[Federal Register Volume 64, Number 193 (Wednesday, October 6, 1999)]
[Notices]
[Pages 54370-54393]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-25795]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is

[[Page 54371]]

publishing this regular biweekly notice. Public Law 97-415 revised 
section 189 of the Atomic Energy Act of 1954, as amended (the Act), to 
require the Commission to publish notice of any amendments issued, or 
proposed to be issued, under a new provision of section 189 of the Act. 
This provision grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license upon a 
determination by the Commission that such amendment involves no 
significant hazards consideration, notwithstanding the pendency before 
the Commission of a request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from September 11, 1999, through September 24, 
1999. The last biweekly notice was published on September 22, 1999 (64 
FR 51343 ).

Notice of Consideration of Issuance of Amendments to Facility 
Operating Licenses, Proposed no Significant Hazards Consideration 
Determination, and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration 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 November 5, 1999, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. 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

[[Page 54372]]

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: Rulemakings and 
Adjudications Staff, 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 at the local public document room for 
the particular facility involved.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of amendments request: September 1, 1999.
    Description of amendments request: The proposed amendment requests 
the following changes to the Technical Specifications:
    1. Change the definition of Azimuthal Power Tilt in Technical 
Specification 1.1;
    2. Correct the peak linear heat rate safety limit in Technical 
Specification 2.1.1.2;
    3. Correct the DC voltage range listed in Surveillance Requirements 
3.8.3.9 and 3.8.1.15;
    4. Correct the loss of voltage and degraded voltage settings in 
Surveillance Requirement 3.3.6.2;
    5. Correct the list of core operating limits in Technical 
Specification 5.6.5.a;
    6. Correct a note on Technical Specification Figure 2.1.1-1;
    7. Remove references to Unit 2, Cycle 12 in various Technical 
Specifications; and
    8. Correct a typographical error in Technical Specification 5.6.
    Specifically, the Proposed Technical Specifications are as follows:
    1. Technical Specification 1.1 is proposed to be changed to replace 
the definition of Azimuthal Power Tilt with a new definition.
    2. Technical Specification 2.1.1.2 is proposed to be changed by 
replacing the peak linear heat rate safety limit with less than or 
equal to 22kW/ft.
    3. Technical Specification SR 3.3.6.2 is proposed to be changed by 
replacing the degraded voltage function with transient degraded voltage 
and steady-state degraded voltage functions.
    4. Technical Specification SRs 3.8.1.9 and 3.8.1.15 are proposed to 
be changed by replacing the steady-state voltage range with the range 
of greater than or equal to 4060 volts and less than or equal to 4400 
volts.
    5. Technical Specification 5.6.5.a is proposed to be changed by 
adding Technical Specifications 3.1.4 and 3.3.1 to the list.
    6. Technical Specification Figure 2.1.1-1 is proposed to be changed 
by removing the reference to Figure B2.1-1.
    7. Various Technical Specifications and Figure 2.1.1-1a.
    8. Technical specification 5.6.5.b, Item 41.ii is proposed to be 
changed by correcting CEN-199(B)-P to CEN-119(b)-P.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
of consequences of an accident previously evaluated.

Change the Definition of Azimuthal Power Tilt

    In their Infobulletin 97-07, Revision 1, Asea Brown Boveri, 
Inc.,--Combustion Engineering, Inc. (ABB-CE) stated that they had 
found a discrepancy in the Technical Specification definition of 
azimuthal power tilt. This discrepancy was found to exist in all CE 
Nuclear Steam supply System analog plants that use CECOR for 
monitoring and surveillance, and that use ABB-CE safety analysis 
methodology. Calvert Cliffs is one of those plants.
    The value of Tq (Azimuthal tilt magnitude) as used in the 
azimuthal power tilt formula now in Technical Specification 1.1 is 
not conservative in all cases. With the proposed definition, Tq is 
the maximum fractional increase in power that can occur anywhere in 
the core because of tilt. Since Tq is the maximum value, it is 
consistently conservative. This is the appropriate measured value of 
tilt to be used in verifying that the tilt assumed in establishing 
safety limits has not been exceeded.
    Therefore, changing the definition of azimuthal power tilt as 
proposed will not involve a significant increase in the probability 
of consequences of an accident previously evaluated.

Correct the Peak Linear Heat Rate Safety Limit

    When Improved Standard Technical Specifications (ITS) were 
written, the peak linear heat rate safety limit of [less than or 
equal to] 21 kW/ft was inadvertently written in Technical 
specification 2.1.1.2. the correct number is [less than or equal to] 
22kW/ft. the peak linear heat rate safety limit was established at 
[less than or equal to] 22 kW/ft in License Amendment Nos. 88 (Unit 
1) and 61 (Unit 2). This number was valid for both units at the time 
of implementation of ITS.
    Therefore, changing the peak linear heat rate safety limit to a 
number previously approved by the Nuclear Regulatory Commission 
(NRC) will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

Correct the Diesel Generator Loss of Voltage and Degraded Voltage 
Settings

    When the ITS were written, a single set of numbers for the 
degraded voltage function was provided in Technical Specification 
Surveillance Requirement (SR) 3.3.6.2. The degraded voltage function 
should have been expressed as transient degraded voltage and steady-
state degraded voltage. This separation of two types of degraded 
voltage functions was approved in License Amendment Nos. 226 (Unit 
1) and 200 (Unit 2), which were issued before the ITS were approved.
    Therefore, changing the degraded voltage function to the 
transient degraded voltage and steady-state degraded voltage 
functions previously approved by the NRC will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.

Correct the Diesel Generator Voltage Range

    Technical Specification SRs 3.8.1.9 and 3.8.1.15 require each 
diesel to be started from a stand-by condition. Surveillance 
requirement 3.8.1.9 requires that the generator reach [greater than 
or equal to] 3740 volts within 10 seconds. After steady-state 
conditions are reached, both SRs require the generator to maintain a 
voltage range of greater than 3740 volts and [less than or equal to] 
4580 volts.

[[Page 54373]]

    The Baltimore Gas and Electric Company ITS conversion added 
voltage requirements to SRs 3.8.1.9 and 3.8.1.15 consistent with SR 
3.8.1.3. License Amendment Nos. 226 and 200 changed the voltage 
requirement for SR 3.8.1.3 to [greater than or equal to] 4060 volts 
and [less than or equal to] 4400 volts. The voltage was not 
corrected in SRs 3.8.1.9 and 3.8.1.15 when the Technical 
Specifications were changed to ITS.
    Therefore, changing the voltage in SRs 3.8.1.9 and 3.8.1.15 to 
voltage previously approved by the NRC will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.

Correct the List of Core Operating Limits

    Technical Specification 5.6.5.a lists Technical Specifications 
that are to be included in the core operating limits and documented 
in the Core Operating Limits Report (COLR). In the transition to 
ITS, Technical Specifications 3.1.4 (Control Element Assembly 
Alignment) and 3.3.1 (Reactor Protective System--Operating) were 
inadvertently omitted from the list. The complete list is currently 
in the COLR.
    Therefore, restoring Technical Specification 5.6.5.a to a list 
previously approved by the NRC will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Correct Figure 2.1.1-1

    A note of Technical Specification Figure 2.1.1-1 was changed in 
License Amendment Nos. 227 (Unit 1) and 201 (Unit 2) (ITS) to delete 
reference to Figure B2.1-1. Figure B2.1-1 was deleted from the 
Technical Specification Bases in the transition to ITS. In License 
Amendment Nos. 228 (Unit 1) and 202 (Unit 2), an old version of 
Figure 2.1.1-1 was used, and the reference to Figure B2.1-1 was thus 
inadvertently put back in the note. The proposed correction will 
replace the reference to Figure B2.1-1 with the wording approved in 
License Amendment Nos. 227 and 201.
    Therefore, returning the note in Figure 2.1.1-1 to the wording 
previously approved by the NRC will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Remove References to Unit 2, Cycle 12

    License Amendment Nos. 228 and 202 added notes to indicate areas 
in the Technical Specifications that had special application to 
Cycle 12 of Unit 2 only. Cycle 12 of Unit 2 ended in May 1999. Since 
these notes no longer have application, they are proposed to be 
removed. Additionally, Figure 2.1.1-la applies only to Unit 2, Cycle 
12, and it is proposed to be removed.
    Therefore, removal of information no longer applicable to either 
unit is an administrative change and will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Correct a Typographical Error

    Technical Specification 5.6.5.b, Item 41.ii is being corrected 
to change the number of the publication ``BASSS, Use of the Incore 
Detector System to Monitor the DNB-LCO on Calvert Cliffs Unit 1 and 
Unit 2'' from CEN-199(B) to CEN-119(B)-P. Correction of a 
typographical error does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from an accident previously evaluated.

Change the Definition of Azimuthal Power Tilt

    In their Infobulletin 97-07, Revision 1, ABB-CE stated that they 
had found a discrepancy in the Technical specification definition of 
azimuthal power tilt. This discrepancy was found to exist in all CE 
Nuclear Steam Supply System analog plants that use CECOR for 
monitoring and surveillance and that use ABB-CE safety analysis 
methodology. Calvert Cliffs is one of those plants.
    The value of Tq (azimuthal tilt magnitude) as used in the 
azimuthal power tilt formula now in Technical specification 1.1 is 
not always the most conservative in all cases. With the proposed 
definition, Tq is the maximum fractional increase in power that can 
occur anywhere in the core because of tilt. Since Tq is the maximum 
value, it is conservative. This is the appropriate measured value of 
tilt to be used in verifying that the tilt assumed by ABB-CE in 
establishing safety limits has not been exceeded.
    Therefore, changing the definition of azimuthal power tilt as 
proposed will not create the possibility of a new or different type 
of accident from any accident previously evaluated.

Correct the Peak Linear Heat Rate

    When the ITS were written, a value of peak linear heat rate 
[less than or equal to] 21 kW/ft was inadvertently written in 
Technical Specification 2.1.1.2. The correct number is [less than or 
equal to] 22 kW/ft. The required peak linear heat rate was 
established at [less than or equal to] 22 kW/ft in License Amendment 
Nos. 88 and 61. This number was valid for both units at the time of 
implementation of ITS.
    Therefore, changing the value of peak linear heat rate to a 
value previously approved by the NRC will not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.

Correct the Diesel Generator Loss of Voltage and Degraded Voltage 
Settings

    When the ITS were written, a single set numbers for the degraded 
voltage function was provided in Technical specification SR 3.3.6.2. 
The degraded voltage function should have been expressed as 
transient degraded voltage and steady-state degraded voltage. This 
separation of two types of degraded voltage functions was approved 
in License Amendment Nos. 226 and 200, which were issued before the 
ITS were approved.
    Therefore, changing the degraded voltage function to the 
transient degraded voltage and steady-state degraded voltage 
functions previously approved by the NRC will not create the 
possibility of a new or different type of accident from any accident 
previously evaluated.

Correct the Diesel Generator Voltage Range

    Technical Specification SRs 3.8.1.9 and 3.8.1.15 require that 
each diesel be started from a stand-by condition. Surveillance 
Requirement 3.8.1.9 requires that the generator reach [greater than 
or equal to] 3740 volts within 10 seconds. After steady-state 
conditions are reached, both SRs require the generator to maintain a 
voltage range of greater than 3740 volts and [less than or equal to] 
4580 volts.
    The Baltimore Gas and Electric Company ITS conversion added 
voltage requirements to SRs 3.8.1.9 and 3.8.1.15 consistent with SR 
3.8.1.3. License Amendment Nos. 226 and 200 changed the voltage 
requirement for SR 3.8.1.3 to [greater than or equal to] 4060 volts 
and [less than or equal to] 4400 volts. The voltage was not 
corrected in SRs 3.8.1.9 and 3.8.1.15 when the Technical 
Specifications were changed to ITS.
    Therefore, changing the voltage in SRs 3.8.1.9 and 3.8.1.15 to a 
voltage previously approved by the NRC will not create the 
possibility of a new or different type of accident from any accident 
previously evaluated.

