[Federal Register Volume 64, Number 212 (Wednesday, November 3, 1999)]
[Notices]
[Pages 59796-59813]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-28598]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from October 8, 1999, through October 22, 1999. 
The last biweekly notice was published on October 20, 1999 (64 FR 
56526).

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

[[Page 59797]]

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 6D59, 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 December 10, 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 electronically from 
the ADAMS Public Library component on the NRC Web site, 
http://www.nrc.gov (the Electronic Reading Room). If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the Nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: 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 electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

Carolina Power & Light Company, et al., Docket No. 50-325, 
Brunswick Steam Electric Plant, Unit 1, Brunswick County, North 
Carolina

    Date of amendment request: September 28, 1999.
    Description of amendment request: The licensee has proposed to 
revise Technical Specification (TS) 2.1.1, ``Reactor Core Safety 
Limits,'' and TS 5.6.5, ``Core Operating Limits Report.'' These 
revisions would remove cycle-specific safety limit restrictions which 
are no longer necessary.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 59798]]

issue of no significant hazards consideration, which is presented 
below:

    1. The proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The procedures for determining the MCPR [Minimum Critical Power 
Ratio] Safety Limit are described in General Electric Standard 
Application for Reactor Fuel (i.e., topical report NEDE-24011-P-A, 
otherwise referred to as GESTAR II). The basis for the MCPR Safety 
Limit calculation is to ensure that greater than 99.9 percent of all 
fuel rods in the core avoid transition boiling in the event of a 
postulated accident. The existing MCPR Safety Limit preserves this 
margin to transition boiling and fuel damage. The MCPR Safety Limits 
for the BSEP [Brunswick Steam Electric Plant], Unit 1 TSs, and their 
use in determining cycle-specific operating limits documented in the 
Core Operating Limits Report, are determined using NRC-approved 
methods (i.e., GESTAR II). The use of these methods ensures that the 
MCPR Safety Limit values are within the existing design and 
licensing bases, and cannot increase the probability or consequences 
of an accident previously evaluated.
    2. The proposed license amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The MCPR Safety Limit is a TS numerical value that has been 
established to ensure that fuel damage from transition boiling does 
not occur in at least 99.9 percent of the fuel rods in the core as a 
result of a limiting postulated accident. The MCPR Safety Limit is 
not an accident initiator; therefore, it cannot create the 
possibility of any new type of accident. The MCPR Safety Limits are 
calculated using NRC-approved methods. The function, location, 
operation, and handling of the fuel will remain unchanged. In 
addition, the initiating sequence of events for previously evaluated 
accidents has not been changed. Therefore, no new or different kind 
of accident has been created.
    3. The proposed license amendment does not involve a significant 
reduction in a margin of safety.
    The MCPR Safety Limit preserves the existing margin to 
transition boiling and fuel damage in the event of a postulated 
accident. The margin of safety, as defined in the TS Bases, will 
remain the same. The MCPR Safety Limit remains unchanged, and will 
ensure that greater than 99.9 percent of all fuel rods in the core 
will avoid transition boiling if the limit is not violated, thereby 
preserving the fuel cladding integrity. The MCPR Safety Limits will 
continue to be calculated using NRC-approved generic and cycle-
specific methodologies that are described in GESTAR II. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602
    NRC Section Chief: Ron Hernan, Acting.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson 
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: September 28, 1999.
    Description of amendment request: The amendment revises Technical 
Specifications (TS) surveillance requirement (SR) 3.7.6.2 ``Component 
Cooling Water (CCW) System,'' to change the CCW pump automatic start 
actuation signal basis from Engineered Safety Feature Actuation Signal 
(ESFAS) to Loss-of-Power Diesel Generator (LOP DG). This change is 
required to reflect the original plant design which was not properly 
incorporated during conversion of the TS to Improved TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Carolina Power & Light (CP&L) Company has evaluated the proposed 
Technical Specification change and has concluded that it does not 
involve a significant hazards consideration. The CP&L conclusion is 
in accordance with the criteria set forth in 10 CFR 50.92. The bases 
for the conclusion that the proposed change does not involve a 
significant hazards consideration are discussed below.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change to Surveillance Requirement (SR) 3.7.6.2 
does not involve any physical alteration of plant systems, 
structures or components, changes in parameters governing normal 
plant operation, or methods of operation. The safety function of the 
Loss of Power (LOP) Diesel Generator (DG) start signal for the 
Component Cooling Water (CCW) pumps is to start the CCW pumps in 
order to provide the minimum heat removal capability assumed in the 
safety analysis for the systems to which it supplies cooling water. 
The CCW System provides a heat sink for the removal of process and 
operating heat from safety related components during a Design Basis 
Accident (DBA) or transient. During normal operation, the CCW System 
also provides this function for various nonessential components, as 
well as the spent fuel storage pool. The CCW System serves as a 
barrier to the release of radioactive byproducts between potentially 
radioactive systems and the Service Water System, and thus to the 
environment. The CCW pumps start upon receipt of a LOP DG start 
signal from undervoltage on the emergency bus. The LOP DG start 
signal to the CCW pumps is not an Engineered Safety Features 
Actuation System (ESFAS) signal. Since this proposed change only 
corrects the description of the start signal, the proposed change 
does not involve an increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve any physical alteration of 
plant systems, structures or components, changes in parameters 
governing normal plant operation, or methods of operation. The 
proposed change does not introduce a new mode of operation or 
changes in the method of normal plant operation. Therefore, the 
possibility of a new or different kind of accident from any accident 
previously evaluated is not created.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change corrects the word description of the start 
signal for the CCW pumps and does not alter any plant design margin 
or analysis assumption as described in the Updated Safety Analysis 
Report. The proposed change does not affect any limiting safety 
system setpoint, calibration method, or setpoint calculation. 
Therefore, the proposed change does not involve a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602 .
    NRC Section Chief: Sheri R. Peterson.

CBS Corporation (licensee), Westinghouse Test Reactor, Waltz Mill 
Site, Westmoreland, Pennsylvania, Docket No. 50-22, License No. TR-
2

    Date of amendment request: September 15, 1999, as supplemented on 
October 4, 1999.
    Description of amendment request: CBS Corporation is the licensee 
for the Westinghouse Test Reactor (WTR) at Waltz Mill, Pennsylvania. 
The licensee is authorized to only possess the reactor and a 
decommissioning plan has been approved.
    The licensee is planning to revise four Technical Specifications 
(TS) in their approved Decommissioning Plan. The

[[Page 59799]]

first TS change deals with what doors need to be closed when restricted 
activities are taking place within containment. Access to containment 
is through three locations, i.e., the truck lock door and the east and 
west airlock doors. Each entry point has two doors, an outer door and 
an inner door. In the existing TS either door could be closed except 
during personnel ingress or egress or while equipment is being passed 
through the doorways. In the proposed TS the licensee has specified the 
following. For the truck lock door the inner door to containment needs 
to be closed. The reason given for the change is that the containment 
boundary is more accurately defined as the interior access door between 
the truck lock area and containment. The truck lock area was 
transferred to the SNM-770 license in April 1970 and the outer doors 
are controlled by this license.
    For the east and west airlock doors, fire doors with an interior 
crash bar have been installed at the outer door as a safety feature to 
minimize the risk of personnel being trapped in containment during an 
emergency. The airlock doors (inner doors) do not allow quick and 
efficient egress during a postulated fire in containment; therefore, 
the original air lock doors have been removed and confinement is 
maintained by the newly installed fire doors.
    Therefore the proposed TS require that the inner truck lock door be 
closed and the outer east and west lock doors be closed except during 
personnel ingress or egress or while equipment is being passed through 
the doorways, and this meets the original goal of the existing TS.
    The second TS change deals with the condition of the containment 
when the containment is open for removal of materials and equipment. In 
the existing TS Restricted Activities in containment are suspended. In 
the proposed TS, containment extension is permitted if an enclosure is 
provided around the opening to effectively isolate the containment from 
the outside environment. If these extensions are not in place, all 
Restricted Activities in containment are suspended. Negative pressure 
(airflow into containment) is maintained in containment in the existing 
as well as the proposed TS. Containment isolation is effectively 
maintained under the proposed TS as it was in the existing TS.
    The third TS change deals with the control of access into 
containment. In the existing TS the outer doors in the air lock and the 
truck lock outer doors shall be locked or blocked closed to prevent 
unauthorized entry except when authorized personnel are inside the 
containment building or outside with the door in view. In the proposed 
TS access into containment is through a Health Physics (HP) control 
point, which is on the first floor of the G-Building. To prevent 
unauthorized entry the accesses into and out of containment shall be 
locked or blocked closed except when this access control point is 
supervised and the provisions of the first TS change are implemented.
    Normal access to the containment is through a door in the G-
Building basement (east and west airlock doors). The G-Building 
basement is a ``Radiation Area''. Routine activities during the day may 
require workers to exit containment (rest, lunch, equipment change out, 
etc). Locking or blocking the doors after workers temporarily exit 
during the working day does not minimize radiation dose and reduces 
worker efficiency. Access control will be established on the first 
floor of the G-Building outside the radiation area. Therefore, the 
access control point would provide positive control into and out of 
containment and meets the original intent of the TS.
    The fourth TS is being changed to include the HP control point in 
the monthly visual surveillance, which assures that accesses into 
containment are locked or blocked when no on is inside containment and 
the HP control point is not occupied.
    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 
considerations. The proposed amendment to a license of a facility 
involves no significant hazards consideration if 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 the margin of safety.
    The staff agrees with the licensee's no significant hazards 
consideration determination submitted on September 15, 1999, for the 
following reason:
    The changes are consistent with the original intent of the TS, 
i.e., to maintain confinement during Restricted Activities and to 
prevent uncontrolled spread of contamination. Access control is still 
being maintained.
    Based on a review of the licensee's analysis, and on the staff's 
analysis detailed above, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: William David Wall, Assistant General 
Counsel, CBS Corporation, 11 Stanwix Street, Pittsburgh, Pennsylvania 
15222.
    NRC Branch Chief: Ledyard B. Marsh.

