[Federal Register Volume 64, Number 125 (Wednesday, June 30, 1999)]
[Notices]
[Pages 35199-35221]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-16489]


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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 Pub. L. 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 June 5, 1999, through June 18, 1999. The 
last

[[Page 35200]]

biweekly notice was published on June 16, 1999 (64 FR 32284).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By July 30, 1999, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
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:

[[Page 35201]]

Rulemakings and Adjudications Staff, or may be delivered to the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington DC, by the above date. A copy of the petition should 
also be sent to the Office of the General Counsel, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and to the attorney 
for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendments request: May 26, 1999.
    Description of amendments request: The proposed amendment would 
revise Technical Specification 3.3.1, ``Reactor Protective System (RPS) 
Instrumentation--Operating,'' to change the RPS reactor coolant flow 
trip setpoints. The change is intended to reduce spurious reactor trip 
hazards.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed change will change the Reactor Protection 
System (RPS) reactor coolant flow trip setpoints. The RPS functions 
to mitigate the consequences of an accident. The changes to the low 
reactor coolant flow trip setpoints will reduce or eliminate 
unnecessary challenges to the RPS. Therefore, the proposed change 
will not involve a significant increase in the probability of an 
accident previously evaluated.
    These changes will result in an increased time delay for the RPS 
low reactor coolant flow trip. The reanalysis of the affected UFSAR 
[updated final safety analysis report] Chapter 15 events (UFSAR 
15.3.4, Reactor Coolant Pump Shaft Break with Loss of Offsite Power 
and UFSAR 15.1.5, Steam System Piping Failures Inside and Outside 
Containment--Modes 1 and 2 Operations), with the increased time 
delay, shows that the dose consequences for these events remain 
bounded by the UFSAR analysis. Therefore, this change does not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed change will change the RPS reactor coolant flow 
trip setpoints. The RPS functions to mitigate the consequences of an 
accident. The changes to the low reactor coolant flow trip setpoints 
will reduce or eliminate unnecessary challenges to the RPS. The 
proposed change only changes the mitigating actions of the RPS, 
without changing the required function of the RPS. Therefore, the 
change to the low reactor coolant flow trip setpoints does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed change will change the RPS reactor coolant flow 
trip setpoints. The reanalysis of the affected UFSAR Chapter 15 
events (UFSAR 15.3.4, Reactor Coolant Pump Shaft Break with Loss of 
Offsite Power and UFSAR 15.1.5, Steam System Piping Failures Inside 
and Outside Containment--Modes 1 and 2 Operations), with the revised 
reactor coolant flow trip setpoints, shows that the minimum DNBR 
[departure from nucleate boiling ratio] and SAFDLs [specified 
acceptable fuel design limits] for these events remain bounded by 
the UFSAR analysis. 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 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: June 2, 1999.
    Description of amendment request: The proposed amendment would 
relocate Shearon Harris Nuclear Power Plant (HNP) 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 HNP 
TS to the Quality Assurance Program Description currently located in 
the HNP Final Safety Analysis Report Section 17.3. Future changes to 
the associated relocated TS would be processed in accordance with 10 
CFR 50.54(a). The proposed 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.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This TS change relocates administrative requirements from HNP TS 
to the Quality Assurance Program Description (QAPD). The proposed 
amendment will not introduce any new equipment or require existing 
equipment to function different from that previously evaluated in 
the Final Safety Analysis Report (FSAR) or TS.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment will not introduce any new equipment or 
require existing equipment to function different from that 
previously evaluated in the Final Safety Analysis Report (FSAR) or 
TS. The changes are consistent with NUREG-1431, Revision 1 and the 
Commission's Final Policy Statement on Technical Specification 
improvements. The proposed amendment will not create any new 
accident scenarios, because the change does not introduce any new 
single failures, adverse equipment or material interactions, or 
release paths.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    This TS change relocates administrative requirements from HNP TS 
to the Quality

[[Page 35202]]

Assurance Program Description (QAPD). The QAPD will be revised to 
include the requirements associated with this proposed change. NRC 
Administrative Letter 95-06 states that administrative requirements 
for review and audit and the independent safety engineering group 
may be relocated from TS to the quality assurance program. HNP 
proposes relocating the associated requirements from TS to the QAPD 
intact. Future changes to these requirements will be processed in 
accordance with 10 CFR 50.54(a). This proposed TS change is 
administrative in nature and does not alter NRC acceptance limits 
with respect to accident mitigation or accident analysis.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Section Chief: Sheri R. Peterson.

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

    Date of amendment request: July 22 and October 22, 1998; May 6, 
1999.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TS) to reflect the licensee's planned use of 
fuel supplied by Westinghouse. The staff has published a Notice of 
Consideration of Issuance of Amendments and Proposed No Significant 
Hazards Consideration Determination on November 3, 1998 (63 FR 69338) 
covering the July 22 and October 22, 1998, submittals. In the May 6, 
1999, submittal the licensee proposed to expand the original amendment 
request, revising Section 5.6.5 of the Technical Specifications. 
Section 5.6.5 specifies a list of NRC-approved topical reports that the 
licensee is required to use to determine reactor core operating limits. 
The licensee proposed to update this list to show the current approval 
status of these topical reports.
    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 for the proposed changes conveyed by the May 6, 1999, 
submittal. The NRC staff has reviewed the licensee's analyses against 
the standards of 10 CFR 50.92(c). The NRC staff's analysis is presented 
below.

First Standard

    No. The proposed changes to Section 5.6.5 will not affect the 
safety function, and will not involve any change to the design or 
operation of any plant system or component. The topical reports were 
previously approved by the NRC staff under separate licensing actions. 
The use of methodologies in these approved topical reports will ensure 
that previously evaluated accidents remain bounding. Therefore, no 
accident probabilities or consequences will be impacted.