Correct the List of Core Operating Limits

    Technical Specification 5.6.5.a lists Technical specifications 
that are to be included in the core operating limits and documented 
in the COLR. In the transition to ITS, Technical Specifications 
3.1.4 (Control Element Assembly Alignment) and 3.3.1 (Reaction 
Protective System--Operating) were inadvertently omitted from the 
list. The complete list is currently in the COLR.
    Therefore, restoring Technical Specification 5.6.5.a to a list 
previously approved by the NRC will not create the possibility of a 
new or different type of accident from any accident previously 
evaluated.

Correct Figure 2.1.1-1

    A note on Technical Specification Figure 2.1.1-1 was changed in 
License Amendment Nos. 227 and 201 (ITS) to delete reference to 
Figure B2.1-1. Figure B2.1-1 was deleted from the Technical 
Specification Bases in the transition of ITS. In License Amendment 
Nos. 228 and 202, an old version of Figure 2.1.1-1 was used, and the 
reference to Figure B2.1-1 was thus inadvertently put back in the 
note. The proposed correction will replace the reference to Figure 
B2.1-1 with the wording approved in License Amendment Nos. 227 and 
201.
    Therefore, removal of information no longer applicable to either 
unit is an administrative change and will not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.

Remove References to Unit 2, Cycle 12

    License Amendment Nos. 228 and 202 added notes to indicate areas 
in the Technical Specifications that had special application to 
Cycle 12 of Unit 2 only. Cycle 12 of Unit 2 ended in May 1999. Since 
these notes no longer have application, they are

[[Page 54374]]

proposed to be removed. Additionally, Figure 2.1.1-1a applies only 
to Unit 2, Cycle 12, and is proposed to be removed.
    Therefore, removal of information no longer applicable to either 
unit is an administrative change and will not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.

Correct a Typographical Error

    Technical Specification 5.6.5.b, Item 41.ii is being corrected 
to change the number of the publication ``BASSS, Use of the Incore 
Detector System to Monitor the DNB-LCO on Calvert Cliffs Unit 1 and 
Unit 2'' from CEN-199(B)-P to CEN-119(B)-P. Correction of a 
typographical error will not create the possibility of a new or 
different type of accident from any accident previously evaluated.
    3. Would not involve a significant reduction in the margin of 
safety.

Change the Definition of Azimuthal Power Tilt

    The margin of safety in this case is whether the azimuthal power 
tilt calculation shows the highest (most conservative) value for Tq 
(azimuthal tilt magnitude).
    The value of Tq as used in the azimuthal power tilt formula now 
in Technical Specification 1.1 is not always the most conservative 
in all cases. With the proposed definition, Tq is the maximum 
fractional increase in power that can occur anywhere in the core 
because of tilt. Since Tq is the maximum value, it is conservative. 
This is the appropriate measured value of tilt to be used in 
verifying that the tilt assumed in establishing safety limits has 
not been exceeded.
    Therefore, changing the definition of azimuthal power tilt as 
proposed will not involve a significant reduction in the margin of 
safety.

Correct the Peak Linear Heat Rate Safety Limit

    The margin of safety in this case was previously approved by the 
NRC in License Amendment Nos. 88 and 61.

Correct the Diesel Generator Loss of Voltage and Degraded Voltage 
Settings

    The margin of safety in this case was previously approved by the 
NRC in License Amendment Nos. 226 and 200.

Correct the Diesel Generator Voltage Range

    The margin of safety in this case was previously approved by the 
NRC in License Amendment Nos. 226 and 200.

Correct the List of Core Operating Limits

    Technical Specification 5.6.5.a lists Technical specifications 
that are to be included in the core operating limits and documented 
in the COLR. In the transition to ITS, Technical Specifications 
3.1.4 (Control Element Assembly Alignment) and 3.3.1 (Reactor 
Protective System--Operating) were inadvertently omitted from the 
list. The complete list is currently in the COLR.
    Therefore, restoring Technical Specification 5.6.5.a to a list 
previously approved by the NRC will not involve a significant 
reduction in the margin of safety.

Correct Figure 2.1.1-1

    A note on Technical Specification Figure 2.1.1-1 was changed in 
License Amendment Nos. 227 and 201 (ITS) to delete reference to 
Figure B2.1-1. Figure B2.1-1 was deleted from the Technical 
Specification Bases in the transition to ITS. In License Amendment 
Nos. 228 and 202, an old version of figure 2.1.1-1 was used, and the 
reference to Figure B2.1-1 was thus inadvertently put back in the 
note. The proposed correction will replace the reference to Figure 
B2.1-1 with the wording approved in License Amendment Nos. 227 and 
201.
    Therefore, returning the note in Figure 2.1.1-1 to the wording 
previously approved by the NRC will not involve a significant 
reduction in the margin of safety.

Remove References to Unit 2, Cycle 12

    License Amendment Nos. 228 and 202 added notes to indicate areas 
in the Technical Specifications that had special application to 
Cycle 12 of Unit 2 only. Cycle 12 of Unit 2 ended in May 1999. Since 
these notes no longer have application, they are proposed to be 
removed. Additionally, Figure 2.1.1-1a applies only to Unit 2, Cycle 
12, and it is proposed to be removed.
    Therefore, removal of information no longer applicable to either 
unit is an administrative change and will not involve a significant 
reduction in the margin of safety.

Correct a Typographical Error

    Technical specification 5.6.5.b, Item 41.ii is being corrected 
to change the number of the publication ``BASSS, Use of the Incore 
Detector system to Monitor the DNB-LCO on Calvert cliffs Unit 1 and 
Unit 2'' from CEN-199(B)-P to CEN-119(B)-P. Correction of a 
typographical error 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: S. Singh Bajwa.

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

    Date of amendment request: August 26, 1999.
    Description of amendment request: The proposed amendment would 
revise TS 3/4.9.4, ``Containment Building Penetrations,'' and its 
associated Bases to allow penetrations which provide direct access from 
the containment atmosphere to the outside atmosphere to remain open 
during refueling operations provided certain administrative controls 
are met.
    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 does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Containment is not an accident initiating system as described in 
the Final Safety Analysis Report. This change is applicable only in 
Mode 6 during Core Alterations or movement of irradiated fuel (which 
occurs when the unit is shutdown). The proposed change will not 
modify equipment used for fuel movement or core alterations within 
the HNP [Harris Nuclear Plant] Containment Building. Administrative 
controls will be used to isolate containment in the event of a fuel 
handling accident. The consequences of a Fuel Handling Accident 
inside containment will increase as a result of this change. 
However, the proposed administrative controls will require closure 
of containment prior to exceeding standard review plan dose limits 
due to a radiological release from a design basis fuel handling 
accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change provides for administrative controls and 
operating restrictions for air lock doors consistent with previous 
guidance authorized by the Commission for similar nuclear power 
plants. Containment is not an accident initiating system as 
described in the Final Safety Analysis Report. Fuel Handling 
Accidents have been previously analyzed for the Harris Nuclear 
Plant.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    Administrative controls will be used to isolate containment in 
the event of a fuel handling accident. The proposed administrative 
controls will require closure of containment prior to exceeding 
standard review plan dose limits due to a radiological release from 
a design basis fuel handling accident.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.


[[Page 54375]]


    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.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Section Chief: Sheri R. Peterson.

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

    Date of amendment request: July 29, 1999.
    Description of amendment request: The proposed change to the 
Arkansas Nuclear One, Unit 2, Technical Specifications would allow the 
performance of a special inspection of the steam generator tubes during 
an upcoming mid-cycle outage. This mid-cycle outage is planned for the 
purpose of performing inspections in selected areas of the steam 
generator tube bundle where previous inspections have revealed tube 
degradation. The proposed change would limit the initial inspection 
scope to these identified areas and includes a scope expansion criteria 
to address unexpected conditions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    An evaluation of the proposed change has been performed in 
accordance with 10 CFR 50.91(a)(1) regarding no significant hazards 
considerations using the standards in 10 CFR 50.92(c). A discussion 
of these standards as they relate to this amendment request follows:
    Criterion 1--Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    This change has no actual impact on any previously analyzed 
accident in the final safety analysis report (FSAR). A double-ended 
break of one steam generator tube is postulated as part of the ANO-2 
design basis accident evaluation. The change permits Entergy 
Operations to determine the appropriate scope and expansion criteria 
for a special steam generator tube inspection that is being 
performed at a frequency more conservative than that of the 
augmented inservice inspection program included in the TSs 
[Technical Specifications]. The special inspection will find and 
repair certain steam generator tubing flaws that would otherwise 
remain in service until the next scheduled refueling outage. The 
increased inspection frequency reduces the probability that a flaw 
in a steam generator tube could grow to a size that would affect the 
leakage or structural integrity of the tube. The augmented inservice 
inspection program contained in the TSs is not being modified.
    This change does not modify any parameter that will increase 
radioactivity in the primary system or increase the amount of 
radioactive steam released from the secondary safety valves or 
atmospheric dump valves in the event of a tube rupture.
    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 scope of this change does not establish a potential new 
accident precursor. The design basis accident analyses for ANO-2 
include the consequences of a double-ended break of one steam 
generator tube which bounds other postulated failure mechanisms. The 
proposed change would permit determination of alternate inspection 
criteria for a special inspection which is in addition to the 
periodic inservice inspections required by the TSs. The equipment 
used in the special inspection would not affect any plant components 
differently than those used for current TS required inspections.
    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.
    As previously stated, a double-ended rupture of one steam 
generator tube is accounted for in the ANO-2 design basis accident 
analysis. Considering that the 2P99 special inspection is in 
addition to the inservice inspection program defined in the ANO-2 
TSs and that leakage detection capability is not being modified, 
performance of a special inspection of any scope will increase the 
margin of safety over the current TS requirements.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Therefore, based upon the reasoning presented above and the 
previous discussion of the amendment request, Entergy Operations 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.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of amendment request: August 17, 1999.
    Description of amendment request: The proposed amendment would 
remove the voltage-based repair criteria, F* repair criteria, and 
sleeving methodologies from the Unit 1 Technical Specifications (T/S) 
and clarify the Bases sections accordingly.
    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.
    This change removes the interim steam generator tube plugging 
criteria from the T/S and reinstates the original T/S criteria 
consistent with Unit 2 (which does not have significantly degraded 
steam generators). The current T/S allow for continued operation 
with tubes that demonstrate indications per F* and voltage-based 
criteria. The basis used to justify the interim criteria is specific 
to the Unit 1 original steam generators (OSGs) and does not apply to 
the replacement steam generators (RSGs).
    The proposed change returns the plugging criteria for the steam 
generator tubes to the original licensing basis. The criteria are in 
accordance with NUREG-0452, (old) ``Standard Technical 
Specifications.'' The plugging criteria are based on a minimum wall 
thickness due to wastage as determined by ASME [American Society of 
Mechanical Engineers] Section XI. The proposed change is 
conservative in nature because it does not allow for continued 
operation with F* and voltage-based degraded tubes. Because of this, 
the probability of a steam generator tube rupture (SGTR) is not 
increased.
    The potential for a SGTR is also not increased as demonstrated 
in the qualification analysis and testing for the RSGs. The program 
for periodic in-service inspection monitors the integrity of the SG 
tubing to provide reasonable assurance that there is sufficient time 
to take proper and timely corrective action if any tube degradation 
is detected. The tube inspections themselves are not initiators of a 
SGTR. Therefore, this change is not expected to increase the 
probability of a SGTR during normal or accident conditions.
    Unit 1 will continue to apply the T/S maximum primary-to-
secondary leakage limit of 150 gallons per day (gpd) through any one 
SG to minimize the potential for excessive leakage. The EPRI 
[Electric Power Research Institute]-recommended 150 gpd limit