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

    Date of amendment request: June 2, 1999, as supplemented August 25, 
1999.
    Description of amendment request: The proposed amendment would 
relocate the quality assurance (QA) related requirements to the 
licensee's Quality Assurance Program Description (QAPD) in accordance 
with NRC Administrative Letter (AL) 95-06, ``Relocation of Technical 
Specifications Administrative Controls Related to Quality Assurance,'' 
dated December 12, 1995. Specifically, Technical Specification (TS) 
Section 6.5, ``Review and Audit,'' TS Section 6.8, ``Procedures and 
Programs,'' and TS Section 6.10, ``Record Retention'' would be 
relocated from the current TS to the QAPD in accordance with 10 CFR 
50.36 (60 FR 30957).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously analyzed?
    Response: This amendment application does not involve a 
significant increase in the probability or consequences of an 
accident previously analyzed. The relocation of the administrative 
controls from the Technical Specification to the Quality Assurance 
Program Description (QAPD) does not alter the performance or 
frequency of these activities. Any future changes to the QA Program 
Description, which might constitute a reduction in commitments, are 
governed by 10 CFR 50.54(a). Therefore, sufficient controls for 
these requirements exist and these changes do not involve a 
significant increase in the probability or consequences of an 
accident previously analyzed.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    Response: This amendment application does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. The proposed changes involve the relocation of 
requirements from the Technical Specifications to the QAPD.

[[Page 59800]]

Relocation of these requirements does not affect plant equipment or 
the way the plant operates. The functions continue to be performed 
in the identical manner as they are currently being performed. 
Therefore, the proposed revisions can not create a new or different 
kind of accident.
    3. Does the proposed license amendment involve a significant 
reduction in a margin of safety?
    Response: This amendment application does not involve a 
significant reduction in a margin of safety. The requested Technical 
Specification revisions relocate the administrative control 
requirements from the Technical Specifications to the QAPD. These 
requirements are not being altered by this relocation. The functions 
continue to be performed in the identical manned as they are 
currently being performed. Any future changes to the QA Program 
Description, which might constitute a reduction in commitments, are 
governed by 10 CFR 50.54(a). Therefore, sufficient controls for 
these requirements exist and these 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Section Chief: Sheri Peterson.

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

    Date of amendment request: July 30, 1999 (NRC-99-0048).
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to include provisions related 
to enabling the oscillation power range monitor (OPRM) upscale trip 
function in the average power range monitor. This change is associated 
with the power range neutron monitoring (PRNM) system installed during 
the last refueling outage. The associated Bases would also be revised.
    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 is to enable the OPRM Upscale Function that 
is contained in the previously installed PRNM equipment. Enabling 
the OPRM hardware provides the long-term stability solution required 
by Generic Letter 94-02. This hardware incorporates the Option III 
detect and suppress solution reviewed and approved by the NRC in the 
Reference 6, 7, and 8 [of the licensee's application dated July 30, 
1999] Licensing Topical Reports and their Supplements. The OPRM is 
designed to meet all requirements of GDC [General Design Criteria] 
10 and 12 by automatically detecting and suppressing design basis 
thermal-hydraulic power oscillations prior to violating the fuel 
MCPR [minimum critical power ratio] Safety Limit. The OPRM system 
provides this protection in the region where Interim Corrective 
Actions (ICAs) restricted operation because of stability concerns. 
Thus, the ICA restrictions on plant operation are deleted from the 
TS, including region avoidance and the requirement for the operator 
to manually scram the reactor with no recirculation loops operating. 
Operation at high core powers with low core flows may cause a 
slight, but not significant, increase in the probability that an 
instability may occur. This slight increase is acceptable because 
subsequent to the automatic detection of an instability, the OPRM 
Upscale function provides an automatic scram signal to the RPS that 
is faster than the operator-initiated manual scram required by the 
current ICAs. Because of this rapid automatic action, the 
consequences of an instability event are not increased as a result 
of the installation of the OPRM system because it eliminates 
dependence on operator actions.
    Based on the above discussion, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change permits Fermi 2 to enable the OPRM power 
oscillation detect and suppress function provided in previously 
installed PRNM hardware, and it simultaneously deletes certain 
restrictions which preclude operation in regions of the power-flow 
map where oscillations potentially may occur. Enabling the OPRM 
Upscale function does not create any new system hardware interfaces 
nor create any new system interactions. Potential failures of the 
OPRM Upscale function result either in failure to perform a 
mitigation action or in spurious initiation of a reactor scram. 
These failures would not create the possibility of a new or 
different kind of accident.
    Based on the above discussion, the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The change does not involve a significant reduction in the 
margin of safety.
    The OPRM Upscale function implements BWROG [Boiling Water 
Reactor Owners Group] Stability Option III, which was developed to 
meet the requirements of GDC 10 and GDC 12 by providing a hardware 
system that detects the presence of thermal-hydraulic instabilities 
and automatically initiates the necessary actions to suppress the 
oscillations prior to violating the MCPR Safety Limit. The NRC has 
reviewed and accepted the Option III methodology described in the 
Reference 6, 7, and 8 [of the licensee's application dated July 30, 
1999] Licensing Topical Reports and their supplements, and concluded 
that this solution will provide the intended protection. Therefore, 
it is concluded that there will be no reduction in the margin of 
safety as defined in the TS as a result of enabling the OPRM Upscale 
function and simultaneously removing the operating restrictions 
previously imposed by the ICAs.
    Based on the above discussion, the proposed change does not 
involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: September 10, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Surveillance Requirements (SRs) 
3.8.4.1, 3.8.4.6, and 3.8.6.2 to accommodate changes in battery 
parameters associated with the replacement of the Division I battery. 
The licensee also plans to revise the Bases section for SR 3.8.6.2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes do not involve a change in the manner in 
which the plant is operated. TS Sections [SRs] 3.8.4.1, 3.8.4.6, 
3.8.6.2 and Bases Surveillance Requirement Section 3.8.6.2 are being 
revised to reflect the new Division I battery cell/system 
characteristics and associated requirements. The new battery will 
have an increased capacity over the present battery, while 
maintaining the existing battery system voltage requirements. This 
is possible because the present and new battery specific gravity 
(1.215) and type (lead calcium) are the same. Also, the end of 
battery system discharge voltage remains the same as 210 VDC. The 
Division I batteries will continue

[[Page 59801]]

to furnish power to redundant essential loads as required and as 
designed. The new surveillance requirement voltages are based on the 
same volts/cell criteria used for the existing batteries. 
Furthermore, failure or malfunction of the station batteries does 
not initiate any of the analyzed accidents previously evaluated in 
the UFSAR [updated final safety analysis report]. The changes 
described will therefore not involve an increase in the probability 
or consequences of an accident previously evaluated.
    2. The changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The new battery is Class 1E qualified equipment and is being 
maintained within the same overall design parameters as the existing 
battery. That is, the battery terminal voltage on float voltage 
conditions (2.167 volt[s]/cell), overvoltage conditions (2.5 volts/
cell) and charger capability (2.15 volts/cell) are the same as the 
original design. Furthermore, the end of system discharge voltage of 
the battery system is maintained the same; therefore, there is no 
negative impact to plant loads supplied by the batteries. Failures 
of the batteries and chargers have been considered in both the 
existing and modified configurations. The proposed changes will not 
change performance or reliability nor introduce any new or different 
failure modes or common mode failure and will therefore not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The changes do not involve a significant reduction in the margin 
of safety.
    The changes act to increase overall battery capacity from 560 
ampere-hours to 1200 ampere-hours with the minimum battery discharge 
voltage remaining at 210 VDC (or 105 VDC per battery). The battery 
terminal voltage on float voltage conditions (2.167 volt[s]/cell), 
overvoltage conditions (2.5 volts/cell) and charger capability (2.15 
volts/cell) are the same as the original design. The new surveillance 
requirement voltages are based on the same volts/cell criteria used for 
the existing batteries. The batteries' ability to satisfy the design 
requirements (battery duty cycle) of the dc system will not be reduced 
from original plant design and will therefore not have any negative 
impact to plant loads [that] the battery supplies. The proposed changes 
therefore do not involve a reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Section Chief: Claudia M. Craig.