Second Standard

    No. The proposed changes would not lead to any hardware or 
operating procedure change. Hence no new equipment failure modes or 
accidents from those previously evaluated will be created.

Third Standard

    No. Margin of safety is associated with confidence in the design 
and operation of the plant; specifically, the ability of the fission 
product barriers to perform their design functions during and following 
an accident. The proposed changes to Section 5.6.5 do not involve any 
change to plant design, operation, or analysis. Thus the margin of 
safety previously analyzed and evaluated is maintained.
    Based on this analysis, it appears that the three standards of 10 
CFR 50.92(c) are satisfied for the proposed changes to Section 5.6.5. 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, Charlotte, North Carolina.
    Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina
    NRC Section Chief: Richard L. Emch, Jr.

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 May 27, 
1999.
    Description of amendment request: The proposed amendments would 
revise the Improved Technical Specifications (TS), Updated Final Safety 
Analysis Report, and Core Operating Limits Report to incorporate 
Topical Report (TR) DPC-NE-3005-P, ``Thermal-Hydraulic Transient 
Analysis Methodology.'' This analysis has been completed for Unit 2 and 
is ongoing for Units 1 and 3. Therefore, the proposed changes that 
reflect the TR provisions affect Unit 2 only. Other proposed changes 
affect all three units. Specifically, (1) a note to TS Surveillance 
Requirement (SR) 3.4.1.2, ``RCS [Reactor Coolant System] Pressure, 
Temperature, and Flow DNB [Departure from Nucleate Boiling] Limits,'' 
would be modified to address application of the delta-Tcold 
limits; (2) TS 3.4.10, ``Pressurizer Safety Valves,'' would be modified 
to increase the setpoint range of the lift settings for the pressurizer 
safety valves for the Oconee unit that has been analyzed in accordance 
with the TR and state that the range is not changed for the other 
units; (3) a statement to SR 3.4.10.1 would be added that will specify 
the pressurizer safety valve lift setpoint in order to clarify the 
difference between the operability setpoint range for a test lift and 
the range required when the setpoint is reset following the 
surveillance test; (4) TS 3.7.4, ``Atmospheric Dump Valve (ADV) Flow 
Paths,'' would be added to address the applicability and required 
actions related to the ADS valves; (5) TS 3.9.7, ``Unborated Water 
Source Isolation Valves,'' would be added to require valves that are 
used to isolate unborated water sources to be secured in the closed 
position while in Mode 6, incorporate SRs, and provide required actions 
if one or more of the valves is not secured in the closed position; (6) 
TS 5.6.5b would be changed to update the Core Operating Limits Report 
references; and (7) the appropriate Bases would be changed to reflect 
the above changes, other changes consistent with the revisions to the 
TR analysis, and the Updated Final Safety Analysis Report revisions 
that were provided in the submittal.
    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

[[Page 35203]]

Report DPC-NE-3005-P, ``UFSAR Chapter 15 Transient Analysis 
Methodology.'' On July 30, 1997, Duke submitted Topical Report DPC-
NE-3005-P to the NRC for approval. The NRC found DPC-NE-3005-P 
acceptable, with noted exceptions, in a Safety Evaluation issued on 
October 1, 1998. To resolve the noted NRC exceptions, Duke submitted 
Revision 1 of DPC-NE-3005-P to the NRC for review on February 1, 
1999. Additional information regarding Revision 1 of DPC-NE-3005-P 
was submitted on April 19 and May 5, 1999. This LAR is dependent 
upon the NRC approval of Revision 1 of DPC-NE-3005-P. [This Topical 
Report was approved by the NRC on 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 
preliminary calculations show that the applicable radiological and 
environmental acceptance criteria 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 a 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.
    Based upon the preceding evaluation, performed pursuant to 10 
CFR 50.92, Duke has concluded that the proposed changes to the 
Oconee Nuclear Station Technical Specifications, Bases, UFSAR, and 
O2C18 COLR will not involve a significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Conee County Library, 501 West 
South Broad Street, Walhalla, South Carolina
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC.
    NRC Section Chief: Richard L. Emch, Jr.

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: May 24, 1999
    Description of amendment request: The proposed amendments would 
revise the maximum local fuel pin centerline temperature safety limit 
in Technical Specification 2.1.1.1 from the limit determined using the 
TACO2 fuel performance computer code to the value determined using a 
newer TACO3 computer code.
    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.

    The following discussion is a summary of the evaluation of the 
changes contained in this proposed amendment against the 10 CFR 
50.92 (c) requirements to demonstrate that all three standards for 
no significant hazards consideration are satisfied. A no significant 
hazards consideration is indicated 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 a margin of safety.

First Standard

    Implementation of this amendment would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The use of the revised maximum local fuel pin 
centerline temperature limit is appropriate since the new limit uses 
a fuel melt temperature which has been conservatively reduced to 
account for code uncertainties in calculating fuel centerline 
temperature. NRC has previously found the use of the TACO3 code by 
DPC [Duke Power Company] in performing reload licensing to be 
acceptable. The use of the revised limit for fuel analyzed using an 
approved code ensures centerline fuel melting is avoided by ensuring 
the maximum fuel temperature is less than the melting temperature of 
the fuel. Therefore this change would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Second Standard

    Implementation of this amendment will not create the possibility 
of a new or different kind of accident from any previously 
evaluated. The use of the revised maximum local fuel pin centerline 
temperature limit has no affect on accident precursors. 
Implementation of this amendment will not impact any plant systems 
that are accident initiators. No other modifications are being 
proposed in the plant that would result in the creation of a new 
accident mechanism. Also, no changes are being made to the way the 
plant is operated; therefore, no new failure mechanisms will be 
initiated.