[[Page 54376]]

provides for leakage detection and plant shutdown in the event of an 
unexpected tube leak and minimizes the potential for excessive 
leakage or tube burst in the event of main steamline break (MSLB) or 
loss-of-coolant accident (LOCA) conditions. This lower limit is more 
restrictive than the limit (500 gpd per SG and total leakage of 1440 
gpd) utilized for determination of offsite dose and also provides 
further assurance that the probability of a SGTR is not increased.
    The design basis doses calculated for postulated accidents 
involving degradation of SG tubes, such as SGTR and MSLB accidents, 
as presented in UFSAR chapter 14 accident analysis, have been 
evaluated. The SGTR consequences continue to be bounded by the 
design basis analyses due to the allowable leakage rate specified by 
this change. The proposed T/S leakage rate is maintained at 150 gpd 
per SG. However, the maximum leakage of 500 gpd per SG and total 
leakage of 1440 gpd for all four generators was used to determine 
offsite dose in UFSAR chapter 14. The MSLB consequences are 
decreased by installation of the RSGs due to the reduction in 
primary-to-secondary leakage during the MSLB. Under the approved 
interim plugging criteria, a leak rate of 8.4 gpm was determined to 
be the upper limit for allowable primary-to-secondary leakage in the 
faulted steam generator. This leakage, combined with the 150 gpd 
leakage from the non-faulted SGs, was determined to limit the 
offsite dose to 10% of the 10 CFR 100 limits. Following replacement 
of the SGs, the leakage is limited during the MSLB to 150 gpd for 
both the faulted and unfaulted SGs. Therefore, the Unit 1 MSLB dose 
will be bounded by the current Unit 2 dose analysis, which is less 
than 10% of 10 CFR 100 limits.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Removing application of voltage-based repair criteria, F* repair 
criteria, and sleeving methodologies upon installation of the RSGs 
will not introduce significant or adverse changes to the plant 
design basis that could lead to a new or different kind of accident 
being created. This change does not change the overall objective of 
surveillance activities--maintaining the structural integrity of 
this portion of the reactor coolant system. The surveillance 
activities are performed during outages. The proposed change in the 
surveillance program returns the program to the initial licensing 
basis. No new failures are created.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Removing the application of voltage-based and F* repair criteria 
and sleeving methodologies does not involve a reduction in the 
margin of safety. The RSG tubing has been shown to retain adequate 
structural and leakage integrity during normal, transient, and 
postulated accident conditions consistent with GDC 14, 15, 30, 31, 
and 32 of 10 CFR [Part] 50 [A]ppendix A. The RSG tubing has been 
designed and evaluated consistent with the ASME Section III, 1989 
edition. The proposed plugging criteria are based on ASME Section XI 
and do not allow for operation with indications identified by F* and 
voltage-based criteria. The proposed program for periodic in-service 
inspection of the RSGs monitors the integrity of the SG tubing to 
provide reasonable assurance that there is sufficient time to take 
proper and timely corrective action if any tube degradation is 
present. The proposed program is consistent with NUREG-0452 and was 
the basis for the original Unit 1 T/S surveillance program.
    The proposed change maintains the T/S maximum primary-to-
secondary leakage at 150 gpd per generator to minimize the potential 
for excessive leakage. This limit provides for leakage detection and 
shutdown in the event of an unexpected tube leak and minimizes the 
potential for excessive leakage or tube burst in the event of a MSLB 
or LOCA. Because this limit is maintained, the margin of safety is 
maintained.
    Therefore, it is concluded that this 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.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, MI 49085.
    Attorney for licensee: Jeremy J. Euto, 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: September 10, 1999.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification (T/S) 3/4.4.7 so that the surveillance 
requirement does not need to be performed when the reactor is defueled 
with no forced circulation. The proposed revision to T/S 3/4.4.7 also 
includes changes to Tables 3.4-1 and 4.4-3. A change is proposed to 
Unit 1 T/S Table 4.4-3 to revise the reactor coolant system (RCS) 
chemistry sampling frequency from three times per 7 days with a maximum 
interval of 72 hours to a frequency of at least once per 72 hours. An 
editorial change to Unit 1 Tables 3.4-1 and 4.4-3 would relocate the 
asterisk for the footnote to a position adjacent to the parameter 
``dissolved oxygen,'' from its current position next to the allowable 
chemistry limit in Table 3.4-1 and the analysis frequency in Table 4.4-
3. An editorial change would also correct the footnote for Table 3.4-1 
for Unit 1 and Unit 2 by making the word ``limit'' plural, as it 
applies to both the steady-state and transient limits.
    Changes are also proposed to revise Surveillance Requirement 
4.11.2.2 by deleting the phrase ``by analysis of the Reactor Coolant 
System noble gases.''
    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 to the RCS chemistry sampling requirements 
do not affect the probability of a loss-of-coolant accident or steam 
generator tube rupture, which are evaluated in Sections 14.3 and 
14.2.4, respectively, of the Updated Final Safety Analysis Report 
(UFSAR). RCS contaminant limits are maintained to reduce the 
potential for RCS leakage or failure due to corrosion. Sampling the 
RCS for contaminants does not initiate an accident. Deleting the 
requirement to obtain samples when the reactor is defueled does not 
modify any plant equipment or affect plant operation and therefore 
does not introduce any new accident initiators or precursors. 
Suspension of RCS chemistry sampling when the reactor is defueled 
does not increase the potential for RCS leakage or failure because 
the corrosive effects of the contaminants is minimal during this 
low-temperature, low-pressure condition. To ensure elevated 
contaminant levels would be detected and corrected prior to 
subjecting the system to a high-temperature condition, chemistry 
sampling will be reinstated within 72 hours of re-establishing 
forced circulation and prior to entering Mode 6. Removing the 
restriction for analyzing primary coolant chemical contaminants at 
least three times every seven days does not change the maximum 
surveillance interval. This change allows the sample to be collected 
two or three times per week, consistent with the maximum 72-hour 
interval. The 72-hour sampling and analysis interval is consistent 
with the current requirement in the Unit 2 T/S, and industry 
guidance in NUREG-0452, ``Standard Technical Specifications.'' The 
72-hour interval continues to provide adequate assurance that 
concentrations in excess of the limits are detected in sufficient 
time to take corrective actions. Therefore, the probability of 
occurrence of a previously evaluated accident is not increased.
    This change does not alter the quantity of radioactive material 
in any system during normal plant operation, the amount of

[[Page 54377]]

shielding provided by plant systems, or the mitigative capabilities 
of any system following an event. Therefore, the consequences of a 
previously evaluated accident are not increased.
    The editorial changes to the RCS chemistry T/S provide 
consistency between the Unit 1 and Unit 2 T/S and the Standard 
Technical Specifications. These changes do not affect the design or 
operation of any system, structure, or component in the plant. The 
accident analysis assumptions and results are unchanged. No new 
failures or interactions are created.
    The amount of radioactive material in the gas storage tanks is 
controlled to ensure that, in the event of a rupture of one of these 
tanks, the resulting total body exposure to an individual at the 
nearest site boundary would not exceed 0.5 rem. The accidental waste 
gas release event is summarized in Section 14.2.3 of the UFSAR. 
Sampling to determine the radioactivity levels in the tanks does not 
initiate an accident or identify any accident precursors. The 
increased sampling flexibility does not change the method of 
operating the waste gas system, nor does it modify any interfaces 
with other plant systems. Therefore, this change does not increase 
the probability of occurrence of an accidental waste gas release 
event.
    Implementation of a different sampling method does not change 
the maximum quantity of radioactive material specified in the T/S 
Limiting Condition for Operation (LCO). The sampling method has no 
effect on normal plant gaseous radwaste activities, so the 
composition of the radioactive gaseous nuclides present in the tank 
at the time of the event is not affected. As the proposed revision 
allows a change to the method of sampling but does not affect the 
radioactivity limit for the gas storage tanks, the proposed change 
does not increase the consequences of an accidental waste gas 
release event.
    Therefore, the probability of occurrence or the consequences of 
accidents previously evaluated are not increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes to revise the RCS chemistry sampling 
frequency and to suspend RCS chemistry sampling when the reactor is 
defueled with no forced circulation does not change the method of 
operating any equipment or the operational limits of any equipment. 
The proposed changes do not introduce any new failure mechanisms to 
the RCS or any other plant systems. The proposed change does not 
involve any physical alterations to any plant equipment, and causes 
no change in the method by which any plant system performs its 
function. Editorial changes to footnotes for Tables 3.4-1 and 4.4-3 
provide consistency between the T/S for Unit 1 and Unit 2, but do 
not change the methods of operating any equipment or introduce any 
new failure mechanisms.
    The proposed change to eliminate the prescriptive waste gas tank 
sampling method does not introduce any new failure mechanisms to the 
waste disposal system, involve any physical changes to the waste 
disposal system or any other plant systems, or change the way any 
plant systems are operated. This change does not change any 
interfaces between the waste disposal system and any other plant 
systems. The proposed changes continue to ensure the system is 
operated within the existing limit established by the T/S LCO. Thus, 
no adverse safety considerations are introduced by this proposed 
change to the T/S.
    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 margin of safety pertinent to the RCS chemistry surveillance 
is related to the concentration of chemical contaminants that would 
expedite corrosion of the RCS piping and components and the period 
of time during which the system is allowed to operate outside the T/
S limits. The proposed changes to the RCS chemistry surveillance do 
not alter either of these criteria. These proposed changes do not 
affect any safety limits or T/S parameter limits. The proposed 
changes do not introduce new equipment, equipment modifications, or 
new or different modes of plant operation. These changes do not 
affect the operational characteristics of any equipment or systems. 
The editorial changes to footnotes for Tables 3.4-1 and 4.4-3 
provide consistency between the T/S for Unit 1 and 2, but do not 
affect the acceptance criteria or surveillance frequencies for this 
T/S.
    The margin of safety pertinent to the waste gas storage tanks is 
related to the quantity of radioactivity that would be released in 
the unlikely event of a tank rupture. The proposed change to the gas 
storage tank T/S eliminates the prescriptive sampling methodology, 
but does not affect the requirement to periodically quantify the 
radioactive gaseous material in the gas storage tanks. The proposed 
change does not affect the quantity of radioactivity allowed in the 
gas storage tanks, nor does it alter the methodology, assumptions, 
or results of any safety analyses. The proposed change to delete the 
prescriptive sampling method does not affect any safety limits or T/
S parameter limits.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, MI 49085.
    Attorney for licensee: Jeremy J. Euto, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: Claudia M. Craig.

National Aeronautics Space Administration (NASA), Docket No. 50-30, 
NASA Test Reactor, Erie County, Ohio

    Date of amendment request: March 25, 1999, as supplemented by 
letter dated August 10, 1999.
    Description of amendment request: The proposed amendment would 
change Lewis Research Center (LeRC) to Glenn Research Center (GRC).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed amendment will change the name of the Licensee for 
the Plum Brook Reactor Facility (PBRF) TR-3 license, a possession 
only license, from Lewis Research Center (LeRC) to the Glenn 
Research Center (GRC). The amendment request is necessary because 
NASA has changed the name of the Lewis Research Center to the Glenn 
Research Center at Lewis Field under legislative action and signed 
into law (sec. 434, P.L. 105-276, 112 Stat. 2461) on October 21, 
1998. The effective date of this name change was March 1, 1999. 
NASA, GRC will retain the PBRF license and the responsibility to 
continue maintaining the PBRF Reactor Facility in a safe protected 
storage mode under the current TR-3 possess-but-not-operate license. 
In addition, the current plans to provide a PBRF decommissioning 
plan to the NRC by the end of CY 1999 and the eventual 
decommissioning by the end of CY 2007 have not changed.
    There will be no change in the funding status of the GRC in 
either maintaining the PBRF facility in the safe protected storage 
mode or the eventual decommissioning. NASA, as a government agency, 
remains responsible for the continuing funding of both activities.
    In addition, there will be no change in the personnel who are 
responsible for maintaining the present TR-3 license or in 
developing the PBRF Decommissioning Plan.
    The proposed amendment does not require any physical change to 
the PBRF Facility, changes to the Technical Specifications or 
procedures under the PBRF TR-3 License other than the name change 
from LeRC to GRC. The proposed change does not increase the 
probability of any accident or increased risk to the public safety.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident condition 
previously evaluated.
    (2) Would not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed amendment does not modify the PBRF facility 
configuration or licensed activities. Therefore, no additional 
accident conditions are introduced.