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 amendment request: April 5, 1999; supplemented October 7, 
1999.
    Description of amendment request: The proposed amendments would 
revise the Improved Technical Specifications (TS), Updated Final Safety 
Analysis Report (UFSAR), and Core Operating Limits Report to 
incorporate Topical Report (TR) DPC-NE-3005-P, ``Thermal-Hydraulic 
Transient Analysis Methodology.'' The proposed changes are: (1) 
Modification of a note for TS Surveillance Requirement (SR) 3.4.1.2, 
``RCS [Reactor Coolant System] Pressure, Temperature, and Flow DNB 
[Departure from Nucleate Boiling] Limits,'' to add that the SR would 
apply for the condition where there is a 0 deg.F delta-Tcold setpoint; 
(2) modification of TS 3.4.10, ``Pressurizer Safety Valves,'' to 
increase the setpoint range of the lift settings for the pressurizer 
safety valves; (3) modification of SR 3.4.10.1 to specify that the 
pressurizer safety valve lift settings shall be within plus or minus 1 
percent; (4) addition of TS 3.7.4, ``Atmospheric Dump Valve (ADV) Flow 
Paths,'' to address the applicability and required actions related to 
the ADS valves; (5) addition of TS 3.9.7, ``Unborated Water Source 
Isolation Valves,'' to require valves that are used to isolate 
unborated water sources to be secured in the closed position while in 
Mode 6, provide required actions if one or more of the valves is not 
secured in the closed position, and related SRs; (6) TS 5.6.5b would be 
changed to update the Core Operating Limits Report references; and (7) 
modification of the appropriate Bases to reflect the above changes and 
consistentcy with the revision to the TR analysis. In addition, 
proposed changes to the UFSAR revisions were provided.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. The proposed changes to the Technical Specifications, Bases, 
Updated Final Safety Analysis Report (UFSAR), and Core Operating 
Limits Report (COLR) incorporate the accident analyses established 
in Topical Report DPC-NE-3005-P, ``UFSAR Chapter 15 Transient 
Analysis Methodology, Revision 1.'' On February 1, 1999, Duke 
submitted Topical Report DPC-NE-3005-P to the NRC for approval. The 
NRC found DPC-NE-3005-P acceptable as noted in SER [Safety 
Evaluation Report] dated May 25, 1999.
    The analyzed events are initiated by the failure of specific 
plant structures, systems or components. These proposed changes do 
not impact the condition or performance of those structures, systems 
or components.
    The revised accident analyses in DPC-NE-3005-P demonstrate that 
the applicable acceptance criteria are met. In addition, the 
calculations show that the applicable radiological and environmental 
acceptance criteria will continue to be met.
    Based on the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated?
    No. The proposed changes do not involve a physical alteration of 
the plant. No new or different equipment is being installed, and no 
installed equipment is being operated in a new or different manner. 
Where setpoints and operating limits have been revised, the revised 
accident analyses demonstrate that the applicable acceptance 
criteria are met. As a result, no new failure modes are being 
introduced.
    Based on the above, the proposed changes do not create the 
possibility of any new or different kind of accident from any 
accident previously evaluated.
    3. Involve a significant reduction in a margin of safety?
    No. The margin of safety is established through the design of 
the plant structures, systems and components, the parameters within 
which the plant is operated, and the establishment of the setpoints 
for the actuation of equipment relied upon to respond to an event. 
The proposed changes do not involve a physical alteration of the 
plant. No new or different equipment is being installed, and no 
installed equipment is being operated in a new or different manner. 
Where setpoints and operating limits have been revised, the revised 
accident analyses in DPC-NE-3005-P demonstrate that the applicable 
acceptance criteria are met.
    Based on the above, the proposed changes do not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC.
    NRC Section Chief: Richard L. Emch, Jr.

[[Page 59802]]

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

    Date of amendment request: September 9, 1999.
    Description of amendment request: The proposed amendment would 
increase the authorized rated thermal power level of 3579 megawatts 
thermal by 5 percent to 3758 megawatts thermal. The proposal follows 
the NRC-approved generic format and content for Boiling Water Reactor 
power uprate licensing topical reports documented in NEDC-31897P-A, 
``Generic Guidelines for General Electric Boiling Water Reactor Power 
Uprate,'' and NEDC-31984P, ``Generic Evaluations of General Electric 
Boiling Water Reactor Power Uprate.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    (1) Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The increase in power level discussed herein will not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    The probability (frequency of occurrence) of Design Basis 
Accidents occurring is not affected by the increased power level, as 
the regulatory criteria established for plant equipment (ASME code, 
IEEE standards, NEMA standards, Regulatory Guide criteria, etc.) are 
still complied with at the uprated power level. An evaluation of the 
boiling water reactor (BWR) probabilistic risk assessments concludes 
that the calculated core damage frequencies do not significantly 
change due to power uprate. Scram setpoints (equipment settings that 
initiate automatic plant shutdowns) are established such that there 
is no significant increase in scram frequency due to uprate. No new 
challenge to safety-related equipment results from power uprate.
    The changes in consequences of hypothetical accidents which 
would occur from 102% of the uprated power, compared to those 
previously evaluated from greater than or equal to 102% of the 
original power, are in all cases insignificant, because the accident 
evaluations from power uprate compared with 105% of original power 
do not result in exceeding the NRC-approved acceptance limits. The 
spectrum of hypothetical accidents and transients has been 
investigated, and shown to meet the plant's currently licensed 
regulatory criteria. In the area of core design, for example, the 
fuel operating limits such as Maximum Average Planar Linear Heat 
Generation Rate (MAPLHGR) and Safety Limit Minimum Critical Power 
Ratio (SLMCPR) are still met at the uprated power level, and fuel 
reload analyses will show plant transients meet the criteria 
accepted by the NRC as specified in NEDO-24011, ``GESTAR II.'' 
Challenges to fuel (ECCS performance) are evaluated, and shown to 
still meet the criteria of 10 CFR 50.46 and Appendix K (Section 4.3 
above, and Regulatory Guide 1.70 Safety Analysis Report Section 
6.3).
    Challenges to the containment have been evaluated, and the 
containment and its associated cooling systems will continue to meet 
10 CFR Appendix A Criterion 38, Long Term Cooling, and Criterion 50, 
Containment.
    Radiological release events (accidents) have been evaluated, and 
shown to meet the guidelines of 10 CFR 100 (Regulatory Guide 1.70 
Safety Analysis Report Chapter 15).
    (2) Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    As summarized below, this change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Equipment that could be affected by power uprate has been 
evaluated. No new operating mode, safety-related equipment lineup, 
accident scenario or equipment failure mode was identified. The full 
spectrum of accident considerations defined in Regulatory Guide 1.70 
has been evaluated and no new or different kind of accident has been 
identified. Power uprate uses existing technology, and applies it 
within the capabilities of already existing plant equipment in 
accordance with existing regulatory criteria and includes NRC 
approved codes, standards, and methods. General Electric has 
designed BWRs of higher power and no new power dependent accidents 
have been identified.
    The technical specifications needed to implement power uprate 
require some small adjustments, with no change to the plant's 
physical configuration. All technical specification changes have 
been evaluated and are acceptable.
    (3) Will the change involve a significant reduction in a margin 
of safety?
    As summarized below, this change will not involve a significant 
reduction in a margin of safety.
    The calculated loads on all affected structures, systems and 
components remain within their design allowables for all design 
basis event categories. No NRC acceptance criteria are exceeded. 
Some design and operational margins are affected by power uprate, 
however, the margins of safety originally designed into the plant 
are not affected by power uprate. Because the plant configuration 
and reactions to transients and hypothetical accidents do not exceed 
the presently approved NRC acceptance limits, power uprate does not 
involve a significant reduction in a margin of safety.