Third Standard

    The revised maximum local fuel pin centerline temperature limit 
has been appropriately reduced to account for uncertainties in 
predicting centerline fuel temperatures. NRC has previously found 
the use of the TACO3 code by DPC in performing reload licensing to 
be acceptable. Therefore, implementation of this amendment would not 
involve a significant reduction in a margin of safety.
    Therefore, Duke has concluded that the proposed amendment does 
not involve a significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC.
    NRC Section Chief: Richard L. Emch, Jr.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver 
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

    Date of amendment request: May 27, 1999.
    Description of amendment request: The proposed changes would 
relocate the seismic monitoring instrumentation requirements contained 
in Technical Specification (TS) 3/4.3.3.3 to the Licensing Requirements 
Manual based on the guidance provided in Generic Letter 95-10, 
``Relocation of Selected Technical Specifications Requirements

[[Page 35204]]

Related to Instrumentation.'' The Bases section for Specification 3/
4.3.3.3 will also be relocated to the LRM. The appropriate Index pages, 
Table Index page (Unit No. 1 only), TS pages and Bases pages will be 
revised to reflect the removal of the seismic monitoring 
instrumentation specification from the TSs. An additional specification 
page will be added to reflect that Specification Number 3/4.3.3.4 is 
not used. This additional page will also denote the number of the 
following page. The Bases section will also be modified to denote that 
Specification Number 3/4.3.3.4 is not used.
    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 amendment would relocate Technical Specification 
(TS) 3/4.3.3.3 titled ``Seismic Instrumentation'' and the associated 
Bases section to the Licensing Requirements Manual (LRM) (based on 
the guidance provided in Generic Letter (GL) 95-10, ``Relocation of 
Selected Technical Specification Requirements Related to 
Instrumentation''). The proposed amendment would also revise the TS 
Index and Beaver Valley Power Station (BVPS) Unit No. 1 List of 
Tables to reflect the relocation of this TS and associated Bases. 
The relocated Specification will be controlled in accordance with 
the requirement of 10 CFR 50.59, ``Controls, Tests, and 
Experiments.'' Additional administrative changes are also included 
to reflect that Specification Number 3/4.3.3.4 is not used.
    The proposed amendment does not involve a significant increase 
in the probability of an accident previously evaluated because no 
changes are being made to any accident initiator. No analyzed 
accident scenario is being changed. The initiating condition and 
assumptions remain as previously analyzed. The failure of the 
seismic monitoring instrumentation to detect a seismic event is not 
an accident initiating event.
    The seismic monitoring instrumentation performs no role in 
mitigating a seismic event or in achieving a safe shutdown condition 
after a seismic event has occurred. Seismic instrumentation is not 
assumed to function in the safety analysis. The seismic 
instrumentation is not associated with a process variable, design 
feature, or operating restriction that is an initial condition of a 
Design Basis Accident (DBA) or transient that either assumes the 
failure of or presents a challenge to the integrity of a fission 
product barrier. Seismic instrumentation does not actuate any 
protective equipment or play any direct role in the mitigation of an 
accident. The capability of the plant to withstand a seismic event 
or other design basis accident is determined by the initial design 
and construction of systems, structures, and components. This 
instrumentation is used to alert operators to the seismic event and 
evaluate the plant response.
    The proposed revisions to the Index pages, Table Index page 
(BVPS Unit No. 1 only), Specification pages and Bases pages are 
administrative in nature and do not affect plant safety.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed amendment does not involve any physical changes to 
the plant or the modes of plant operation defined in Appendix A of 
the operating license. The proposed amendment does not involve the 
addition or modification of plant equipment nor does it alter the 
design or operation of plant systems. Seismic instrumentation does 
not actuate any protective equipment or play any direct role in the 
mitigation of an accident. The capability of the plant to withstand 
a seismic event or other design basis accident is determined by the 
design and construction of systems, structures, and components. This 
instrumentation is used to alert operators to the seismic event and 
evaluate the plant response.
    Therefore, operation of the facility in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed amendment does not involve revisions to any safety 
limits or safety system setting that would adversely impact plant 
safety. The proposed amendment does not affect the ability of 
systems, structures or components important to ensure the safe 
shutdown of the facility, or the mitigation and control of accident 
conditions within the facility. In addition, the proposed amendment 
does not affect the ability of safety systems to ensure that the 
facility can be maintained in a shutdown or refueling condition for 
extended periods of time, or the availability of sufficient 
instrumentation and control capability for monitoring and 
maintaining the unit status.
    The proposed revisions to the Index pages, Table Index page 
(BVPS Unit No. 1 only), Specification pages and Bases pages are 
administrative in nature and do not affect plant safety.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: S. Singh Bajwa.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver 
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