[[Page 54378]]

    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequence of an accident.
    (3) Would not involve a significant reduction in a margin of 
safety.
    This amendment is required because of the name change from LeRC 
to GRC. NASA will continue to be financially responsible to maintain 
the PBRF Facility under the existing TR-3 License.
    Furthermore, the GRC personnel for the eventual PBRF 
decommissioning and contract support personnel reporting to GRC will 
continue to be technically qualified to maintain the PBRF under the 
safe protected storage mode. There has been no effective change in 
the personnel who will be responsible to implement the eventual 
decommissioning effort that will be required under the future PBRF 
Decommissioning Plan.
    Plum Brook's existing qualified contractors remained in place 
following the name change. The requested amendment does not involve 
any changes in the performance of current licensed activities and 
these activities will continue in their current form without changes 
or interruptions of any kind.
    The proposed amendment does not alter any margin of safety 
because it does not involve any changes in the PBRF Facility or 
licensed activities under the TR-3 License. All activities will 
continue in the current form without changes or interruptions of any 
kind as a result of the name.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: N/A.
    Attorney for licensee: Elias T. Naffah, MS 500-118, NASA, Glenn 
Research Center, 21000 Brookpark Road, Cleveland Ohio 44135.
    NRC Branch Chief: Ledyard B. Marsh.

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: July 16, 1999.
    Description of amendment request: Proposed relocation of Technical 
Specifications 3/4.9.3.2, ``Refueling Operations, Spent Fuel 
Temperature,'' 3/4.9.3.3, ``Refueling Operations, Decay Time,'' 3/
4.9.5, ``Refueling Operation, Communications,'' 3/4.9.6, ``Refueling 
Operation, Crane Operability--Containment Building,'' and 3/4.9.7, 
``Refueling Operations, Crane Travel--Spent Fuel Storage Building,'' to 
the Millstone, Unit No. 2 Technical Requirements Manual. The associated 
Bases pages and index pages will be modified to address the proposed 
change.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Technical Specification 3/4.9.3.2, ``Refueling Operations, Spent 
Fuel Pool Temperature,'' is proposed to be relocated to the TRM 
where future changes will be controlled in accordance with 10 CFR 
50.59. This specification limits spent fuel pool temperature to be 
less than or equal 140  deg.F to ensure the resin in the spent fuel 
cooling demineralizers will not degrade and the temperature and 
humidity are compatible with personnel comfort and safety 
requirements. Additionally, the requirement ensures that the design 
temperature of the fuel pool cooling system, liner/building 
structures, and racks is not exceeded. Relocation of this Technical 
Specification to the TRM does not imply any reduction in its 
importance in limiting the spent fuel pool bulk temperature to be 
less than or equal to 140  deg.F. Spent fuel pool bulk temperature 
is a design bases process variable which is used to establish the 
required heat removal capabilities of the spent fuel heat removal 
system. In the unlikely event of total loss of cooling water flow to 
the spent fuel pool, the pool water temperature may reach 212  deg.F 
within approximately 9 hours and will result in a boiling condition. 
This event does not represent a challenge to the fuel cladding, as a 
fission product barrier, unless the fuel becomes uncovered. The 
requirement on storage pool water level is covered by Technical 
Specification 3/4.9.12, ``Storage Pool Water Level,'' which requires 
a minimum of 23 feet of water over the top of irradiated fuel 
assemblies. Therefore, spent fuel pool bulk temperature is not by 
itself a process variable that is an initial condition of a design 
basis accident. This Technical Specification does not cover a 
process variable, design feature, or operating restriction that is 
an initial condition of a design basis accident or transient 
analysis that either assumes the failure of or presents a challenge 
to the integrity of a fission product barrier. It does not cover a 
structure, system, or component that is part of the primary success 
path which functions or actuates to mitigate a design basis accident 
or transient that either assumes the failure of or presents a 
challenge to the integrity of a fission product barrier. The 
proposed change will not alter the way pool temperature is measured, 
nor will it alter any of the assumptions used in the spent fuel pool 
fuel handling accident analysis. Relocation of this Technical 
Specification to the TRM does not degrade the performance of any 
safety systems or prevent actions assumed in the accident analysis, 
nor does it alter any of the assumptions made in the analysis that 
could increase the consequences of accidents. Therefore, this change 
will not significantly increase the probability or consequences of 
an accident previously evaluated.
    Technical Specification 3/4.9.3.3, ``Refueling Operations, Decay 
Time,'' is proposed to be relocated to the TRM where future changes 
will be controlled in accordance with 10 CFR 50.59. This 
specification requires the reactor to remain in Mode 5 or 6 until 
the most recent core offload has decayed a sufficient time to ensure 
alternate cooling is available during this time to cool the spent 
fuel pool should a failure occur in the Spent Fuel Pool Cooling 
System. Alternate cooling would be provided by the Shutdown Cooling 
System. Relocation of this Technical Specification to the TRM does 
not imply any reduction in its importance in insuring that the most 
recent core offload has decayed a sufficient time. If the 
requirement to remain in Mode 5 or 6 until the most recent core 
offload has decayed for 504 hours is not satisfied, the spent fuel 
pool cooling system may not have the capability to remove decay heat 
and stay below the Technical Specification limit of 140  deg.F. In 
the unlikely event of total loss of cooling water flow to the spent 
fuel pool, the pool water temperature may reach 212  deg.F in less 
than 9 hours and will result in a boiling condition. This event does 
not represent a challenge to the fuel cladding, as a fission product 
barrier, unless the fuel becomes uncovered. The requirements on 
storage pool water level is covered by Technical Specification 3/
4.9.12, ``Storage Pool Water Level,'' which requires a minimum of 23 
feet of water over the top of irradiated fuel assemblies. Therefore, 
this requirement to remain in Mode 5 or 6 until the most recent core 
offload has decayed for 504 hours is not by itself a process 
variable that is an initial condition of a design basis accident. 
This Technical Specification does not cover a process variable, 
design feature, or operating restriction that is an initial 
condition of a design basis accident or transient analysis that 
either assumes the failure of or presents a challenge to the 
integrity of a fission product barrier. It does not cover a 
structure, system, or component that is part of the primary success 
path which functions or actuates to mitigate a design basis accident 
or transient that either assumes the failure of or presents a 
challenge to the integrity of a fission product barrier. The 
proposed change will not alter the requirement that the most recent 
core offload has decayed a sufficient time, nor will it alter any of 
the assumptions used in the spent fuel pool fuel handling accident 
analysis. Relocation of this Technical Specification to the TRM does 
not degrade the performance of any safety systems or prevent actions 
assumed in the accident analysis, nor does it alter any of the 
assumptions made in the analysis that could increase the 
consequences of accidents. Therefore, this change will not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    Technical Specification 3/4.9.5, ``Refueling Operations, 
Communications,'' is proposed to be relocated to the TRM where 
future

[[Page 54379]]

changes will be controlled in accordance with 10 CFR 50.59. This 
specification requires communication between the control room and 
the refueling station, to ensure any abnormal change in the facility 
status, as indicated on the control room instrumentation, can be 
communicated to the refueling station personnel. Relocation of this 
Technical Specification to the TRM does not imply any reduction in 
its importance in insuring communication between the control room 
and the refueling station. This Technical Specification does not 
cover a process variable, design feature, or operating restriction 
that is an initial condition of a design basis accident or transient 
analysis that either assumes the failure of or presents a challenge 
to the integrity of a fission product barrier. It does not cover a 
structure, system, or component that is part of the primary success 
path which functions or actuates to mitigate a design basis accident 
or transient that either assumes the failure of or presents a 
challenge to the integrity of a fission product barrier. The 
proposed change will not alter the requirement on communication 
between the control room and the refueling station, nor will it 
alter any of the assumptions used in the spent fuel pool fuel 
handling accident analysis. Relocation of this Technical 
Specification to the TRM does not degrade the performance of any 
safety systems or prevent actions assumed in the accident analysis, 
nor does it alter any of the assumptions made in the analysis that 
could increase the consequences of accidents. Therefore, this change 
will not significantly increase the probability or consequences of 
an accident previously evaluated.
    Technical Specification 3/4.9.6, ``Refueling Operations, Crane 
Operability--Containment Building,'' is proposed to be relocated to 
the TRM where future changes will be controlled in accordance with 
10 CFR 50.59. This specification ensures the lifting device on the 
refueling machine has adequate capacity to lift the weight of a fuel 
assembly and a control element assembly, and that an automatic load 
limiting device is available to prevent damage to the fuel assembly 
during fuel movement. Relocation of this Technical Specification to 
the TRM does not imply any reduction in its importance in insuring 
that the lifting device on the refueling machine has adequate 
capacity. The automatic load limiting device and/or physical stops 
are not monitored and controlled during operation, nor are they 
assumed to function to mitigate the consequences of a design basis 
accident. The automatic load limiting device is checked on a 
periodic basis to ensure operability. This Technical Specification, 
which ensures the lifting device on the refueling machine has 
adequate capacity, does not cover a process variable, design 
feature, or operating restriction that is an initial condition of a 
design basis accident or transient analysis that either assumes the 
failure of or presents a challenge to the integrity of a fission 
product barrier. The proposed change will not alter the requirement 
that the lifting device on the refueling machine has adequate 
capacity, nor will it alter any of the assumptions used in the 
accident analysis. Relocation of this Technical Specification to the 
TRM does not degrade the performance of any safety systems or 
prevent actions assumed in the accident analysis, nor does it alter 
any of the assumptions made in the analysis that could increase the 
consequences of accidents. Therefore, this change will not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    Technical Specification 3/4.9.7, ``Refueling Operations, Crane 
Travel--Spent Fuel Storage Pool Building,'' is proposed to be 
relocated to the TRM where future changes will be controlled in 
accordance with 10 CFR 50.59. This specification ensures loads in 
excess of one fuel assembly containing a control element assembly, 
plus the weight of the fuel handling tool, will not be moved over 
other fuel assemblies in the spent fuel storage racks. Therefore, in 
the event of a drop of this load, the activity released is limited 
to that contained in one fuel assembly. Relocation of this Technical 
Specification to the TRM does not imply any reduction in its 
importance in insuring that loads in excess of 1800 pounds (except 
of a consolidated fuel storage box) are prohibited from travel over 
irradiated fuel. While this Technical Specification does address an 
operating restriction assumed in the accident analysis, there is no 
process variable that can be monitored during power operation of the 
plant. Crane interlocks and/or physical stops are used to assure 
that this requirement is met, but indication of the operation of the 
interlocks and/or physical stops is not available in the control 
room. These features inhibit movement of the crane so that 
monitoring is not necessary. This Technical Specification does not 
cover a structure, system, or component that is part of the primary 
success path which functions or actuates to mitigate a design basis 
accident or transient that either assumes the failure of or presents 
a challenge to the integrity of a fission product barrier. The 
proposed change will not alter the requirement that the crane 
interlocks and/or physical stops are OPERABLE, nor will it alter any 
of the assumptions used in the spent fuel pool fuel handling 
accident analysis. Relocation of this Technical Specification to the 
TRM does not degrade the performance of any safety systems or 
prevent actions assumed in the accident analysis, nor does it alter 
any of the assumptions made in the analysis that could increase the 
consequences of accidents. Therefore, this change will not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    Revision of Index Pages IX and XIII and the proposed change to 
Bases sections, by relocating them to the TRM, are administrative 
changes. Therefore, this change will not significantly increase the 
probability or consequences of an accident previously evaluated. The 
proposed changes do not alter how any structure, system, or 
component functions. There will be no effect on equipment important 
to safety. The proposed changes have no effect on any of the design 
basis accidents previously evaluated. Therefore, this License 
Amendment Request does not impact the probability of an accident 
previously evaluated, nor does it involve a significant increase in 
the consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. They do not alter the way any 
structure, system, or component functions and do not alter the 
manner in which the plant is operated. The proposed changes do not 
introduce any new failure modes. 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 relocation of Technical Specification 3/4.9.3.2, 
``Refueling Operations, Spent Fuel Pool Temperature,'' to the TRM 
does not imply any reduction in its importance in limiting the spent 
fuel pool bulk temperature to less than or equal to 140  deg.F. The 
proposed change will not alter the way pool temperature is measured. 
It will not alter any of the assumptions used in the spent fuel pool 
fuel handling accident analysis, nor will it cause any safety system 
parameters to exceed their acceptance limit. The proposed relocation 
of Technical Specification 3/4.9.3.3, ``Refueling Operations, Decay 
Time,'' to the TRM does not imply any reduction in its importance in 
insuring that the most recent core offload has decayed a sufficient 
time. The proposed change will not alter the requirement that the 
most recent core offload has decayed a sufficient time, it will not 
alter any of the assumptions used in the spent fuel pool fuel 
handling accident analysis, nor will it cause any safety system 
parameters to exceed their acceptance limit. The relocation of 
Technical Specification 3/4.9.5, ``Refueling Operations, 
Communications,'' to the TRM does not imply any reduction in its 
importance in insuring communication between the control room and 
the refueling station. The proposed change will not alter the 
requirement on communication between the control room and the 
refueling station, it will not alter any of the assumptions used in 
the spent fuel pool fuel handling accident analysis, nor will it 
cause any safety system parameters to exceed their acceptance limit. 
The relocation of Technical Specification 3/4.9.6, ``Refueling 
Operations, Crane Operability--Containment Building,'' to the TRM 
does not imply any reduction in its importance in insuring that the 
lifting device on the refueling machine has adequate capacity. The 
proposed change will not alter the requirement that the lifting 
device on the refueling machine has adequate capacity, it will not 
alter any of the assumptions used in the accident analysis, nor will 
it cause any safety system parameters to exceed their acceptance 
limit. The relocation of Technical Specification 3/4.9.7, 
``Refueling Operations, Crane Travel--Spent Fuel Storage Pool 
Building,'' to the TRM does not imply any reduction in its 
importance in insuring that loads in excess of 1800 pounds (except 
of a consolidated fuel storage box) are prohibited from travel over 
irradiated fuel. The proposed change will not