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

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

    Date of amendment request: September 9, 1999.
    Description of amendment request: The proposed amendment would 
revise Perry Operating License Appendix B, the Perry Environmental 
Protection Plan. The proposed change will eliminate the requirement in 
the Environmental Protection Plan to sample Lake Erie sediment in the 
Perry and Eastlake Plant area for Corbicula, since Corbicula and zebra 
mussels have already been identified, and control and treatment plans 
have been implemented which are effective on both species.
    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 Perry Plant water source (Lake Erie) is now known to have 
mussels and clams present. Therefore, it is no longer necessary to 
use lake sampling techniques designed to provide advance notice of 
their arrival. Treatment programs and monitoring for system fouling 
are in place. The treatment programs and system monitoring for 
fouling makes it highly likely that equipment degradation due to 
Corbicula would be avoided or readily identified, allowing time for 
corrective actions. Therefore, the programs will ensure that plant 
systems remain capable of performing their intended functions. Since 
the lake sampling was designed to allow time to implement a control 
program, and the control program is now in place, elimination of the 
lake sampling program will not involve a significant increase in the 
probability or radiological consequences of an accident previously 
evaluated.
    (2) The proposed change would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change will eliminate the lake sampling program 
designed to detect the arrival of Corbicula, a particular species of 
clam, at the Perry Plant. Since the clam is now known to exist in 
the vicinity, and control methods are developed and implemented, 
advanced detection is no longer required. Since the proposed change 
involves only a monitoring program and does not change or modify the 
design, maintenance or operation of any plant equipment, the 
proposed change would not create the possibility of a new or 
different

[[Page 59803]]

kind of accident from any accident previously evaluated.
    (3) The proposed change will not involve a significant reduction 
in the margin of safety.
    The current requirements for aquatic monitoring are designed to 
detect Corbicula prior to plant cooling water systems and heat 
exchangers becoming infested with clams and flow becoming degraded, 
and thus reducing the cooling available to safety systems.
    Since an effective control method has already been implemented, 
the deletion of a lake sampling method to provide advance warning of 
clams in the area provides no significant benefit. The proposed 
change will continue to provide the same level of protection against 
system or component fouling that currently exists, thus the proposed 
change will not involve a significant reduction in the margin of 
safety.

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

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

    Date of amendment request: September 9, 1999.
    Description of amendment request: The proposed amendment includes 
nine separate changes to the Perry technical specifications. The 
proposed changes include increasing the minimum water volume of the 
condensate storage tank, clarification of minimum ECCS pump 
differential pressures, clarifications to Required Action and Condition 
statements, as well as minor nomenclature and editorial changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    (1) The proposed changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    A summary of the proposed changes is:
    1. (Condensate Storage Tank (CST) Level-Low.) The Allowable 
Values for the CST low water level limits (Technical Specification 
(TS) Table 3.3.5.1-1 Function 3.d and Table 3.3.5.2-1 Function 3) 
are being revised from greater than or equal to 59,700 gallons to 
greater than or equal to 90,300 gallons based on recent revisions to 
calculations taking into account potential vortex issues. This 
change also results in raising the TS Surveillance Requirement (SR) 
3.5.2.2.b value for the normal CST level limit to greater than or 
equal to 249,700 gallons.
    2. (Emergency Core Cooling System Pump Differential Pressure) TS 
SRs 3.5.1.4 and SR 3.5.2.5 are being revised to better describe what 
the differential pressures listed in the SRs represent at Perry 
Nuclear Power Plant, in lieu of the phrase ``pump differential 
pressure'.
    3. (RCIC/RHR Steam Line Flow-High) The proposed change revises 
the nomenclature on a table to match the plant-specific instrument 
nomenclature.
    4. (Containment Average Temperature-To-Relative-Humidity) This 
revision is a clarification to prevent misinterpretation of the 
Required Actions.
    5. (Containment Vacuum Breakers) T 3.6.1.11 Required Action A.2 
is being revised to clarify the proper actions to take if the 
required number of vacuum breakers is not operable. Required Action 
A.2 is being revised to add the word ``required'.
    6. (Reporting Requirements) TS Administrative Controls Reporting 
Requirement 5.6.1 is being revised to clarify the definition of the 
time period of the report. ``Calendar'' is being removed from the 
term ``calendar year'' to clarify the time period that the 
Occupational Radiation Exposure Report is required to cover, to be 
consistent with the revised wording in 10 CFR 20.1003.
    7. (High Radiation Area) TS Administrative Control 5.7 is being 
revised to update the titles of individuals responsible for 
radiation protection. The term ``health physics'' is being revised 
to ``radiation protection'' to be consistent with plant terminology.
    8. (ECCS Instrumentation) Required Action E.1 Note 1 is being 
revised for consistency with other specifications. The word ``in'' 
is being added.
    9. (Electrical Power Systems) In TS 3.8.3, the word 
``continued'' is being added to the bottom of the page for 
consistency with other specifications.
    The CST level change is adjusted in a conservative direction, as 
recommended by NRC inspectors during a Safety System Functional 
Inspection (SSFI) that was conducted in the spring of 1997. The 
current setpoints were reviewed and determined to be adequate, 
however it was suggested that some additional margin should be 
added. The ``low level'' limits are being raised to move the 
setpoint further away from the level at which vortexing would begin, 
and the normal water level limit is also being raised to ensure that 
at least 150,000 gallons of water would be available for HPCS and 
RCIC. Since the existing limits are already considered adequate, and 
the proposed changes are in the conservative direction, the proposed 
change does not involve a significant increase in the probability or 
radiological consequences of an accident previously evaluated.
    The other eight proposed changes are administrative only, and 
can have no effect on any previously evaluated accident scenario. 
These eight changes have no effect on plant hardware, plant design, 
safety limit settings, or system operation and therefore do not 
modify or add any initiating parameters that would significantly 
increase the probability of an accident previously evaluated, or the 
radiological consequences of an event.
    (2) The proposed changes would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes will raise the Condensate Storage Tank 
level, which is conservative, and also includes some administrative 
changes to improve clarity, update titles or terminology. None of 
these changes can create the possibility of a new of different kind 
of accident from any accident previously evaluated.
    (3) The proposed changes will not involve a significant 
reduction in the margin of safety.
    The Condensate Storage Tank level change increases the margin of 
safety by providing more margin between the setpoint that causes the 
HPCS and RCIC suctions to shift from the CST to the Suppression Pool 
and the beginning of the formation of a vortex at their pump 
suctions. The other administrative changes have no effect on the 
margin of safety. Therefore the proposed change will not involve a 
significant reduction in the margin of safety.

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

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

    Date of amendment request: September 14, 1999.
    Description of amendment request: The proposed amendment would 
delete one Operating License Condition, and revise another. License 
Condition 2.C.10 regarding controls over the containment air locks 
during plant outages would be deleted due to the effective 
implementation of Shutdown Safety administrative controls at Perry. 
License Condition 2.F would be revised to clarify the intent of 
reporting requirements for violations of the technical specifications 
and the Environmental Protection Plan.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 59804]]

issue of no significant hazards consideration which is presented below:

    (1) The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes delete or revise two Operating License 
Conditions, one that addresses administrative controls on air locks 
during refueling outages, and one regarding reporting of violations 
of the technical specifications and the Environmental Protection 
Plan.
    These proposed changes to the Operating License are 
administrative only, and have no effect on any previously evaluated 
accident scenario. The proposed changes have no effect on plant 
hardware, plant design, safety limit setting, or plant system 
operation and therefore do not modify or add any initiating 
parameters that would significantly increase the probability of an 
accident previously evaluated.
    The changes will not alter the operation of equipment assumed to 
be available for the mitigation of accidents or transients, nor will 
they alter the operation of equipment important to safety previously 
evaluated in the accident analyses.
    The proposed activity does not affect accident mitigation 
capabilities or the radiation release amounts for postulated 
accidents. Since there are no changes to previous accident analyses, 
the radiological consequences associated with these analyses remain 
unchanged.
    Therefore, the proposed change does not significantly increase 
the probability or consequences of an accident previously evaluated.
    (2) The proposed change would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes are administrative in nature, and do not 
involve any physical alteration of the plant (no new or different 
type of equipment will be installed). They do not alter the design 
assumptions, conditions, configuration of the facility or the manner 
in which the plant is operated. The proposed changes have no impact 
on component and system interactions.
    The safety functions of plant structures, systems, and 
components are also not changed in any manner, nor is the 
reliability of any structure, system, or component reduced.
    The proposed changes are not providing for operation in a mode 
that is not already evaluated. These changes do not affect the 
operation of any systems or components, nor do they involve any 
potential initiating events that would create any new or different 
kind of event.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    (3) The proposed change will not involve a significant reduction 
in the margin of safety.
    The proposed changes are administrative in nature (they delete 
or revise two license conditions). Administrative controls will 
continue to be applied to the opening of the air locks during plant 
shutdown periods, and to the reporting of violations of the 
technical specifications and the Environmental Protection Plan.
    There is no impact on safety limits or limiting safety system 
settings. The changes do not affect any plant safety parameters or 
setpoints. No physical or operational changes to the facility will 
result from the proposed changes.
    The proposed changes have no impact on any safety analysis 
assumptions. Consequently, no margin of safety as described in the 
Final Safety Analysis Report or defined in the basis of any 
technical specification is reduced as a result of these changes. 
These proposed changes do not detrimentally affect the ability of 
structures, systems, and components important to safety to fulfill 
their intended safety functions.
    Therefore, the proposed changes do not cause a significant 
reduction in the margin of safety.