    Date of amendment request: May 27, 1999.
    Description of amendment request: The proposed amendments would (1) 
revise the frequency for performing the CHANNEL FUNCTIONAL TEST (CFT) 
of the manual initiation functional units specified in the Beaver 
Valley Power Station, Unit Nos. 1 and 2, Engineered Safety Features 
Actuation System (ESFAS) Instrumentation Technical Specifications (TSs) 
from monthly, with an accompanying footnote which allows the manual 
initiation to be tested on a refueling interval, to each refueling 
interval; (2) Revise footnotes associated with TS ESFAS tables; (3) 
revise associated TS Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change revises the frequency notation specified for 
the channel functional test of the manual initiation functions 
listed on Table 4.3-2 of TS 3/4.3.2, ``Engineered Safety Feature 
Actuation System (ESFAS) Instrumentation.'' The proposed change 
revises the current TS requirement for surveillance testing these 
functions to clarify that testing be performed on a refueling basis. 
The revision to the surveillance frequency specified in Table 4.3-2 
does not physically impact the Instrumentation, its setpoints, or 
the actual frequency at which the manual initiation functions are 
tested. The revision eliminates the potential for confusion 
regarding the testing required for the manual initiation function by 
deleting Footnote (1) to Table 4.3-2. The proposed change to the 
Surveillance Requirements of Table 4.3-2 for the manual initiation 
functions eliminates the need for Footnote (1). Footnote (1) 
requires testing the manual actuation switches every 18 months and 
performing a Channel Functional Test on all other circuitry 
associated with manual safeguards actuation every 31 days. As there 
is no other circuitry for which a 31 day CFT is applicable, the 
proposed change simplifies the TS requirement consistent with the 
current Standard TS for Westinghouse plants. Footnote (1) is 
consistent with early versions of the Standard Technical 
Specifications of

[[Page 35205]]

NUREG-0452; however, later versions of the Standard Technical 
Specifications and the Improved Standard Technical Specifications of 
NUREG-1431 simply require testing manual initiation functions on a 
refueling or 18 month basis. The proposed refueling frequency for 
testing this instrumentation recognizes that the manual initiation 
functions can not be tested at power since this would introduce the 
potential for a significant plant transient.
    The deletion of Table 4.3-2 Footnote (1) resulted in renumbering 
Footnote (2) to (1). In addition, expired Unit 2 Table 4.3-2 
Footnote (3) (only applicable to the first refueling outage) was 
also deleted. In addition, changes to the TS bases are made to 
further clarify the channel functional test requirements. The 
reorganization of the Table 4.3-2 footnotes and bases modifications 
are considered to be editorial changes.
    The manual initiation instrumentation will continue to be tested 
in the same manner as before (every refueling). This test frequency 
is consistent with the licensing basis for testing this 
instrumentation described in the Updated Final Safety Analysis 
Report (UFSAR) and with the testing frequency specified in the 
standard Westinghouse Plant TS. Therefore, this test frequency is 
considered adequate to verify instrumentation operability. In 
addition, failure of a manual initiation function is not an accident 
initiator. As such, the ESFAS instrumentation will continue to be 
capable of providing the required safety functions described in the 
UFSAR. Therefore, operation of the facility in accordance with the 
proposed amendment does not involve a significant increase in the 
probability or consequence 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?
    There are no hardware changes associated with this license 
amendment nor are there any changes in the method by which any 
safety-related plant system performs its safety function. No new 
accident scenarios, transient precursors, failure mechanisms or 
limiting single failures are introduced as a result of these 
changes. These changes do not introduce any adverse effects or 
challenges to any safety-related systems. No change is required to 
any system configurations, plant equipment or analyses. Therefore, 
these changes will not create the possibility of any new or 
different kind of accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The margin of safety depends on the maintenance of specific 
operating parameters and systems within design requirements. 
Updating the manual initiation function surveillance interval 
requirements specified on ESFAS TS Table 4.3-2 and deleting Table 
4.3-2 Footnote (1) reflects the standard Westinghouse Plant TS 
requirements for this instrumentation and is consistent with the 
design and operation of the plant as described in the UFSAR. In 
addition, the proposed change does not reduce the current refueling 
interval testing performed on this instrumentation. The refueling 
test frequency specified for this instrumentation is consistent with 
industry standards and considered adequate to ensure the affected 
manual initiation functions are maintained operable. The proposed 
change will improve the clarity of the TS requirement by eliminating 
the potential for confusion as to when the surveillances are 
required to be performed. As such, the proposed change continues to 
ensure that the operation of the affected instrumentation is 
maintained within its design requirements and that it continues to 
be capable of providing the required safety functions described in 
the UFSAR. Therefore, operation of the facility in accordance with 
the proposed amendment will not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: S. Singh Bajwa.

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

    Date of amendment request: June 1, 1999.
    Description of amendment request: The proposed amendment would 
revise the surveillance requirements and applicable Bases relevant to 
inservice inspection requirements for the portions of the once-through 
steam generator (OTSG) tubes adjacent to the primary cladding region of 
the upper and lower OTSG tubesheets.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Criterion 1--Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The OTSGs are used to remove heat from the reactor coolant 
system during normal operation and during accident conditions. The 
OTSG tubing forms a substantial portion of the reactor coolant 
pressure boundary. An OTSG tube failure is a breach of the reactor 
coolant pressure boundary and is a specific accident analyzed in the 
Arkansas Nuclear One, Unit 1 (ANO-1), Safety Analysis Report (SAR).
    The purpose of the periodic surveillance performed on the OTSGs 
in accordance with ANO-1 Technical Specification (TS) 4.18 is to 
ensure that the structural integrity of this portion of the reactor 
coolant system will be maintained. The TS plugging limit of 40% of 
the nominal tube wall thickness requires tubes to be repaired or 
removed from service because the tube may become unserviceable prior 
to the next inspection. Unserviceable is defined in the TS as the 
condition of a tube if it leaks or contains a defect large enough to 
affect its structural integrity in the event of an operating basis 
earthquake, a loss-of-coolant accident, or a steam line or feedwater 
line break. The proposed TS change allows OTSG tubes with axial TEC 
[tube end cracking] indications that do not extend from the cladding 
region into the carbon steel interface within the tube-to-tubesheet 
rolled joint of the tubesheets to remain in service with existing 
degradation exceeding the existing 40% through-wall (TW) plugging 
limit.
    Extensive testing and plant experience has illustrated that TEC 
flaws confined to this area within the OTSG will not result in tube 
burst or significant tube leakage under MSLB [main steamline break] 
conditions. Potential leakage from tubes with TEC will be bounded by 
the MSLB evaluation presented in the SAR. Therefore, allowing TEC 
flaws in this specific region to remain in service will not alter 
the conditions assumed in the current ANO-1 accident analysis for 
OTSG tube failures under postulated accident conditions. In 
addition, the condition of the OTSG tubes in this region are 
monitored during regular inspection intervals to assess for evidence 
of growth. Any growth noted will be addressed through the 
operational assessment. Therefore, Entergy Operations has determined 
that the identification, monitoring, assessment, and corrective 
action programs * * * [associated with the proposed changes] 
sufficiently support this change request.
    Application of the TEC alternate repair criteria will allow 
leaving tubes with TEC indications found in the defined area of the 
tubesheets in service while ensuring safe operation by monitoring 
and assessing the present and future conditions of the tubes. 
Through the inspection, monitoring, and assessment programs 
previously mentioned, and the on-line leak detection capabilities 
available during plant operation, continued safe operation of ANO-1 
is reasonably assured.
    Therefore, the application of the TEC alternate repair criteria 
* * * does not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    Criterion 2--Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The implementation of the TEC alternate repair criteria will not 
result in any failure mode not previously analyzed. The OTSGs are 
passive components. The intent of the TS surveillance requirements 
are being met by these proposed changes in that adequate structural 
integrity will be maintained. Potential leakage under MSLB 
conditions will remain bounded by the current SAR analysis. 
Additionally, the proposed change does not introduce any new modes 
of plant operation.