[[Page 54380]]

alter the requirement that the crane interlocks and/or physical 
stops are OPERABLE, it will not alter any of the assumptions used in 
the spent fuel pool fuel handling accident analysis, nor will it 
cause any safety system parameters to exceed their acceptance limit. 
Revision of Index Pages IX and XIII and the proposed change to Bases 
sections by eliminating the sections corresponding to the relocated 
Technical Specifications are administrative changes. These changes 
will not alter any of the assumptions used in the spent fuel pool 
fuel handling accident analysis, nor will it cause any safety system 
parameters to exceed their acceptance limit. The proposed changes do 
not affect any of the assumptions used in the accident analysis, nor 
do they affect any operability requirements for equipment important 
to plant safety. Therefore, the proposed changes will not result in 
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.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    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.

PECO Energy Company, Docket No. 50-352, Limerick Generating Station, 
Unit 1, Montgomery County, Pennsylvania

    Date of amendment request: June 7, 1999.
    Description of amendment request: The proposed change to the 
Technical Specifications (TSs), if approved, will reflect the permanent 
deactivated configuration of the ``wet'' instrument reference leg 
isolation valve HV-61-102 which originally connected the Drywell Floor 
and Equipment Drain Tanks to level instruments outside the containment. 
The TS changes affecting TS Table 3.6.3-1, ``Primary Containment 
Isolation Valves,'' and its associated notations will reflect the 
current plant configuration. More specifically, TS Section 3/4.6.3, 
``Primary Containment Isolation Valves,'' Table 3.6.3-1, Penetration 
Number 230B will be revised to designate the function of valve HV-61-
102 as ``Deactivated,'' the maximum isolation time for valve HV-61-102 
will be eliminated, and notations 1, 23, and 29 will be replaced with a 
new notation indicating the permanent configuration of the subject 
valve.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:

    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The closed valve, HV-61-102, has no effect on the function of 
the Drywell Sump/Equipment Drain Tanks, other safety-related 
systems, or other containment penetrations. The current status of 
the valve is locked closed, de-energized, and the motor operator 
cannot be accidentally actuated. In addition, the line is capped 
downstream of the isolation valve. As described above, the valve is 
considered to be in a passive configuration, where a malfunction is 
not expected and cannot cause an increase in the probability of a 
malfunction to itself or other safety-related equipment. The 
potential for increased releases outside the containment due to 
breaching of the valve assembly is no greater than that of the 
isolation design previously evaluated.
    Therefore, the proposed change to the TSs does not involve a 
significant increase in the probability of occurrence or 
consequences of an accident previously evaluated in the Safety 
Analysis Report.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The abandoned isolation valve conforms to approved isolation 
configurations, and its structural integrity has not been degraded 
by the modified configuration. The original function of valve HV-61-
102 was only to provide isolation of the instrument line. Following 
the modification, the valve is independent of the function of the 
Drywell Sump/ Equipment Drain Tanks, other safety-related systems, 
and other penetrations. Since the valve is passive and has no 
requirements to be operated, it cannot create a different type of 
malfunction on itself or other safety-related systems. In addition, 
the valve is specifically designed to isolate and is essentially 
passive during accident conditions, it has no activity that could be 
the initiator of an accident of a different type.
    Therefore, the proposed changes to the TSs do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety. Isolation valve HV-61-102 in its 
proposed permanent configuration meets the margin of safety 
described in TS Bases 3/4.6.3 since it is kept closed under all 
operational conditions and will not be under the constraint of TS 
closing times in order to maintain releases within specifications. 
The proposed changes have no impact on any safety analysis 
assumptions.
    Therefore, the proposed TS changes do not involve a significant 
reduction in the 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.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101.
    NRC Section Chief: James W. Clifford.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of amendment request: July 23, 1999, as supplemented on 
September 13, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Surveillance Requirement 4.8.1.1.2 to 
allow the 24-hour emergency diesel generator endurance run to be 
performed during power operation (i.e., Modes 1 and 2) instead of 
restricting the test to when the reactor was shutdown.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change to Technical Specification Surveillance 
Requirement (SR) 4.8.1.1.2.d.7 (24-hour emergency diesel generator 
(EDG) endurance run test) to eliminate the restriction to perform 
the test during shutdown conditions does not involve a significant 
increase in the probability of any previously evaluated accident. 
Although paralleling or connecting the EDG to off-site power for the 
test could induce an electrical distribution system perturbation, 
the same possibility exists when the EDG is tested during the 
monthly 1-hour loaded surveillance test (SR 4.8.1.1.2 a 2). This 
risk during testing the EDG monthly at power was reviewed and found 
acceptable by the NRC. Further, none of the automatic actuations and 
interlocks in the tested portion of the electrical system or the EDG 
control system are disabled during the 24-hour endurance run. Thus, 
the onsite safety-related electrical system remains protected from 
potential faults and perturbations.
    The ability and capability [o]f the EDG to perform their safety 
function (mitigate the consequences of a previously evaluated

[[Page 54381]]

accident) is also unaffected. This capability was demonstrated not 
only by the tests conducted in the EDG manufacturer's plant, but 
continue to be demonstrated by surveillance testing performed at the 
station.
    This testing verifies specific design criteria, which assure 
continued EDG operability even during testing. Examples of presently 
performed Technical Specification testing that demonstrate the 
ability and capability of the EDG to perform its safety functions 
are:
     SR 4.8.1.1.2. d. 2 requires, in part, that on a load 
rejection of greater than 820 KW, the voltage and frequency be 
restored to acceptable values within 4 seconds.
     This surveillance demonstrates the ability of the EDGs 
to withstand a loss of load, as it would occur in a normal 
safeguards equipment controller (SEC) actuation, without 
compromising its ability to be ready to accept a new loading 
sequence and carry its design safety function.
     SR 4.8.1.1.2. d. 9 requires, in part, that with the EDG 
operating in a test mode (connected to its bus), a simulated safety 
injection signal overrides the test mode by (1) returning the diesel 
generator to standby operation and (2) automatically energizing the 
emergency loads with offsite power.
    This surveillance demonstrates the ability of the EDGs to be 
disconnected from the grid, if in a test mode, on an accident 
signal, and be ready to accept a new loading sequence and carry its 
design safety function.
     SR 4.8.1.1.2. a. 2 requires, in part, that every 31 
days each EDG be demonstrated OPERABLE by synchronizing it to the 
grid for greater than or equal to 60 minutes.
    Note that this proposed amendment request eliminates a 
discrepancy between the current requirement to perform the 24 hour 
run during shutdown and SR 4.8.1.1.2.a.2, which would allow a 24 
hour run at power.
    Additionally, PSE&G performed an assessment of the potentially 
added risk of an additional 24 hours of on-line EDG testing. The 
unavailability of all three EDGs was increased in the Probabilistic 
Safety Analyses (PSA) for both Salem Units 1 and 2 to correspond to 
an additional 24 hours per cycle out-of-service time each 18-month 
operating cycle. The unavailability was changed from 1.86E-02/year 
to 2.0E-2/year. The increase in the baseline internal events core 
damage frequency (CDF) was determined to be 1.6E-07 events/year for 
both Salem Units 1 and 2. Based on the definition provided in 
Regulatory Guide 1.174, Paragraph 2.2.4, this increase is considered 
a very small increase in risk (less than 1.0E-06 events/year).
    Therefore, the proposed amendment, including proposed 
administrative controls, does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment to Technical Specification Surveillance 
Requirement 4.8.1.1.2.d.7 (24-hour endurance run test) to eliminate 
the restriction to perform the test during shutdown conditions does 
not physically modify the facility, introduce a new failure mode, or 
propose a different operational mode of the AC electrical power 
sources, or Emergency Diesel Generators.
    Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The AC Electrical distribution system has been designed to 
provide sufficient redundancy and reliability to ensure the 
availability of the EDGs to provide the required safety function 
under design basis events to protect the power plant, the public and 
plant personnel. Specifically, the ability of the EDGs to separate 
from the off-site power source has been designed and tested per 
Technical Specifications requirements.
    Performance of the 24-hour endurance run during power operations 
will not affect the availability of any of the required power 
sources, nor the capability of the EDGs to perform their intended 
safety function. Furthermore, performing the test when the 
undervoltage protection of the 4160-V vital buses required by the 
Salem Station Technical Specification 3.3.2.1 is operable, provides 
for an added level of protection to the EDG that is not available 
while shutdown.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of application for amendments: August 30, 1999 (TS 99-08).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah (SQN) Technical Specification (TS) requirements to 
provide alternatives to the requirement of actually measuring response 
times.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This change to the TS does not result in a condition where the 
design, material, and construction standards that were applicable 
prior to the change are altered. The same RTS [Reactor Trip System] 
and engineered safety feature actuation system (ESFAS) 
instrumentation is being used, the time response allocations/
modeling assumptions in the [Final Safety Analysis Report] Chapter 
15 analyses are still the same, only the method of verifying time 
response is changed. The proposed change will not modify any system 
interface and could not increase the likelihood of an accident since 
these events are independent of this change. The proposed activity 
will not change, degrade or prevent actions, or alter any 
assumptions previously made in evaluating the radiological 
consequences of an accident described in the Final Safety Analysis 
Report. Therefore, the proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    This change does not alter the performance of pressure [or] 
differential pressure transmitters, solid state protection system 
racks, nuclear instrumentation, or input and output master/slave 
relays used in the plant protection systems. Applicable sensors, 
solid state protection system (SSPS) racks, nuclear instrumentation, 
and relays will still have response time verified by test prior to 
placing the equipment in operational service and after any 
maintenance that could affect the response time of that equipment. 
Changing the method of periodically verifying instrument response 
time for certain instruments from RTT [Response Time Test] to 
calibration and channel checks or functional test will not create 
any new accident initiators or scenarios. Therefore, the proposed 
amendment does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    This change does not affect the total system response time 
assumed in the safety analysis. The periodic system response time 
verification method for selected pressure and pressure differential 
sensors and SSPS racks, nuclear instrumentation, or logic systems is 
modified to allow use of actual test data or engineering data 
(various Westinghouse WCAPs [topical reports]). The method of 
verification still provides assurance that the total system response 
time is within that assumed in the safety analysis, since 
calibration checks and functional tests will detect any degradation 
which might significantly affect equipment response time. Therefore, 
the proposed license amendment request does not result in a 
significant reduction in margin of safety.


[[Page 54382]]


    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Sheri R. Peterson.