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

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: October 12, 1999.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification (T/S) Surveillance Requirement (SR) 
4.6.2.2.d for the spray additive system to relocate the details 
associated with the acceptance criteria and test parameters to the 
associated T/S Bases. Additionally, certain administrative text format 
changes are being proposed.
    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 relocate the details associated with the 
acceptance criteria and test parameters from the T/S SR to the 
associated Bases and do not affect system operability or 
performance. The format changes in the text on each page are 
administrative in nature and do not result in any change in plant 
operation. Relocation of this information to the Bases is 
administrative in nature and does not affect the probability or 
consequences of any accident previously evaluated. No actual change 
to the requirement is made. Actual plant operation is not affected 
by the administrative changes. No methods of operation of plant 
systems, structures or components are changed. Operation of accident 
mitigation features is not changed. Consequently, there is no affect 
upon the probability of any previously analyzed accident, transient, 
accident initiators, or precursor events. Additionally, because 
there is no actual change in plant design or operation, there is no 
affect upon radioactive material inventories, plant shielding, or 
effluent release points. Therefore, these changes do not 
significantly increase the probability of occurrence or consequences 
of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes relocate the details associated with the 
acceptance criteria and test parameters from the 
T/S SR to the associated Bases and do not affect system operability 
or performance. The format changes in the text on each page are 
administrative in nature and do not result in any change in plant 
operation. Facility operation and procedures are not changed. 
Relocation of this information to the Bases is administrative in 
nature and does not affect [sic] create any new accident scenarios, 
accident initiators, or precursor events. 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 proposed changes relocate the details associated with 
acceptance criteria and test parameters from the T/S SR to the 
associated Bases and do not modify T/S safety settings, setpoints, 
or other values. The format changes in the text on each page are 
administrative in nature and do not result in any change in plant 
operation. There is no effect upon operating margins and accident 
margins because the administrative changes do not change the manner 
of operation of plant systems, structures, or components. Plant 
emergency and abnormal operating procedures are not affected. There 
is no change of actual testing methodology, test parameters, or 
acceptance criteria. The response of the plant to an event is the 
same. Potential offsite doses are unaffected because operation of 
the facility is unchanged. Relocation of the testing details to the 
Bases is acceptable because controls are in place for T/S Bases 
changes which require evaluation of changes under the provisions of 
10 CFR 50.59. 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

[[Page 59805]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Jeremy J. Euto, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: September 30, 1999.
    Description of amendment request: The proposed amendment would 
change the Technical Specification surveillance periodicity 
requirements for the control room emergency filtration system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    During an accident, the Control Room Emergency Filtration [EFT] 
System provides filtered air to pressurize the Control Room to 
minimize the activity, and therefore the radiological dose, inside 
the Control Room. Technical Specification surveillance requirements 
are established in order to ensure that the EFT System will perform 
its safety function during an accident. The proposed amendment 
eliminates unnecessary testing which is not required to show that 
the filters are operable and which causes unnecessary wear and tear 
on the system. The remaining surveillances adequately show that the 
system is operable and capable of performing its safety function. 
Dose to the public and the Control Room operators are not affected 
by the proposed change.
    The proposed Technical Specification change does not introduce 
new equipment operating modes, nor does the proposed change alter 
existing system relationships. The proposed amendment does not 
introduce new failure modes.
    Therefore, the proposed amendment will not significantly 
increase the probability or the consequences of an accident 
previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed Technical Specification change does not introduce 
new equipment operating modes, nor does the proposed change alter 
existing system relationships. The proposed amendment does not 
introduce new failure modes. The proposed surveillance requirements 
are consistent with industry and regulatory guidance and show that 
the system is capable of performing its safety function. System 
reliability is enhanced by the proposed change by eliminating 
unnecessary wear on the system.
    Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment will not involve a significant reduction 
in the margin of safety.
    The proposed amendment is within current industry and regulatory 
standards for testing filters. The proposed amendment maintains 
margins of safety. Off-site and Control Room dose assessments are 
not affected by the proposed amendment, since the ability of the EFT 
System to perform its safety function is shown by the proposed 
surveillance requirements. The proposed change to the surveillance 
provides assurance that the system will perform at the filter 
efficiency used in the evaluation of the radiological consequences 
of the postulated events. Therefore, the proposed amendment will not 
involve a significant reduction in the margin of safety.

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

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

    Date of amendment request: September 30, 1999.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications associated with the Safety Limit 
Minimum Critical Power Ratios (SLMCPRs) in order to support the 
operation of Hope Creek in the upcoming Cycle 10 with a mixed core of 
General Electric (GE) and Asea Brown Bovieri/Combustion Engineering 
(ABB/CE) fuel. In addition, administrative changes would be made to the 
Technical Specifications to reflect the change in fuel vendor from GE 
to ABB/CE.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The derivation of the revised SLMCPRs for Hope Creek for 
incorporation into the Technical Specifications, and its use to 
determine cycle-specific thermal limits, have been performed using 
NRC [U.S. Nuclear Regulatory Commission] approved methods. These 
calculations do not change the method of operating the plant and 
have no effect on the probability of an accident initiating event or 
transient.
    There are no significant increases in the consequences of an 
accident previously evaluated. The basis of the MCPR Safety Limit is 
to ensure that no mechanistic fuel damage due to clad overheating is 
calculated to occur if the limit is not violated. The new SLMCPRs 
preserve the existing margin to transition boiling and the 
probability of fuel damage is not increased.
    Removal of the cycle specific footnote for the Safety Limit 
applicability will not involve a significant increase in the 
probability or consequences of an accident previously evaluated 
since the change is administrative and does not affect the plant or 
fuel design or operation.
    Likewise, the proposed changes to the Average Planar Heat 
Generation Rate (APLHGR), Minimum Critical Power Ratio (MCPR), 
Recirculation Loop Limiting Condition for Operation (LCO) Action 
Statements, and references to fuel vendor analyses and reports do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated. The changes to the 
APLHGR, MCPR and Recirculation Loop LCOs are considered to be 
administrative in nature since the Core Operating Limits Report 
(COLR) will continue to be used to appropriately control and limit 
the bounds of plant operation with slow control rods or during 
single recirculation loop operation, and the COLR will still be 
developed in accordance with NRC approved methods. Similarly, the 
revised references to the fuel vendor throughout the Technical 
Specifications are also considered to be administrative in nature 
since they reflect the current status of NRC approval of 
methodologies utilized by PSE&G [Public Service Electric and Gas 
Company] and the fuel vendor to develop operating and safety limits 
for the fuel and core designs. These proposed changes do not alter 
the method of operating the plant and have no effect on the 
probability of an accident initiating event or transient.
    There are no significant increases in the consequences of an 
accident previously evaluated. The basis of the COLR and the PSE&G 
and fuel vendor methodologies is to ensure that no mechanistic fuel 
damage is calculated to occur if the limits on plant operation are 
not violated. The COLR will continue to preserve the existing margin 
to fuel damage and the probability of fuel damage is not increased.
    Therefore, the proposed change does not involve an 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.

[[Page 59806]]