[[Page 35206]]

    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3--Does Not Involve a Significant Reduction in the 
Margin of Safety.
    The application of an alternate repair criteria for TEC provides 
adequate assurance with margin that ANO-1 steam generator tubes will 
retain their structural integrity under normal and accident 
conditions. The structural requirements of TEC affected tubes have 
been evaluated satisfactorily and meet or exceed regulatory 
requirements. The tubing region where TEC occurs is constrained 
within the tubesheet bore; therefore, there is no additional risk 
associated with tube rupture. Main steam line break leakage rates 
for these tubes are reasonably assured to remain within the 
assumptions of the accident analysis by proper application of the 
TEC alternate repair criteria program. Because no appreciable impact 
is evidenced on the tubes structural integrity or its potential 
leakage rate, the margin to safety remains unaltered.

    Therefore, this change does not involve a significant reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: June 1, 1999.
    Description of amendment request: The amendments would revise the 
St. Lucie, Units 1 and 2, Technical Specifications (TS), Sections 
3.5.2, to allow up to 7 days to restore an inoperable Low Pressure 
Safety Injection System train to operable status. The amendments would 
also revise the associated surveillance requirements and TS Bases 
sections to be consistent with the revisions to TS Section 3.5.2. Minor 
editorial changes for the specified Recirculation Actuation Signal 
(RAS) verification test are also included to ensure the terminology 
used in the specification is consistent with plant design.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments for St. Lucie Plant, Units 1 and 2 will 
extend the action completion/allowed outage time (AOT) for a single 
inoperable Low Pressure Safety Injection (LPSI) train from 72 hours 
to 7 days. A LPSI train is designed as a part of each Emergency Core 
Cooling System (ECCS) subsystem to supplement Safety Injection Tank 
(SIT) inventory during the early stages of mitigating a Design Basis 
Accident. As such, components of the LPSI system are not accident 
initiators, and an extended AOT to restore operability of an 
inoperable LPSI train would not increase the probability of 
occurrence of accidents previously analyzed.
    The safety analyses for both St. Lucie Units demonstrate that 
ECCS performance acceptance criteria are satisfied with only one of 
the two redundant ECCS subsystems operating during the postulated 
Design Basis Accident. The proposed technical specification 
revisions involve the AOT for a single inoperable LPSI train, and do 
not change the conditions assumed for the minimum amount of 
operating equipment needed for accident mitigation. Therefore, the 
consequences of an accident previously evaluated will not be 
significantly increased.
    In addition to the preceding evaluation, a Probabilistic Safety 
Analysis (PSA) was performed to quantitatively assess the risk 
impact of the proposed amendments. It was concluded from the results 
of that assessment that the risk contribution of the AOT extension 
is very small, and that the net impact of the proposed amendment can 
be risk beneficial.
    The editorial corrections proposed for the specified RAS 
verification test do not alter existing test requirements and have 
no impact on the accident analyses. Therefore, operation of either 
facility in accordance with its proposed amendment would not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendments will not change the physical plant or 
the modes of plant operation defined in either Facility License. The 
changes do not involve the addition or modification of equipment nor 
do they alter the design of plant systems. Therefore, operation of 
either facility in accordance with its proposed amendment would not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The margin of safety associated with the ECCS system is 
established by acceptance criteria for system performance defined in 
10 CFR 50.46. The proposed amendments will not change these 
acceptance criteria or the operability requirements for equipment 
that is used to achieve such performance as demonstrated in the 
plant safety analyses. Moreover, an integrated assessment of the 
risk impact of extending the AOT for a single inoperable LPSI train 
has concluded that the risk contribution is very small, LPSI system 
reliability can potentially be improved, and the net impact of the 
proposed change can be risk beneficial. The editorial corrections 
proposed for the specified RAS verification test do not alter 
existing test requirements and have no impact on the accident 
analyses. Therefore, operation of either facility in accordance with 
its proposed amendment would not involve a significant reduction in 
a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Sheri R. Peterson.