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

    Date of application for amendments: August 30, 1999 (TS 99-10).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah (SQN) Technical Specifications (TS) to provide 
clarification to the requirements for containment isolation valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed revisions enhance the technical specification (TS) 
requirements to provide greater consistency with the standard TS in 
NUREG-1431. This revision proposes changes to the requirements for 
containment isolation valves in Specifications 3.6.3. A proposed 
revision relocates a surveillance requirement (SR) from SQN TS 
3.6.1.1, ``Containment Integrity'' to SQN TS 3.6.3, ``Containment 
Isolation Valves.'' A proposed revision to TS 3.6.3, Action (a), a 
new Action (b), and a proposed revision to SR 4.6.3.2 provide 
improvements to the existing TS requirements. The proposed revisions 
are not the result of changes to plant equipment, system design, 
testing methods, or operating practices. The modified requirements 
will allow some relaxation of current action requirements, and SRs. 
These changes provide more appropriate requirements in consideration 
of the safety significance and the design capabilities of the plant 
as determined by the improved standard TS industry effort. SQN TS 
3.6.3, ``Containment Isolation Valves,'' continues to provide 
controls to ensure these valves isolate within the time limits 
assumed in the safety analyses. Operability of these valves 
continues to assure that the containment isolation function assumed 
in the safety analyses is maintained. Since these proposed revisions 
will continue to support the required safety functions without 
modification of the plant features, the probability of an accident 
is not increased.
    The provisions proposed in this change request will continue to 
maintain an acceptable level of protection for the health and safety 
of the public and will not significantly impact the potential for 
the offsite release of radioactive products. The overall effect of 
the proposed change will result in specifications that have 
equivalent or improved requirements compared to existing 
specifications for containment isolation valve operability and will 
not significantly increase the consequences of an accident.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed revisions are not the result of changes to plant 
equipment, system design, testing methods, or operating practices. 
The modified requirements will allow some relaxation of current 
action requirements, and a SR consistent with NUREG-1431. These 
changes provide more appropriate requirements in consideration of 
the safety significance and the design capabilities of SQN's 
containment isolation system. The specifications for containment 
isolation valves serve to provide controls for maintaining the 
containment pressure boundary. TVA's proposed changes does not 
contribute to the generation of postulated accidents. Since the 
function of the containment isolation valves and their associated 
systems remains unchanged, and the effects do not contribute to 
accident generation, the proposed changes will not create the 
possibility of a new or different kind of accident.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed changes will not result in changes to system design 
or setpoints that are intended to ensure timely identification of 
plant conditions that could be precursors to accidents or potential 
degradation of accident mitigation systems. Operability requirements 
for SQN's containment isolation valves remain unchanged. TVA's 
proposed revisions provide some relaxation and flexibility to 
existing actions and a SR; however, the addition of a new action 
requirement for a 31-day periodic verification of valve position 
provides conservative administrative controls to ensure containment 
isolation function is maintained. The action times are acceptable 
considering the redundant features of containment penetration flow 
paths and the allowed time intervals that have been developed by the 
industry and NRC.
    TVA's revisions will continue to provide the necessary actions 
to minimize the impact of inoperable containment isolation valves 
and will provide testing activities that will ensure containment 
isolation system operability. The setpoints and design features that 
support the margin of safety are unchanged and actions for 
inoperable systems continue to provide appropriate time limits and 
compensatory measures. Accordingly, the proposed changes will not 
significantly reduce the margin of safety.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Sheri R. Peterson.

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

    Date of application for amendments: August 30, 1999 (TS 99-11).
    Brief description of amendments: The proposed amendments would add 
Sequoyah (SQN) Technical Specification (TS) 3.0.7 to address the use of 
interim provisions upon discovery of unintended TS action.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    TVA proposes the addition of a new definition and limiting 
condition for operation (LCO) that will allow the interim correction 
of erroneous TS requirements until NRC's review of an amendment 
request is completed. This allowance will only apply to those errors 
that are clearly in conflict with the intended purpose of the TS 
requirement. The proposed revision will not alter any plant 
equipment or operating practices or deviate from the intended 
application of the TS requirements. Therefore, the probability of an 
accident is not increased by this revision. Likewise, the 
consequences of an accident is not increased because the proposed 
allowance will maintain the underlying intent of the TS 
requirements, the plant licensing basis, and plant nuclear safety.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed revision to the SQN TSs will not alter plant 
equipment or operating practices. The intent of the TS requirements 
will be maintained to ensure the assumed initial conditions for 
accidents and the availability of mitigation systems in the event of 
an postulated accident. The proposed addition will not promote 
activities that have

[[Page 54383]]

the potential to generate accidents. Therefore, the proposed 
revision will not create the possibility of an accident of a new or 
different kind.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    TVA's proposed revision to add an allowance to correct erroneous 
TS requirements will not alter plant systems or those setpoints and 
limits that are use[d] to maintain safety functions. Any corrections 
implemented in accordance with the proposed allowance will be 
consistent with the underlying intent of the TSs. TVA will pursue 
timely correction of such errors through the license amendment 
process while temporarily utilizing the corrected requirement. This 
will ensure that inadequate TS requirements are resolved with NRC in 
an acceptable time interval. Implementation of the proposed revision 
will enhance the ability to maintain the licensing basis and safety 
features of the plant without the need for unnecessary unit 
shutdowns or regulatory activities. Therefore, the proposed revision 
maintains the plant safety features without the reduction of any 
margin of safety.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Sheri R. Peterson

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

    Date of application request: September 8, 1999.
    Description of amendment request: The amendment will authorize 
revisions to the Final Safety Analysis Report (FSAR) to reflect 
increases in the radiological dose consequences in the Callaway FSAR 
for the steam generator tube rupture (SGTR) and main steam line break 
(MSLB) accidents.
    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.
    This change increases the offsite dose consequences for the MSLB 
and SGTR accidents reported in FSAR Sections 15.1 and 15.6. Non-
conservative assumptions regarding letdown flow rate, iodine 
isotopic mix in the source term, resin effeciency, and termination 
of the flash release pathway were identified in the SGTR and MSLB 
radiological consequence analyses. The correction of these non-
conservative assumptions results in an increase in the radiological 
consequences reported in FSAR Tables 15.1-4 and 15.6-5. However, 
these increases are not significant since the new values remain less 
than the 10 CFR 100.11 regulatory requirements and the guideline 
values provided by the Standard Review Plan [NUREG-0800].
    There will be no increase in the probability of previously 
evaluated accidents. This change only involves the modeling and 
calculation of the SGTR and MSLB radiological consequences. [There 
are no equipment or system changes.] Protection system performance 
will remain within the assumptions of the previously performed 
accident analyses since no hardware changes are proposed. The 
protection systems will continue to function in a manner consistent 
with the plant design basis. The proposed change will not affect the 
probability of any event initiators nor will the proposed change 
affect the ability of any safety-related equipment to perform its 
intended function. There will be no degradation in the performance 
of, nor an increase in the number of challenges imposed on, safety-
related equipment assumed to function during an accident situation. 
There will be no change to normal plant operating parameters or 
accident mitigation performance.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This change is the result of a re-analysis of the MSLB and SGTR 
radiological consequences. These accidents were previously analyzed 
in the FSAR. None of the changes in the dose calculation modeling 
create the possibility of a new or different kind of accident.
    There are no hardware changes associated with this amendment 
application nor are there any changes in the method by which any 
safety-related plant system performs its safety function. The change 
will not affect the normal method of plant operation, other than the 
imposition of administrative limits on the concentrations of I-134 
[Iodine-134] and Dose Equivalent I-131 until this amendment 
application is approved by NRC. No new accident scenarios, transient 
precursors, failure mechanisms, or limiting single failures are 
introduced as a result of this change. There will be no adverse 
effect or challenges imposed on any safety-related system as a 
result of this change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The re-analysis of the MSLB and SGTR radiological consequences, 
and the resultant increase in consequences reported in FSAR Tables 
15.1-4 and 15.6-5, ensures that the accident analyses support the 
plant operating conditions allowed by current Technical 
Specification 3.4.8, Reactor Coolant System Specific Activity (ITS 
[Improved Technical Specification] 3.4.16), and current Technical 
Specification 3.7.1.4, Plant Systems Specific Activity (ITS 3.7.18).
    The proposed change does not affect the acceptance criteria for 
any analyzed event nor is there a change to any Safety Analysis 
Limit (SAL). There will be no effect on the manner in which safety 
limits or limiting safety system settings are determined nor will 
there be any effect on those plant systems necessary to assure the 
accomplishment of protection functions. There will be no impact on 
the overpower limit, DNBR [departure from nucleate boiling ratio], 
FQ [heat flux hot channel factor], FdeltaH [nuclear 
enthalpy rise hot channel factor], LOCA PCT [peak cladding 
temperature for the loss-of-coolant accident], peak local power 
density, or any other margin of safety. The radiological dose 
consequence acceptance criteria listed in the Standard Review Plan 
continue to be met.
    Therefore, the proposed change does not involve a significant 
reduction in any 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.
    Local Public Document Room location: Elmer Ellis Library, 
University of Missouri, Columbia Missouri 65201.
    Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

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 54384]]

page cited. This notice does not extend the notice period of the 
original notice.

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

    Date of amendment requests: September 14, 1998.
    Description of amendment requests: The proposed amendments would 
change the runout limits for a safety injection (SI) pump to 675 
gallons per minute (gpm), unless the pump is specifically tested to a 
higher flow rate, not exceeding 700 gpm for both Units 1 and 2. This 
change was initiated upon reevaluation of correspondence from 
Westinghouse sent to the licensee in 1991, which indicated that the 
generic runout limits for Pacific 2'' JTCH pumps was 675 gpm unless 
each specific pump is tested to a higher flow rate. Individual testing 
is necessary due to test variations between pumps which may limit the 
applicability of testing of one pump to another pump due to 
manufacturing tolerances in the sand cast impellers and material 
changes in the pump casing.
    Furthermore, the bases section is being clarified to describe why 
the injection rather than the recirculation mode during flow balancing 
is the minimum resistance and, consequently, more conservative 
configuration for runout considerations.
    Date of publication of individual notice in Federal Register: 
August 31, 1999 (64 FR 47533).
    Expiration date of individual notice: September 30, 1999
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, MI 49085.
    Date of publication of individual notice in Federal Register: 
August 31, 1999 (64 FR 47533).
    Expiration date of individual notice: September 30, 1999.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, MI 49085.

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

    Date of application for amendments: October 8, 1998.
    Brief description of amendments: The amendments would revise 
Technical Specification (TS) 3.3.3.8 for Unit 1 and TS 3.3.3.6 for Unit 
2, ``Post-Accident Instrumentation.'' The proposed changes to the TSs 
will place tighter restrictions on the amount of time the refueling 
water storage tank (RWST) water level instrumentation may be inoperable 
before the limiting conditions for operation in the TSs are applied.
    Date of publication of individual notice in Federal Register: 
August 31, 1999 (64 FR 47532).
    Expiration date of individual notice: September 30, 1999.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, MI 49085.

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

    Date of application for amendments: December 3, 1998.
    Brief description of amendments: The amendments would make 
administrative changes to several Technical Specifications to remove 
obsolete information, provide consistency between Unit 1 and Unit 2, 
provide consistency with the Standard Technical Specifications, provide 
clarification, and correct typographical errors.
    Date of publication of individual notice in Federal Register: 
August 31, 1999 (64 FR 47535).
    Expiration date of individual notice: September 30, 1999.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, MI 49085.

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

    Date of application for amendments: May 21, 1999.
    Brief description of amendments: The amendments would change the 
Technical Specifications (T/S) to allow reactor coolant system 
temperature changes in certain Mode 5 and 6 action statements if the 
shutdown margin is sufficient to accommodate the expected temperature 
change. In addition, footnotes regarding additions of water from the 
refueling water storage tank to the reactor coolant system are 
clarified and relocated to action statements. Additional actions are 
added in Table 3.3-1, ``Reactor Trip System Instrumentation,'' when the 
required source range neutron flux channel is inoperable. Corresponding 
changes are proposed for the bases for T/S 3/4.1.1, ``Boration 
Control,'' and T/S 3/4.1.2, ``Boration Systems.'' Administrative 
changes are proposed to improve clarity. Finally, additions are made to 
shutdown margin T/S surveillance requirements to address use of a boron 
penalty (requirement for additional boron) during residual heat removal 
system operation in Modes 4 and 5.
    Date of publication of individual notice in Federal Register: July 
12, 1999 (64 FR 37574).
    Expiration date of individual notice: August 11, 1999.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, MI 49085.

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 at the local public document rooms for 
the particular facilities involved.

