 8-22-00; 8:45 am]
BILLING CODE 7590-01-P