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

    Date of amendment request: May 13, 1999.
    Description of amendment request: The proposed amendment would make 
changes to the TMI-1 Facility Operating License No. DPR-50 Sections 
2.a, 2.c.(3), and 2.c.(7) to delete obsolete or outdated portions of 
the license conditions, and would change the Bases for Technical 
Specification 3.1.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated. Most of the proposed amendment is only administrative; it 
adds to the Technical Specifications generic references to various 
documents. These changes have no affect upon the plant design or 
operation.

[[Page 35207]]

    The proposed change to the Technical Specification Bases 3.1.1 
is the removal of the specified pressurizer code safety valve flow-
rate for which no basis could be found and the acceptance of a 3% 
setpoint drift (as-found) as per the ASME code. The 3% code limit is 
in accordance with the plant's Inservice Test Program submittal, 
which was evaluated by the NRC staff for the current 10 year 
interval and documented under NRC TAC No. M93777. The [c]orrect 
pressurizer code safety valve flow is provided in the FSAR Table 
4.2-8. The proposed change is supported by a revise[d] Startup 
Accident analysis with the revised safety valve flow-rate at the 3% 
setpoint drift, which demonstrated that the acceptance criteria for 
the event were met with considerable margin. The proposed change 
does not affect the Technical Specification 3.1.1.a, pressurizer 
code safety valve operable (as-left) requirement of [plus or minus] 
1%.
    Therefore, operation in accordance with the proposed amendment 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated, because no 
new failure modes are created by the proposed changes. The 
administrative changes are cosmetic and have no impact on plant 
design or operation.
    3. Operation of the facility in accordance with the proposed 
amendment will not involve a significant reduction in a margin of 
safety. The proposed amendment does not change any operating limits 
for reactor operation.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: S. Singh Bajwa.

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

    Date of amendment request: May 26, 1999.
    Description of amendment request: The proposed amendment would 
approve changes to the TMI-1 Updated Final Safety Analysis Report 
(UFSAR) which would allow use of the EPRI (Electric Power Research 
Institute) Conservative Deterministic Failure Margin (CDFM) methodology 
for seismic analysis of the portions of the auxiliary steam line 
located in the Auxiliary, Control and Fuel Handling buildings at TMI-1. 
The licensee determined that these changes to the UFSAR required prior 
NRC approval in accordance with 10 CFR 50.59.
    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, use of CDFM methodology for the 
analysis of the auxiliary steam system piping, would not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The analysis of the auxiliary steam pipe using the CDFM 
methodology demonstrates that the pipe wall will maintain integrity 
sufficient to prevent adverse impact on safety related equipment 
during a safe shutdown earthquake (SSE). The methodology is based on 
actual earthquake experience data and has been shown to be adequate 
to demonstrate that piping systems will maintain integrity. The CDFM 
methodology was developed by experts in the field of seismic 
analysis and is based on actual earthquake experience and the 
results of dynamic tests with large seismic accelerations. The 
methodology provides a conservative mechanism for analytically 
predicting performance during actual earthquakes, and thus its 
application would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed amendment, use of CDFM methodology for the 
analysis of the auxiliary steam system piping, would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluateed.
    No changes to plant systems, structures or components are 
proposed and no changes to methods of operation [of the plant] are 
involved.
    3. The proposed amendment, use of CDFM methodology for the 
analysis of the auxiliary steam system piping, would not involve a 
significant reduction in a margin of safety.
    No changes are proposed to operating limits or safety system 
settings, or to accident analysis acceptance criteria. The CDFM 
methodology provides a conservative mechanism for analytically 
predicting system performance during actual earthquakes. Its 
application to the auxiliary steam system piping would not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: S. Singh Bajwa.

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

    Date of amendment request: June 4, 1999.
    Description of amendment request: The amendment revises decay heat 
removal capability requirements to ensure that at least two active 
methods of decay heat removal capability will be available during 
shutdown conditions except when the reactor vessel head is removed and 
the fuel transfer canal water level is greater than or equal to 23 feet 
above the reactor vessel flange.
    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:

    GPU Nuclear has determined that this Technical Specification 
Change Request poses no significant hazards as defined by NRC in 10 
CFR 50.92. 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 because the 
proposed changes would remove exceptions for decay heat removal 
system operability requirements during the time the plant is in a 
Refueling Shutdown with the RCS loop not filled. The proposed 
changes effectively add requirements to maintain redundancy in decay 
heat removal systems.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated because the proposed changes 
would not introduce any new failure modes or modify existing 
systems.
    3. Involve a significant reduction in a margin of safety because 
the proposed amendment would not involve changes to the safety 
limits, limiting safety system settings, or operating limits.

    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.

[[Page 35208]]

    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: S. Singh Bajwa.