[[Page 54385]]

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

    Date of application for amendment: September 23, 1998, as 
supplemented on December 7, 1998, and August 10, 1999.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) 3/4.6.1.3, ``Containment Air Locks,'' and its 
associated bases, to clarify the requirements for locking an air lock 
door shut and to make it consistent with NUREG-1431, Revision 1, 
``Standard Technical Specifications, Westinghouse Plants,'' dated April 
1995.
    Date of issuance: September 14, 1999.
    Effective date: September 14, 1999.
    Amendment No.: 90.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications
    Date of initial notice in Federal Register: October 21, 1998 (63 FR 
56239)
    The December 7, 1998, and August 10, 1999, submittals contained 
clarifying information only, and did not change the initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 14, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

    Date of application for amendment: June 15, 1999.
    Brief description of amendment: This amendment changes the 
Technical Specifications to incorporate the performance-based 10 CFR 50 
Appendix J, Option B for Type A tests (containment integrated leakage 
rate tests). Option B will be implemented for Type A testing in 
accordance with NRC Regulatory Guide 1.163, ``Performance-Based 
Containment Leak-Test Program,'' dated September 1995, and Nuclear 
Energy Institute (NEI) Guideline 94-01, Revision 0, ``Industry 
Guideline for Implementing Performance-Based Option of 10 CFR Part 50, 
Appendix J,'' dated July 26, 1995. Type B and C testing (containment 
penetration leakage tests) will continue to be performed in accordance 
with 10 CFR 50 Appendix J, Option A.
    Date of issuance: September 17, 1999.
    Effective date: September 17, 1999.
    Amendment No.: 91.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38023). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 17, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

    Date of application for amendments: July 30, 1999.
    Brief description of amendments: The amendments changed the maximum 
allowable temperature of the ultimate heat sink in the technical 
specifications from 98 degrees Fahrenheit to 100 degrees Fahrenheit. 
The change is in effect from the date of this amendment until September 
30, 1999.
    Date of issuance: September 8, 1999.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 103 and 103.
    Facility Operating License Nos. NPF-72 and NPF-77: The amendments 
revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (64 FR 44962 dated August 18, 1999). The notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by September 17, 1999, but indicated that if the Commission 
makes a final no significant hazards consideration determination any 
such hearing would take place after issuance of the amendments. The 
Commission's related evaluation of the amendments, finding of exigent 
circumstances and final no significant hazards consideration 
determination are contained in a Safety Evaluation dated September 8, 
1999.
    Local Public Document Room location: Wilmington Public Library, 201 
S. Kankakee Street, Wilmington, Illinois 60481.

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: May 3, 1999, as supplemented by 
letter dated September 10, 1999.
    Brief description of amendments: The amendments relocated the 
requirements of Technical Specification (TS) Section 3/4.6.I to the 
Updated Final Safety Analysis Report (UFSAR). TS Section 
3/4.6.I contains reactor coolant chemistry limiting conditions for 
operation (LCO) and surveillance requirements (SR) for conductivity, 
chloride concentration, and pH.
    Date of issuance: September 23, 1999.
    Effective date: Immediately, to be implemented within 30 days 
including relocation of the removed TSs and associated bases to the 
licensee's UFSAR pending change file. In addition, the licensee shall 
include the relocated information in the UFSAR submitted to the NRC, 
pursuant to 10 CFR 50.71(e), except for any information that has been 
changed in accordance with 10 CFR 50.59 and described in the change 
summaries submitted to NRC pursuant to 10 CFR 50.59.
    Amendment Nos.: 173 & 169.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43768). The September 10, 1999, submittal provided additional 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 23, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.

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

    Date of application for amendments: June 29, 1999.
    Brief description of amendments: The amendments increased the notch 
testing surveillance interval of partially withdrawn control rods in 
Technical Specification Surveillance Requirement 3/4.3.C, ``Reactivity 
Control--Control Rod Operability,'' from an interval of once in 7 days 
to once in 31 days.
    Date of issuance: September 23, 1999.

[[Page 54386]]

    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 190 & 187.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 28, 1999 (64 FR 
40905).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 23, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.

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

    Date of application of amendments: May 24, 1999
    Brief description of amendments: The amendments revise the maximum 
local fuel pin centerline temperature safety limit in Technical 
Specification 2.1.1.1 from the limit determined using the TACO2 fuel 
performance computer code to the value determined using a newer TACO3 
computer code.
    Date of Issuance: September 24, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--306, Unit 2--306, Unit 3--306.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35203).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 24, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.

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

    Date of application for amendments: July 22, 1998, and supplemented 
by letters dated October 22, 1998, January 28, May 6, June 24, August 
17 and September 15, 1999.
    Brief description of amendments: The amendments revise various 
sections of the Technical Specifications (Appendix A of the Catawba 
operating licenses) to permit use of Westinghouse's Robust Fuel 
Assemblies for future core reloads.
    Date of issuance: September 22, 1999.
    Effective date: As of the date of issuance and shall be implemented 
prior to beginning the installation of the Westinghouse fuel, currently 
projected to be Fuel Cycle 13 and 11 for Units 1 and 2, respectively.
    Amendment Nos.: Unit 1--180; Unit 2--172.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 64108); May 19, 1999 (64 FR 27317); August 11, 1999 (64 FR 43770) 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated September 22, 1999.
    No significant hazards consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

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

    Date of application for amendments: July 22, 1998, and supplemented 
by letters dated October 22, 1998, and January 28, May 6, June 24, 
August 17 and September 15, 1999
    Brief description of amendments: The amendments revise various 
sections of the Technical Specifications (Appendix A of the McGuire 
operating licenses) to permit use of Westinghouse's Robust Fuel 
Assemblies for future core reloads.
    Date of issuance: September 22, 1999.
    Effective date: As of the date of issuance and shall be implemented 
prior to beginning the installation of the Westinghouse fuel, currently 
projected to be Fuel Cycle 15 and 14 for Units 1 and 2, respectively.
    Amendment Nos.: Unit 1--188; Unit 2--169.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43771); June 30, 1999 (64 FR 35202); December 16, 1998 (64 FR 69388)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 22, 1999.
    No significant hazards consideration comments received: No
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, Charlotte, North Carolina

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

    Date of amendment request: April 9, 1999, as supplemented by letter 
dated July 29, 1999
    Brief description of amendment: The amendment revises the 
requirements associated with the station batteries and the direct 
current (DC) sources to the 125 volt DC switchyard distribution system.
    Date of issuance: September 14, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days from the date of issuance (including issuance of the 
Technical Requirements Manual for use by licensee personnel).
    Amendment No.: 200.
    Facility Operating License No. DPR-51: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 19, 1999 (64 FR 
27321).
    The July 29, 1999, letter provided clarifying and additional 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 14, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

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

    Date of amendment request: June 1, 1999, as supplemented by letters 
dated July 29 and August 19, 1999.
    Brief description of amendment: The amendment revised the Technical 
Specifications to allow, under specific conditions, certain once-
through steam generator (OTSG) tubes with tube end crack indications 
adjacent to the primary cladding region of the upper and lower OTSG 
tubesheets to remain in service.
    Date of issuance: September 14, 1999.
    Effective date: As of the date of issuance and shall be implemented 
prior to reactor startup after refueling outage 1R15.
    Amendment No.: 201.
    Facility Operating License No. DPR-51: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35205).
    The July 29 and August 19, 1999, letters provided clarifying 
information

[[Page 54387]]

that did not change the scope of the June 1, 1999, application and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 14, 1999.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

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

    Date of application for amendment: May 17, 1999.
    Brief description of amendment: The amendment changes Technical 
Specification Section 3.3.8, ``Emergency Diesel Generator Loss of Power 
Start,'' Surveillance Requirement 3.3.8.1 and corresponding basis 
section. The surveillance is revised to make a note included in the 
surveillance consistent with the method of performing the surveillance.
    Date of issuance: September 13, 1999.
    Effective date: September 13, 1999.
    Amendment No.: 187.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38026).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 13, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.

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

    Date of application for amendment: December 23, 1998.
    Brief description of amendment: The proposed amendment revised the 
surveillance frequency for verifying the operability of motor-operated 
isolation valves and condensate makeup valves in the Isolation 
Condenser Technical Specification 4.8.A.1 and Bases page from once per 
month to once per 3 months.
    Date of Issuance: September 24, 1999.
    Effective date: Date of issuance and shall be implemented within 30 
days of issuance.
    Amendment No.: 209.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 7, 1999 (64 FR 
17026).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated September 24, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear 
Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of application for amendment: October 19, 1998, as 
supplemented August 19, 1999.
    Brief description of amendment: The proposed amendment adds 
operability and surveillance requirements to the Technical 
Specifications for the remote shutdown system similar to the standard 
technical specifications for Babcock & Wilcox nuclear plants as 
described in NUREG-1430.
    Date of issuance: September 22, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 216.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 64118). The August 19, 1999, supplement to the application did not 
change the staff's proposed no significant hazards consideration 
determination or expand the scope of the application as originally 
noticed.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 22, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (Regional Depository) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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

    Date of application for amendment: April 30, 1999.
    Brief description of amendment: The amendment revises Duane Arnold 
Energy Center (DAEC) Technical Specification (TS) Surveillance 
Requirement (SR) 3.4.3.1 to revise the safety function lift setpoint 
tolerance limits for the main safety valves (SVs) and the safety/relief 
valves (SRVs).
    Date of issuance: September 22, 1999.
    Effective date: September 22, 1999, to be implemented within 30 
days.
    Amendment No.: 228.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38028).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 22, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, IA 52401.

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

    Date of application for amendment: May 15, 1998, as supplemented by 
letters dated September 25, October 13, December 9 (two letters), 1998; 
January 11, April 1, and April 22, 1999.
    Brief description of amendment: This amendment changes Technical 
Specification (TS) 5.5, ``Storage of Unirradiated and Spent Fuel,'' to 
reflect a planned modification to increase the storage capacity of the 
spent fuel pool from 2776 to 4086 fuel assemblies. It also deletes an 
inappropriate statement and reference within TS 5.5.
    Date of issuance: June 17, 1999.
    Effective date: This license amendment is effective as of the date 
of its issuance to be implemented before spent fuel is stored within 
the new high-density spent fuel rack modules authorized for 
installation and use by this amendment.
    Amendment No.: 167.
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1998 (63 
FR 64973).
    The September 25, October 13, December 9 (two letters) 1998, 
January 11, April 1, and April 22, 1999, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 17, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents

[[Page 54388]]

Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of amendment request: June 23, 1999.
    Description of amendment request: To revise Technical Specification 
(TS) 3.7.6.2 to increase the allowable outage time for the Control Room 
Air Conditioning Subsystem from 30 days to 60 days, on a one-time basis 
for each train, to allow adequate time to replace portions of the 
existing system during the current operating cycle, and to exclude the 
requirements of TS 3.0.4 and TS 4.0.4 during the implementation of the 
modification.
    Date of issuance: September 17, 1999.
    Effective date: As of its date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 62.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications/License.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38032).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 17, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.