    The proposed changes contained in this submittal result from an 
analysis of the reload core using the same fuel types as previous 
cycles and an ABB/CE fuel design with extensive operating 
experience. These changes do not involve any new method for 
operating the facility and do not involve any facility modifications 
for the reload core operation. No new initiating events or 
transients result from these changes. Therefore, the proposed 
Technical Specification changes do not create the possibility of a 
new or different kind of accident, from any accident previously 
evaluated.
    Removal of the cycle specific footnote for the Safety Limit 
applicability does not create the possibility of a new or different 
kind of accident from any accident previously evaluated since the 
change is administrative and does not affect the plant or fuel 
design or operation.
    The changes to the APLHGR, MCPR and Recirculation Loop LCOs are 
considered to be administrative in nature since the Core Operating 
Limits Report (COLR) will continue to be used to appropriately 
control and limit the bounds of plant operation with slow control 
rods or during single recirculation loop operation, and the COLR 
will still be developed in accordance with NRC approved methods. 
These changes do not involve any new method for operating the 
facility and do not involve any facility modifications in addition 
to the new fuel design. No new initiating events or transients 
result from these changes. Therefore, the proposed Technical 
Specification changes do not create the possibility of a new or 
different kind of accident.
    The revised references to the fuel vendor throughout the 
Technical Specifications are also considered to be administrative in 
nature since they reflect the current status of NRC approval of 
methodologies utilized by PSE&G and the fuel vendor to develop 
operating and safety limits for the fuel and core designs. These 
changes do not involve any new method for operating the facility and 
do not involve any facility modifications in addition to the new 
fuel design. No new initiating events or transients result from 
these changes. Therefore, the proposed Technical Specification 
changes do not create the possibility of a new or different kind of 
accident.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety as defined in the Technical Specification 
bases will remain the same. The new SLMCPRs are calculated using NRC 
approved methods, which are in accordance with the current fuel 
designs, and licensing criteria. The MCPR Safety Limit remains high 
enough to ensure that greater than 99.9% of all fuel rods in the 
core will avoid transition boiling if the limit is not violated, 
thereby preserving the fuel cladding integrity. Therefore, the 
proposed Technical Specification changes do not involve a 
significant reduction in a margin of safety.
    Removal of the cycle specific footnote for the Safety Limit 
applicability does not create the possibility of a new or different 
kind of accident from any accident previously evaluated since the 
SLMCPR will continue to be evaluated on a cycle-specific basis.
    The margin of safety as defined in the Technical Specification 
bases will likewise remain unaffected by the proposed changes to 
APLHGR, MCPR and Recirculation Loop LCOs, and the revised references 
to the fuel vendor throughout the Technical Specifications. These 
changes establish controls for plant operation and establish bases 
for fuel analyses that reflect NRC approved methods, and are in 
accordance with the current fuel design and licensing criteria. 
These changes will continue to ensure that the plant is operated 
within specified acceptable fuel design limits. Therefore, the 
proposed Technical Specification changes do not involve a 
significant reduction in a margin of safety.

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

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

    Date of amendment request: September 8, 1999.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3/4.8.1, ``A.C. Sources, 
Operating,'' and associated Bases, by eliminating the requirement for 
accelerated testing of the standby diesel generators and the associated 
reporting requirements. The TS Index would also be revised to reflect 
these changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes do not involve hardware changes nor do they 
affect the operational limits or design of the standby diesel 
generators or power systems. These changes do not alter assumptions 
made in the safety analysis. In conjunction with the maintenance 
rule program, these changes continue to assure the operability and 
reliability of the standby diesel generators while minimizing the 
number of required engine starts and associated wear. These changes 
are also consistent with the guidance provided in Generic Letter 94-
01, ``Removal of Accelerated Testing and Special Reporting 
Requirements for Emergency Diesel Generators.''
    Therefore, the proposed changes do not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes minimize the number of required standby 
diesel generator starts; they do not affect the operational limits 
or design. The performance capability of the standby diesel 
generators is not affected. These changes do not alter the plant 
configuration (no new or different type of equipment will be 
installed) or make changes in methods governing normal plant 
operation. These changes do not alter assumptions made in the safety 
analysis. These changes are also consistent with the guidance 
provided in Generic Letter 94-01.
    Therefore, the changes will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes do not involve a change in the operational 
limits or design of the emergency power system. The design and 
capabilities of the standby diesel generators are not affected by 
these changes. These changes are also consistent with the guidance 
provided in Generic Letter 94-01.
    The proposed changes do not involve a significant reduction in 
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.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: September 8, 1999.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 3/4.8.1, ``A.C. Sources, Operating,'' 
and associated Bases, by relocating the 18-month surveillance to 
subject the standby diesel generator to inspections in accordance with 
procedures prepared in conjunction with its manufacturer's 
recommendations, to the Updated Final Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards

[[Page 59807]]

consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change moves the requirement to perform 
manufacturer's recommended inspections of the Standby Diesel 
Generators from the Technical Specifications to the Updated Final 
Safety Analysis Report (UFSAR). The change does not result in any 
hardware or operating procedure changes. The requirement being 
removed from the Technical Specifications is not the initiator of 
any analyzed event. The UFSAR is maintained using the provisions of 
10 CFR 50.59. Since any changes will be evaluated per 10 CFR 50.59, 
no significant increase in the probability or consequences of an 
accident previously evaluated will be allowed without prior NRC 
approval. Therefore, the changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change moves the requirement to perform 
manufacturer's recommended inspections of the Standby Diesel 
Generators from the Technical Specifications to the Updated Final 
Safety Analysis Report (UFSAR). The change does not alter the plant 
configuration (no new or different type of equipment will be 
installed) or make changes in methods governing normal plant 
operation. The change does not impose different requirements. The 
change does not alter assumptions made in the safety analysis and 
licensing basis. Therefore, the change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change moves the requirement to perform 
manufacturer's recommended inspections of the Standby Diesel 
Generators from the Technical Specifications to the Updated Final 
Safety Analysis Report (UFSAR). The change does not reduce the 
margin of safety since the location of details has no impact on any 
safety analysis assumptions. In addition, the requirement being 
transposed from the Technical Specification to the UFSAR [is the] 
same as the existing Technical Specification. Also, the UFSAR is 
maintained using the provisions of 10 CFR 50.59. Since any changes 
will be evaluated per 10 CFR 50.59, no significant reduction in a 
margin of safety will be allowed without prior NRC approval.

    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.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: September 28, 1999.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to increase the maximum allowable 
leakage rates for main steam isolation valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    TVA proposes to utilize the main steam drain lines to 
preferentially direct MSIV leakage to the main condenser. This drain 
path takes advantage of the large volume of the steam lines and 
condenser to provide holdup and plate-out of fission products that 
may leak through the closed MSIVs. In this approach, the main steam 
lines, steam drain piping, and the main condenser are used to 
mitigate the consequences of an accident to limit potential off-site 
exposures below those specified in 10 CFR 100 and 10 CFR 50 Appendix 
A, GDC 19 for control room dose limits.
    Seismic verification walkdowns and evaluations of representative 
piping/supports were performed to demonstrate the main steam line 
piping and components that comprise the ALT path were rugged, and 
able to perform the safety function of MSIV leakage control 
following an Design Basis Earthquake (DBE). Thus, it has been 
concluded the primary components in the MSIV alternate treatment 
flow path can be relied upon to maintain structural integrity.
    Therefore, the proposed amendment does not involve changes to 
structures, components, or systems which would affect the 
probability of an accident previously evaluated in the Browns Ferry 
Final Safety Analysis Report (FSAR).
    A plant-specific radiological analysis has been performed to 
assess the effects of the proposed increase in MSIV leakage criteria 
in terms of off-site doses and main control room dose. This analysis 
uses the holdup and plate-out factors described in NEDC-31858P, 
Revision 2. The analysis shows the dose contribution from the 
proposed increase in leakage criteria is acceptable compared to 
doses limits prescribed in 10 CFR 100 and 10 CFR 50, Appendix A, GDC 
19. Therefore, the proposed changes do not significantly increase 
the consequences of an accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes require the use of the main steam piping 
and the condenser to process MSIV leakage. This additional function 
does not compromise the reliability of these systems. They will 
continue to function as intended and not be subject to a failure of 
a different kind than previously considered. In addition, MSIV 
functionality will not be adversely impacted by the increased 
leakage limit. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change to TS Surveillance Requirement 3.6.1.3.10 to 
increase the allowable MSIV leakage does not involve a significant 
reduction in the margin of safety. The allowable leak rate specified 
for the MSIVs is used to quantify a maximum amount of leakage 
assumed to bypass containment. The results of the re-analysis 
supporting these changes were evaluated against the dose limits 
contained in 10 CFR 100 for off-site doses and 10 CFR 50, Appendix 
A, GDC 19 for control room doses. Sufficient margin relative to the 
regulatory limits is maintained even when conservative assumptions 
and methods are utilized. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
    NRC Section Chief: Sheri R. Peterson.

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

    Date of amendment request: September 30, 1999.
    Description of amendment request: The proposed amendments consist 
of administrative revisions to the Operating Licenses for BFN Units 1, 
2 and 3 that delete license conditions that have become outdated, are 
no longer applicable, or are redundant, and consolidate license 
conditions which currently exist in two locations in each units' 
Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards

[[Page 59808]]

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 changes requested by this submittal are administrative in 
nature and do not change the way BFN operates. The proposed changes 
are intended to: delete redundant paragraphs, delete requirements 
and authorizations for modifications that have been completed, 
delete an authorization to temporarily store radioactive material on 
site, delete an exemption from a General Design Criterion which has 
expired, and consolidate license conditions which currently exist in 
two locations in each units Technical Specifications.
    The change does not affect any design bases accident or the 
ability of any safe shutdown equipment to perform its design 
function. There are no physical modifications that are required to 
implement this license condition update. There is no impact on plant 
equipment or changes to operating procedures. 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.
    The changes described above are administrative in nature and do 
not change the way BFN operates. There are no physical modifications 
authorized by the proposed changes and there are no procedure or 
process changes that are requested. Changes requested are intended 
to ensure the license conditions reflect the current status of the 
plant. There is no impact on any accident analysis created by this 
change. Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The changes described above are administrative in nature and do 
not change the way BFN operates. There are no procedural or physical 
changes required by this amendment. The license conditions are being 
updated partially as a result of NRC Information Notice 97-43 which 
highlighted the importance of periodically verifying compliance with 
the Operating License. These changes are intended to delete license 
conditions which are no longer needed or are redundant in order to 
ensure the license conditions accurately reflect the current status 
of the licensed facility. The change does not affect any design 
bases accident or the ability of any safe shutdown equipment to 
perform its design function, therefore no margins of safety have 
been affected by any of these changes. Accordingly, the proposed 
amendment does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
    NRC Acting Section Chief: Ronald W. Hernan.