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

    Date of amendment request: April 19, 1999.
    Description of amendment request: The proposed amendment would 
replace the current set of technical specifications for the Millstone 
Unit 1 plant with a new set of technical specifications for the 
permanently shutdown status of the plant.
    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, a summary of which is presented below:
    The proposed change does not:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    This proposed change is consistent with the STS [standard 
technical specifications]. The relocation of requirements from the 
MP1 TS [Millstone Unit 1 Technical Specifications] to the licensee 
controlled documents is consistent with the criteria set forth in 10 
CFR 50.36 for the content of Technical Specifications. The removal 
of definitions, generic LCO [limiting condition for operation] 
actions and generic surveillance requirements has no impact on 
facility SSCs [structure, system, and components] or the methods of 
operation of such SSCs. The deletion of design features and safety 
limits not applicable to the permanently shutdown and defueled 
status of MP1 has no impact on the remaining DBA [design-basis 
accident], the fuel handling accidents in the fuel storage pool. The 
removal of LCOs and surveillance requirements which are related only 
to the operation of the nuclear reactor or only to the prevention, 
diagnosis or mitigation of reactor-related transients or accidents 
do not affect the applicable DBA previously evaluated. The critical 
safety functions involving core reactivity control, reactor heat 
removal, reactor coolant system inventory control and containment 
integrity are no longer necessary at MP1. The proposed accidents 
involving damage to the reactor coolant system, main steam lines, 
reactor core, and the subsequent release of radioactive material are 
no longer possible at MP1. Fuel pool cooling and makeup related 
equipment and support equipment (e.g., electrical power systems) are 
not required to be continuously available since recent analysis 
demonstrated that there is up to ten days before fuel storage pool 
boiling to effect repairs, establish alternate sources of make up 
flow, or establish steady state natural air circulation cooling of 
the Reactor Building atmosphere and fuel storage pool water in the 
event of a loss of cooling and makeup flow to the fuel pool. The 
radioactive decay of the irradiated fuel since shutdown of the 
reactor in November, 1995 has reduced the consequences of the fuel 
handling accident to levels well below those previously analyzed. 
The relevant parameter (water level) associated with the fuel pool 
provides an initial condition for the fuel handling accident 
analyses and is included in the PDTS [Permanently Defueled Technical 
Specifications]. The Reactor Building crane LCOs are retained to 
preserve the engineered controls which preclude a spent fuel cask 
drop from occurring over the fuel storage pool. The deletion and 
modification of provisions of the administrative controls do not 
directly affect the design of SSCs necessary for safe storage of 
irradiated fuel or the methods used for handling and storage of such 
fuel in the fuel pool. The relocation of administrative controls 
related to quality assurance to the Northeast Utilities Quality 
Assurance Program is also consistent with the guidance provided in 
NRC Administrative Letter AL 95-06, ``Relocation of Technical 
Specification Administrative Controls Related to Quality 
Assurance,'' dated December 12, 1995. The changes to the 
administrative controls are administrative in nature and do not 
affect any accidents applicable to the safe storage of irradiated 
fuel or the permanently shutdown and defueled condition of the 
reactor. Therefore, the proposed changes to the MP1 TS do not 
involve any increase in the probability or consequences of any 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes have no impact on facility SSCs affecting 
the safe storage of irradiated fuel or on the methods of operation 
of such SSCs, or handling and storage of such fuel. These changes 
are consistent with the STS and add to the clarity and ease of use 
of the proposed PDTS. The removal of Technical Specifications which 
are related only to the operation of the nuclear reactor or only to 
the prevention, diagnosis, or mitigation of reactor-related 
transients or accidents cannot result in different or more adverse 
failure modes or accidents than previously evaluated because the 
reactor is permanently shutdown and defueled and MP1 is no longer 
authorized to operate the plant. The proposed deletion of provisions 
of the MP1 TS do not affect systems credited in the accident 
analyses for the fuel handling accident in the fuel storage pool at 
MP1. The proposed PDTS continue to require proper control and 
monitoring of safety significant parameters and activities. The 
proposed restriction on the fuel pool level is fulfilled by normal 
operating conditions and preserves initial conditions assumed in the 
analyses of the postulated DBA. Reactor Building crane LCOs are 
retained from current Technical Specifications to preclude the 
possibility of a spent fuel cask drop over the fuel storage pool. 
Therefore, the proposed changes to this section of the MP1 TS would 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The deletion of provisions of the MP1 TS, which are not related 
to the storage of irradiated fuel or which are inconsistent with the 
scope of the STS, will not affect the analyses of the remaining DBA 
applicable to MP1. The postulated DBAs involving the reactor are no 
longer possible due to the permanently shutdown and defueled 
condition of the reactor. The requirements for SSCs which have been 
deleted from the MP1 TS are not credited in the existing accident 
analyses for the remaining applicable postulated accidents and 
therefore, do not contribute to the margin of safety associated with 
the accident analysis. Therefore, the proposed changes to this 
section of the MP1 TS do not involve any reduction in a margin of 
safety.

Conclusion

    NNECO has concluded that the proposed change to the MP1 Technical 
Specifications does not involve a significant hazards consideration as 
defined by 10 CFR 50.92.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: Michael T. Masnik.

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

    Date of amendment request: January 25, 1996, as supplemented April 
26, 1996, September 12, 1996, March 17, 1997, September 9, 1997, 
December 30, 1998, and May 19, 1999.
    Description of amendment request: The proposed changes extend the 
allowed outage time for an emergency diesel generator (EDG) system from 
7 to

[[Page 35209]]

14 days. At FitzPatrick, an EDG system consists of 2 EDGs powering one 
of two emergency AC power buses. The proposal includes provisions for a 
Configuration Risk Management Program (CRMP) consistent with the 
guidance of Regulatory Guide (RG) 1.177, ``An Approach for Plant-
Specific, Risk-Informed Decisionmaking: Technical Specifications.'' The 
NRC staff had previously published a notice on these topics on March 
27, 1996 (61 FR 13532). This revised notice on these topics is required 
to address revisions made in the licensee's supplemental submittals.
    The licensee's January 25, 1996, submittal also proposed two line-
item changes to reduce EDG testing at power and to revise AC power 
requirements for cold shutdown and refueling modes. The two line-item 
changes have not been affected by the supplemental information provided 
by the licensee, so the March 27, 1996, proposed finding of no 
significant hazards considerations remains valid for these items.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the FitzPatrick plant in accordance with the 
additional changes to the proposed Amendment discussed above, would not 
involve a significant hazards consideration as defined in 10 CFR 50.92, 
since it would not:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to the Technical Specifications will allow 
longer Allowed Out of Service Times to perform necessary repair and 
maintenance on Emergency Diesel Generators while at power. This 
extended AOT [allowed outage time] will enhance scheduling of 
preventive maintenance of individual EDGs without significantly 
increasing the probability or consequences of an accident previously 
evaluated. The risk evaluations for the EDGs determined that the 
probability of an accident by increasing the AOT for an EDG System 
from 7 days to 14 days is non-risk-significant.
    Increasing the EDG AOT does not involve physical alteration of 
any plant equipment and does not affect analysis assumptions 
regarding functioning of required equipment designed to mitigate the 
consequences of accidents. Further, the severity of postulated 
accidents and resulting radiological effluent releases will not be 
affected by the increased AOT for an EDG System.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