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: March 17, 1999.
    Brief description of amendment: The amendment changes Technical 
Specifications 3.5.2, ``Emergency Core Cooling Systems--ECCS 
Subsystems--Tavg 300  deg.F;'' 3.7.1.7, ``Plant Systems--
Atmospheric Steam Dump Valves;'' and 3.7.6.1, ``Plant Systems--Control 
Room Emergency Ventilation System.'' The changes will revise: (1) 
Surveillance requirements for the Emergency Core Cooling System valves, 
(2) the atmospheric steam dump valve requirements to focus on the steam 
release path instead of the individual valves, and (3) the allowed 
outage time for the atmospheric steam valves and Control Room Emergency 
Ventilation System. The licensee made changes to the Bases pages 
consistent with the proposed changes to the TSs.
    Date of issuance: August 12, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 238.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 21, 1999 (64 FR 
19559).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 12, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

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

    Date of application for amendment: June 4, 1999.
    Brief description of amendment: The amendment makes administrative 
changes to the Technical Specifications.
    Date of issuance: September 14, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 193.
    Facility Operating License No. DPR-64: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 28, 1999 (64 FR 
40906).
    No significant hazards consideration comments received: No.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 14, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: January 28, 1999, as 
supplemented April 29, 1999, and May 17, 1999. By letters dated April 
29, 1999, and May 17, 1999, the licensee revised the original submittal 
dated January 28, 1999, in response to questions raised by the NRC 
staff.
    Brief description of amendment: The amendment changes the Technical 
Specifications by reducing the number of emergency diesel generators 
required to be operable under certain conditions.
    Date of issuance: September 14, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 194.
    Facility Operating License No. DPR-64: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 2, 1999 (64 FR 
29713). This notice superceded a notice dated April 21, 1999 (64 FR 
19563).
    No significant hazards consideration comments received: No.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 14, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: January 28, 1999, as 
supplemented July 16, 1999.
    Brief description of amendment: The amendment removes lists of 
containment isolation valves from the Technical Specifications (TSs) 
and modifies the TSs accordingly.
    Date of issuance: September 16, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 195.
    Facility Operating License No. DPR-64: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 19, 1999 (64 FR 
27323).
    The July 16, 1999, submittal did not change the staff's initial 
proposed finding of no significant hazards considerations.
    No significant hazards consideration comments received: No.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 16, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

[[Page 54389]]

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

    Date of application for amendment: April 5, 1999.
    Brief description of amendment: The proposed changes would revise 
Appendix A (Section 6.1) and Appendix B (Section 7.1) of the James A. 
FitzPatrick Technical Specifications. The proposed changes would remove 
the position title of General Manager from these sections and would 
state that if the Site Executive Officer is unavailable, he will 
delegate his responsibilities to another staff member, in writing. In 
addition the position title of Resident Manager, used in Appendix B, 
Section 7.1, would be replaced by Site Executive Officer.
    Date of issuance: September 13, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 254.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications and the Environmental Technical 
Specifications.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43775).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 13, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: October 8, 1997.
    Brief description of amendment: The amendment revises actions in 
the Technical Specifications to be taken in the event multiple control 
rods are inoperable.
    Date of issuance: September 21, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 255.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 11, 1998 (63 
FR 6991).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 21, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: December 30, 1998, as 
supplemented September 13, 1999.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) Limiting Condition for Operation 3.7.3 and TS Table 
3.7.3-1. These changes modify the flood protection actions required 
when severe storm warnings that may affect the site are in effect or 
during periods of elevated river water level.
    Date of issuance: September 17, 1999.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 122.
    Facility Operating License No. NPF-57: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: February 24, 1999 (64 
FR 9200).
    The September 13, 1999, supplement provided clarifying information 
that did not change the initial proposed no significant hazards 
determination or expand the scope of the initial Federal Register 
notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 17, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.

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

    Date of application for amendment: May 24, 1999, as supplemented 
June 21, 1999.
    Brief description of amendment: This amendment revises the 
Technical Specifications (TSs) to correct typographical and editorial 
errors, and is considered administrative in nature.
    Date of issuance: September 21, 1999
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 123.
    Facility Operating License No. NPF-57: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35209).
    The June 21, 1999, supplement provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination or expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 21, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of application for amendments: July 2, 1999.
    Brief description of amendments: The amendments delete TS 3/4.3.4, 
``Instrumentation--Turbine Overspeed Protection,'' and its associated 
Bases and relocate the requirements to the licensee-controlled Updated 
Final Safety Analysis Report.
    Date of issuance: September 14, 1999
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 224 and 205.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43776).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 14, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

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

    Date of application for amendments: December 31, 1998 (PCN-501), as 
supplemented June 14, 1999.
    Brief description of amendments: The amendments consist of changes 
to Technical Specification 3.3.5, ``Engineered Safety Features 
Actuation System (ESFAS) Instrumentation,'' and will include 
restrictions on operation with a channel of the refueling water storage 
tank level--low input to the recirculation actuation signal and the 
steam generator pressure--low input or

[[Page 54390]]

steam generator pressure difference--high input to the emergency 
feedwater actuation signal in the tripped condition.
    Date of issuance: September 7, 1999.
    Effective date: September 7, 1999, to be implemented within 30 days 
of issuance.
    Amendment Nos.: Unit 2--157; Unit 3--148.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 28, 1999 (64 FR 
40907).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 7, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.

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

    Date of application for amendments: June 18, 1997 (PCN-478), as 
supplemented May 24 and August 10, 1999.
    Brief description of amendments: The amendments modify the 
Technical Specification surveillance requirements related to diesel 
generator testing to more clearly reflect safety analysis and testing 
conditions as it is performed.
    Date of issuance: September 9, 1999.
    Effective date: September 9, 1999, to be implemented within 30 days 
of issuance.
    Amendment Nos.: Unit 2--158; Unit 3--149.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 31, 1997 (62 
FR 68315) The licensee's letters dated May 24 and August 10, 1999, 
provided updated Technical Specification pages, clarifications, and 
additional information that were within the scope of the original 
Federal Register notice and did not change the staff's initial proposed 
no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 9, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Main Library, University of 
California, P.O. Box 19557, Irvine, California 92713.

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

    Date of amendment request: June 7, 1999.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 2.2.1, ``Reactor Trip System (RTS) Instrumentation 
Setpoints,'' and TS 3.3.2, ``Engineered Safety Features Actuation 
System (ESFAS) Instrumentation,'' and the associated Bases, by removing 
the Total Allowance, Sensor Error, and Z terms (Z is the statistical 
summation of errors excluding sensor and rack drift) from the RTS and 
ESFAS Instrumentation Trip Setpoints Tables. This replaces the five-
column methodology with a two-column methodology that consists of the 
trip setpoint and allowable value columns.
    Date of issuance: September 13, 1999.
    Effective date: September 13, 1999, to be implemented within 30 
days.
    Amendment Nos.: Unit 1--116; Unit 2--104.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35211) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 13, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.

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 24, 1999 (TS 99-06).
    Brief description of amendments: The amendments revise the Sequoyah 
Nuclear Plant Technical Specifications (TS) by adding a footnote to 
allow use of an installed spare electrical inverter, if needed.
    Date of issuance: September 23, 1999.
    Effective date: As of the date of issuance to be implemented no 
later than 45 days after issuance.
    Amendment Nos.: 246 and 237.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TS.
    Date of initial notice in Federal Register: August 2, 1999 (64 FR 
41973) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 23, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

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

    Date of amendment request: June 23, 1999, as supplemented by letter 
dated August 4, 1999.
    Brief description of amendments: The amendments revise Surveillance 
Requirement 3.8.1.13, ``AC Sources--Operating'' to clarify that each 
emergency diesel generator automatic noncritical trip, except for 
engine overspeed and generator differential current, is bypassed on 
either a loss-of-offsite power or a safety injection actuation signal.
    Date of issuance: September 21, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 69 and 69.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38037) The August 4, 1999, letter provided additional and clarifying 
information that did not change the scope of the June 23, 1999, 
application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 21, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.

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

    Date of amendment request: February 12, 1999, as supplemented by 
letter dated June 14, 1999
    Brief description of amendments: The amendments change Technical 
Specification (TS) 3.4.13, ``RCS [Reactor Coolant System] Operational 
Leakage,'' TS 5.5.9, ``Steam Generator (SG) Tube Surveillance 
Program,'' and TS 5.6.10,

[[Page 54391]]

``Steam Generator Tube Inspection Report,'' to implement the 1.0 Volt 
Steam Generator Tube Repair Criteria for CPSES, Unit 1.
    Date of issuance: September 22, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--Amendment No. 70; Amendment No. 70.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24202) The June 14, 1999, supplement provided clarifying information 
that did not change the scope of the February 12, 1999, application and 
the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 22, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.

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

    Date of amendment request: October 2, 1998, as supplemented by 
letters dated July 27 and August 26, 1999.
    Brief description of amendments: The amendments revise Technical 
Specfications for CPSES, Unit 1, to define the F* steam generator tube 
plugging criteria in TS 5.5.9, ``Steam Generator (SG) Tube Surveillance 
Program,'' and associated reporting requirements in TS 5.6.10, ``Steam 
Generator Inspection Report.''
    Date of issuance: September 22, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--Amendment No. 71; Unit 2--Amendment No. 71.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 4, 1998 (63 FR 
59597). The July 27 and August 26, 1999, letters provided clarifying 
information that did not change the scope of the October 2, 1998, 
application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 22, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.

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

    Date of application for amendment: May 5, 1999.
    Brief description of amendment: The amendment revises the technical 
specifications (TSs) to enhance the limiting conditions for operation 
and surveillance requirements relating to the standby liquid control 
system and to incorporate certain provisions of NRC's rule on 
anticipated transients without scram. The change involves the use of 
enriched boron in the standby liquid control system and improves upon 
other aspects of the TSs for this system.
    Date of Issuance: September 17, 1999.
    Effective date: September 17, 1999, and shall be implemented within 
30 days.
    Amendment No.: 175.
    Facility Operating License No. DPR-28. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35214).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated September 17, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

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

    Date of application for amendment: July 12, 1999.
    Brief description of amendment: The amendment revises the values 
for the minimum critical power ratio safety limits and deletes the 
wording classifying the limits as cycle-specific values.
    Date of Issuance: September 21, 1999.
    Effective date: September 21, 1999, and shall be implemented within 
60 days.
    Amendment No.: 176
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 28, 1999 (64 FR 
40910).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated September 21, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for

[[Page 54392]]

example, in derating or shutdown of a nuclear power plant or in 
prevention of either resumption of operation or of increase in power 
output up to the plant's licensed power level, the Commission may not 
have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 
application for amendment, (2) the amendment to Facility Operating 
License, 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 at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By November 5, 1999, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC and at the local public document room for the 
particular facility involved. 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. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    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: Rulemakings and 
Adjudications Staff 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 the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

[[Page 54393]]

Duke Energy Corporation, Docket No. 50-369, McGuire Nuclear Station, 
Unit 1, Mecklenberg County, North Carolina

    Date of application for amendment: August 27, 1999.
    Brief description of amendment: The amendment approves a one-time 
extension of the surveillance frequency for Technical Specifications 
Surveillance Requirement (TSSR) 3.1.4.2 beyond the 25 percent extension 
allowed by TSSR 3.0.2 to the McGuire Nuclear Station, Unit 1. This 
license amendment is effective upon issuance and is to expire upon 
entering Mode 3 during Unit 1 startup following the Unit 1 End of Cycle 
13 refueling outage.
    Date of issuance: September 8, 1999.
    Effective date: As of its date of issuance (September 8, 1999), and 
shall expire upon entering Mode 3 during startup, following the End of 
Cycle 13 refueling outage.
    Amendment No.: Unit 1-186.
    Facility Operating License No. NPF-9: Amendments revised the 
Technical Specifications.
    Press release issued requesting comments as to proposed no 
significant hazards consideration: Yes, September 2, 1999, Charlotte 
Observer.
    Comments received: No.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, consultation with the State of North Carolina, 
and final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated September 8, 1999.
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, Charlotte, North Carolina.
    Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina 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 application for amendments: September 13, 1999.
    Brief description of amendments: The amendments revise the 
Technical Specifications TS 3.7.9, ``Control Room Area Ventilation 
System (CRAVS),'' to establish actions to be taken for an inoperable 
control room ventilation system due to a degraded control room pressure 
boundary. This revision approves changes that would allow up to 24 
hours to restore the Control Room Pressure Boundary (CRPB) to operable 
status when two CRAVS trains are inoperable due to an inoperable CRPB 
in MODES 1, 2, 3, and 4. In addition, a Limiting Condition for 
Operation note would be added to allow the CRPB to be opened 
intermittently under administrative control without affecting CRAVS 
operability.
    Date of issuance: September 22, 1999.
    Effective date: As of the date of issuance and shall be implemented 
upon receipt.
    Amendment Nos.: Unit 1--187; Unit 2--168.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Press release issued requesting comments as to proposed no 
significant hazards consideration: Yes, September 17, 1999, Charlotte 
Observer.
    Comments received: No.
    The Commission's related evaluation and the amendment, finding of 
emergency circumstances, consultation with the State of North Carolina, 
and final no significant hazards consideration determination are 
contained in a Safety Evaluation dated September 22, 1999.
    Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina.
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, Charlotte, North Carolina.
    NRC Section Chief: Richard L. Emch, Jr.

    For the Nuclear Regulatory Commission.

    Dated at Rockville, Maryland, this 29th day of September, 1999.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 99-25795 Filed 10-5-99; 8:45 am]
BILLING CODE 7590-01-P