Previously Published Notices of Consideration of Issuance of 
Amendment 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 of the same as above. They were 
published as individual notices either because the 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 considerations.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, 
Units 1, 2, and 3, Maricopa County, Arizona

    Date of amendment requests: June 8, 1999.
    Brief description of amendments request: The proposed amendments 
would revise Technical Specification (TS) 3.7.15, ``Fuel Storage Pool 
Boron Concentration,'' TS 3.7.17, ``Spent Fuel Assembly Storage,'' and 
TS 4.3.1, ``Criticality,'' to increase spent fuel pool storage capacity 
by crediting soluble boron and decay time in the safety analysis for 
the spent fuel pool storage racks. The proposed amendments would also 
increase the maximum radially averaged fuel enrichment from 4.3 weight 
percent to 4.8 weight percent.
    Date of publication of individual notice in Federal Register: 
September 20, 1999 (64 FR 50835)
    Expiration date of individual notice: October 20, 1999.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, 
Units 1, 2, and 3, Maricopa County, Arizona

    Date of amendment requests: October 8, 1999.
    Brief description of amendments request: The proposed amendment 
would revise Technical Specification (TS) Section 3.8.4, ``DC Sources--
Operating,'' to waive, on a one-time basis, the requirement to perform 
Surveillance Requirement (SR) 3.8.4.8 for Unit 1 channels A, B, and C.
    Date of publication of individual notice in Federal Register: 
October 19, 1999 (64 FR 56369).
    Expiration date of individual notice: For comments on proposed no 
significant hazards consideration determination: November 2, 1999; for 
opportunity for hearing: November 18, 1999.

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

    Date of amendment request: October 20, 1998 (PCN 485), as 
supplemented August 13, 1999.
    Brief description of amendment request: The proposed amendments 
would revise the San Onofre Nuclear Generating Station Units 2 and 3 
technical specifications Surveillance Requirement 3.3.9 to include a 
response time testing requirement for the control room isolation 
signal.
    Date of publication of individual notice in Federal Register: 
October 12, 1999 (64 FR 55311.
    Expiration date of individual notice: November 12, 1999.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination,

[[Page 59809]]

and Opportunity for A Hearing in connection with these actions was 
published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

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

    Date of amendment request: October 27, 1998.
    Brief description of amendment: The amendments update the Operating 
Licenses for the Brunswick Steam Electric Plant, Units 1 and 2.
    Date of issuance: October 5, 1999.
    Effective date: October 5, 1999.
    Amendment No.: 206 and 236.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendment 
revises the Operating Licenses.
    Date of initial notice in Federal Register: December 30, 1998 (63 
FR 71964).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 5, 1999.
    No significant hazards consideration comments received: No.

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 2, 1999, as supplemented on 
September 1, 1999.
    Brief description of amendment: This amendment relocates Technical 
Specification (TS) Section 6.5, ``REVIEW AND AUDIT,'' TS 6.8.2, TS 
6.8.3, and TS Section 6.10, ``RECORD RETENTION,'' intact from the 
Harris Nuclear Plant (HNP) TS to the Quality Assurance Program 
Description (QAPD) currently located in HNP Final Safety Analysis 
Report Section 17.3. Future changes to the associated relocated TS will 
be processed in accordance with 10 CFR 50.54(a). The change is 
consistent with NUREG-1431, Revision 1, ``Standard Technical 
Specifications, Westinghouse Plants,'' dated April 1995, and with the 
guidance provided in NRC Administrative Letter 95-06, ``Relocation of 
Technical Specification Administrative Controls related To Quality 
Assurance,'' dated December 12, 1995.
    Date of issuance: October 19, 1999.
    Effective date: October 19, 1999.
    Amendment No.: 92.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35201).
    The September 1, 1999, submittal 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 October 19, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 30, 1999.
    Brief description of amendments: The amendments revised the 
requirements related to the cross-tie of DC power buses between units, 
remove references to the AT&T batteries which have been replaced at 
Braidwood Station, and remove references to the 10-day allowed outage 
time (AOT) required for replacement of the AT&T batteries at Braidwood, 
Unit 2, which was granted in Amendment Nos. 99 and 99 issued to 
Braidwood Station, Unit Nos. 1 and 2, on March 26, 1999.
    Date of issuance: October 13, 1999.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 111 and 104.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43767).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 13, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 9, 1998, and July 7, 
1999.
    Brief description of amendments: The amendments revised Technical 
Specification Table 3.3.3-2, ``Emergency Core Cooling System Actuation 
Instrumentation Setpoints,'' to modify the degraded voltage second 
level undervoltage relay setpoint and allowable value.
    Date of issuance: October 15, 1999.
    Effective date: Immediately, to be implemented prior to startup 
from L1R08 for Unit 1 and prior to startup from L2R08 for Unit 2.
    Amendment Nos.: 135 and 120.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 13, 1999 (64 FR 
2245) and August 11, 1999 (64 FR 43769).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 15, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 13, 1999, as 
supplemented on August 27, 1999.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) Section 1.0, ``Definitions,'' Item 1.7, ``Core 
Alteration,'' to specify that instrumentation and control rod movements 
are not considered core alterations if there are no fuel assemblies in 
the associated cell. The amendments also revise TS Sections 3/4.1, 3/
4.3, and 3/4.9 to reflect the change in definition. In addition, a 
license condition is added as follows: ``The licensee is prohibited 
from moving any fuel assemblies within the reactor pressure vessel 
unless all control rods except one are fully inserted during refueling 
in Mode 5''.
    Date of issuance: October 18, 1999.

[[Page 59810]]

    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 136 and 121.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: September 8, 1999 (64 
FR 48860).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 18, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application of amendment: June 3, 1999, and as supplemented 
by letter dated August 24, 1999 .
    Brief description of amendment: The amendment revises the Operating 
License to clarify that the license is not terminated until the 
Commission notifies the licensee in writing, and relocates certain 
Technical Specification (TS) requirements to licensee-controlled 
documents. The administrative controls section of the TSs have been 
revised to more closely conform to the standardized TSs. Administrative 
controls have been added for the control of radioactive effluents. A TS 
Bases Control Program has been added. The weight limit for loads 
carried over the spent fuel pool (SFP) has been increased. The 
amendment deletes certain TSs that are either (1) no longer applicable 
to the permanently shutdown and defueled state of the reactor, or (2) 
which duplicate regulatory requirements, or (3) which duplicate 
information located in the Updated Final Safety Analysis Report. A 
number of editorial changes were made to clarify the language used, to 
correct typographical errors, to renumber the listings, to remove 
section numbers that no longer contain requirements, and to renumber 
the pages in the TSs.
    Date of issuance: October 19, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 195.
    Facility Operating License No. DPR-61: The amendment revised the 
Operating License and the Technical Specifications.
    Date of original notice in Federal Register: July 14, 1999 (64 FR 
38024).
    The August 24, 1999, supplement contained clarifications of the 
June 3, 1999 amendment request. The supplemental information did not 
change the staff's initial proposed no significant hazards 
consideration determination nor expand the scope of the original 
notice. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 19, 1999.
    No significant hazards consideration received: No.

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

    Date of amendment request: September 24, 1999.
    Description of amendment request: The amendment revises current 
Technical Specification (TS) 3.6.1.8 by adding footnote ``**'' to 
Action b. The footnote allows continued operation of Fermi 2 with the 
leakage of penetration X-26 exceeding the limit in TS 4.6.1.8.2, 
provided certain compensatory measures are taken. Operation is allowed 
to continue until the next plant shutdown.
    Because the NRC staff issued the Fermi 2 improved standard TSs 
(ITS) on September 30, 1999, with implementation within 90 days, this 
amendment also provides pages that are compatible with the ITS. The 
amendment adds a new special operations TS, ITS 3.10.8, to address the 
compensatory actions and other requirements associated with penetration 
X-26.
    Date of issuance: October 19, 1999.
    Effective date: October 19, 1999, and shall be implemented within 5 
days.
    Amendment No.: 135.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes (64 FR 53421, dated October 1, 1999). The 
notice provided an opportunity to submit comments on the Commission's 
proposed NSHC determination. No comments have been received. The notice 
also provided for an opportunity to request a hearing by November 1, 
1999, but indicated that if the Commission makes a final NSHC 
determination, any such hearing would take place after issuance of the 
amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and final NSHC determination are contained in a 
Safety Evaluation dated October 19, 1999.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Section Chief: Claudia M. Craig.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, 
Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, 
Claiborne County, Mississippi

    Date of application for amendment: June 23, 1999, as supplemented 
by letters dated August 6, September 8, and October 4, 1999.
    Brief description of amendment: The amendment revises Technical 
Specification requirements for handling irradiated fuel in the 
Containment Building and in the Auxiliary Building, and selected 
specifications associated with performing core alterations.
    Date of issuance: October 20, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No: 139.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications and Operating License.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46435).
    The August 6, September 8, and October 4, 1999, submittals provided 
additional clarifying information and did not change the initial 
proposed no significant hazards consideration determination and did not 
expand the scope of the original application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 20, 1999.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, 
Turkey Point Plant, Units 3 and 4, Dade County, Florida

    Date of application for amendments: March 8, 1999.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS), Section 6.0, Administrative Controls, by 
removing requirements that are adequately controlled by existing 
regulations other than 10 CFR 50.36 and the TS. The amendments also 
relocate selected requirements from TS 6.0 to licensee-controlled 
documents or programs (e.g., the final safety analysis report or the 
quality assurance plan). Guidance on the changes was developed by the 
NRC and provided in the Standard Technical Specifications for 
Pressurized Water Reactor Plants, NUREG-1431, and Administrative Letter 
95-06, ``Relocation of Technical Specification

[[Page 59811]]

Administrative Controls Related to Quality Assurance,'' issued on 
December 12, 1995.
    Date of issuance: October 6, 1999.
    Effective date: As of date of issue, to be implemented within 90 
days of issuance.
    Amendment Nos.: 201 and 195.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the TS.
    Date of initial notice in Federal Register: April 7, 1999 (64 FR 
17025).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 6, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 10, 1999, as supplemented 
July 16 and October 4, 1999.
    Brief description of amendment: The amendment revised Duane Arnold 
Energy Center (DAEC) Technical Specification (TS) 2.1.1.2 to revise the 
Safety Limit Minimum Critical Power Ratio (SLMCPR) to support operation 
with GE-12 fuel with a 10x10 pin array.
    Date of issuance: October 20, 1999.
    Effective date: Immediately, to be implemented within 30 days
    Amendment No.: 229.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38029).
    The July 16 and October 4, 1999, letters provided additional 
clarifying information within the scope of the original Federal 
Register notice and did not affect the NRC staff's initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 20, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 14, 1998.
    Brief description of amendments: The amendments revise Technical 
Specification page 3/4 5-6, ``Limiting Conditions for Operation and 
Surveillance Requirements--Emergency Core Cooling Systems (ECCS),'' and 
its associated Bases to change pump runout limits for a safety 
injection pump to 675 gallons per minute (gpm) unless the pump is 
specifically tested to a higher flow rate not to exceed 700 gpm for 
Units 1 and 2.
    Date of issuance: October 21, 1999.
    Effective date: October 21, 1999, with full implementation within 
45 days.
    Amendment Nos.: 229 and 212.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1999 (64 FR 
47533).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 21, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 21, 1999.
    Brief description of amendments: The amendments change the 
Technical Specifications (TS) 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 TS 3/4.1.1, ``Boration 
Control,'' and TS 3/4.1.2, ``Boration Systems.'' Administrative changes 
are proposed to improve clarity. Finally, additions are made to 
shutdown margin TS 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 issuance: October 21, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 230 and 213.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 12, 1999 (64 FR 
37574).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 21, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 31, 1998, as 
supplemented May 17, 1999.
    Brief description of amendment: The amendment revises the technical 
specification reactor pressure vessel (RPV) pressure-temperature limit 
curves, deletes completed RPV sample surveillance requirements, deletes 
the requirement to withdraw a specimen at the next refueling outage, 
removes the standby liquid control system relief valve setpoint, and 
makes associated administrative changes.
    Date of issuance: October 12, 1999.
    Effective date: October 12, 1999, with full implementation within 
45 days.
    Amendment No.: 106.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.

    Date of initial notice in Federal Register: February 10, 1999 (64 
FR 6706). The May 17, 1999, submittal added clarifying information that 
was within the scope of the original Federal Register notice and did 
not change the staff's initial proposed no significant hazards 
considerations determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 12, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 9, 1999.
    Brief description of amendment: The amendment changes the Technical 
Specifications by increasing the allowable outage time for any one 
safety injection pump.
    Date of issuance: October 12, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 196.
    Facility Operating License No. DPR-64: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 2, 1999 (64 FR 
297147).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 12, 1999.
    No significant hazards consideration comments received: No.

[[Page 59812]]

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 29, 1999, as 
supplemented August 2, 1999.
    Brief description of amendment: The amendment changes the Technical 
Specifications by increasing the allowable control rod misalignment 
when operating at or below 85% power.
    Date of issuance: October 14, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 197.
    Facility Operating License No. DPR-64: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 21, 1999 (64 FR 
19564).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 14, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 19, 1999, as supplemented 
by letter dated October 8, 1999.
    Brief description of amendment: The amendment revises the TS to 
incorporate the new Pressure/Temperature Limits Curves consistent with 
the analysis results of reactor specimen W.
    Date of issuance: October 21, 1999.
    Effective date: October 21, 1999.
    Amendment No.: 143.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 8, 1999 (64 
FR 48865). The October 8, 1999, submittal 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 October 21, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 2, 1999 (TS 98-05).
    Brief description of amendments: The amendments delete the Sequoyah 
Nuclear Plant. License Conditions that require an Independent Safety 
Engineering Group.
    Date of issuance: October 12, 1999.
    Effective date: As of the date of issuance to be implemented no 
later than 45 days after issuance.
    Amendment Nos.: 248 and 239.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the License.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24201).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 12, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 18, 1999.
    Brief description of amendment: The amendment revises the 
definition of ``Surveillance Frequency'' to incorporate provisions that 
apply upon the discovery of a missed Technical Specification 
surveillance. This change allows a delay in performing the actions of 
the associated limiting conditions for operation for up to 24 hours or 
up to the limit of the specified frequency, whichever is less, when it 
is discovered that a surveillance was not performed within its 
specified frequency.
    Date of Issuance: October 13, 1999.
    Effective date: October 13, 1999, and shall be implemented within 
30 days.
    Amendment No.: 179.
    Facility Operating License No. DPR-28. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 9, 1999 (64 
FR 48867).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated October 13, 1999.
    No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
Creek Generating Station, Coffey County, Kansas

    Date of amendment request: July 8, 1999, as supplemented by letter 
dated September 2, 1999.
    Brief description of amendment: The amendment increased the 
allowable values for engineered safety features actuation system 
(ESFAS) loss-of-power 4 kV undervoltage trips in the current Technical 
Specifications (TSs) Table 3.3-4 (functional units 8.a and 8.b) and in 
surveillance requirement (SR) 3.3.5.3 of the improved TSs. The word 
``nominal'' is also added to describe the trip setpoint in SR 3.3.5.3 
and in the Bases of the improved TSs. The improved TSs were issued in 
Amendment 123 dated March 31, 1999, but have not yet been implemented.
    Date of issuance: October 12, 1999.
    Effective date: October 12, 1999, to be implemented within 60 days 
from the date of issuance.
    Amendment No.: 128.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43782).
    The September 2, 1999, supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noticed, and did not change the staff's initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 12, 1999.
    No significant hazards consideration comments received: No.

Yankee Atomic Electric Co., Docket No. 50-29, Yankee Nuclear Power 
Station (YNPS) Franklin County, Massachusetts

    Date of application for amendment: March 17, 1999
    Brief description of amendment: Revises the Possession Only License 
by deleting technical specifications related to hours of work and 
putting these requirements in appropriate Administrative Procedures.
    Date of issuance: October 8, 1999.
    Effective date: October 8, 1999, Implementation of this amendment 
includes incorporation of hours of work restrictions into the 
Administrative Procedures as described in the licensee's application 
dated March 17, 1999, and evaluated in the staff's safety evaluation 
attached to the amendment, and written notification to NRC that the 
amendment has been fully implemented.
    Amendment No.: 153.
    Facility Operating License No. DPR-3. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 7, 1999 (64 FR 
17032).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 8, 1999.
    No significant hazards consideration comments received: No.


[[Page 59813]]


    Dated at Rockville, Maryland, this 27th day of October 1999.

    For the Nuclear Regulatory Commission.
Suzanne C. Black,
Deputy Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 99-28598 Filed 11-2-99; 8:45 am]
BILLING CODE 7590-01-P