    [The CRMP provides administrative controls to ensure equipment 
configurations do not result in any significant increase in plant 
risk. In RG 1.177, the NRC staff established a standard for the 
content of the CRMP. The licensee's proposal is consistent with that 
standard, and so does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.]

2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.

    Extending the AOT for an EDG system does not necessitate 
physical alteration of the plant or changes in parameters governing 
normal plant operation. Thus, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated for [the] JAF [FitzPatrick] plant.

[The CRMP provides administrative controls to ensure equipment 
configurations do not result in any significant increase in plant 
risk. These administrative controls do not create any new equipment 
configurations, or provide for operation of equipment in a new or 
different manner. Therefore, the CRMP does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.]

    3. Involve a significant reduction in the margin of safety.
    As discussed above, a Fitzpatrick evaluation determined that the 
change in risk associated with extending the AOT for a[n] EDG System 
is non-risk-significant. In addition, the design provides adequate 
redundancy for safe shut down during the AOT with an EDG System out 
of service. This is supported by the LOCA [loss-of-coolant accident] 
analyses including analyses for long term suppression pool cooling 
and reactor shutdown cooling.

[The CRMP provides administrative controls to ensure equipment 
configurations do not result in any significant increase in plant 
risk. These administrative controls do not create any new equipment 
configurations, or provide for operation of equipment in a new or 
different manner. Therefore, the proposed CRMP 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: S. Singh Bajwa.

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

    Date of amendment request: May 24, 1999.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to correct typographical and 
editorial errors, and is considered administrative in nature.
    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 proposed editorial and administrative changes involve 
typographical errors and/or reflect changes that were previously 
reviewed and approved by the NRC. These changes, therefore, do not 
modify or add any initiating parameters that would significantly 
increase the probability or consequences of any previously analyzed 
accident.
    (2) The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    These proposed changes do not involve any potential initiating 
events that would create the possibility of a new or different kind 
of accident. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) The proposed change does not involve a significant reduction 
in a margin of safety.
    These changes are editorial in nature and/or reflect information 
previously reviewed and approved by the NRC. The proposed changes 
will make the information in the TS consistent with that already 
approved by the NRC. Therefore, the proposed changes do not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.
     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.

[[Page 35210]]

Sacramento Municipal Utility District (the District), Docket No. 50-
312, Rancho Seco Nuclear Station, Sacramento County, California

    Date of amendment request: April 23, 1999.
    Description of amendment request: The proposed amendment would 
change Permanently Defueled Technical Specification (PDTS) D3/4.1, 
``Spent Fuel Pool Level,'' to replace a specific reference to spent 
fuel pool (SFP) level alarm switches with a generic reference to SFP 
level instrumentation. This would allow the licensee to replace the old 
level alarm switches with a new ultrasonic level transmitter.
    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:

    PA-193 will not create a significant increase in the probability 
or consequences of an accident previously evaluated in the SAR 
[Safety Analysis Report], because the proposed PDTS change is 
editorial in nature and only changes the type of equipment that is 
referenced in surveillance specification D4.1.2. The SFP level 
instrument reference in D4.1.2 is changed from a specific reference 
(i.e., SFP level alarm switches) to a more generic reference (i.e., 
SFP level instrumentation). In addition:
    1. SFP level monitoring instrumentation is not relied on to 
mitigate the consequences of the accidents analyzed in the SAR 
(i.e., Fuel Handling Accident, Loss-Of-Offsite-Power event, Liquid 
Tank Ruptures, and Decommissioning Accidents),
    2. PA-193 does not alter the SFP level monitoring, SFP cooling, 
or fuel handling functions during the PDM [Permanently Defueled 
Mode],
    3. PA-193 continues to require an 18-month calibration of SFP 
level instrumentation, and
    4. SFP level and alarm indication in the Control Room is 
maintained with the new SFP level instrumentation. Also, the SFP 
level alarm setpoints remain unchanged with the new SFP level 
detection system.
    PA-193 will not create the possibility of a new or different 
type of accident than previously evaluated in the SAR, because SFP 
level instrumentation does not provide any control function and does 
not affect any equipment associated with SFP cooling, fuel handling, 
or inventory control. The proposed wording change to PDTS D4.1.2 
accommodates upgrading the SFP level instrumentation without 
changing the intent of surveillance specification D4.1.2. Also, the 
new SFP level detection system will (1) maintain the existing SFP 
level alarm setpoints and Control Room indication features and (2) 
have no adverse impact on the SFP level monitoring function.
    PA-193 will not involve a significant reduction in the margin of 
safety, because the proposed PDTS change is editorial in nature and 
necessary and only accommodates replacing an unreliable, antiquated 
SFP level monitoring system with a new, state-of-the-art, ultrasonic 
level detection system. The new SFP level detection system will 
improve the accuracy, reliability, and serviceability of the SFP 
level monitoring function. The District is maintaining the 
requirement to perform a[n] SFP level calibration and is only 
changing the type of equipment that is referenced in D4.1.2 from a 
specific reference (i.e., SFP level alarm switches) to a more 
generic reference (i.e., SFP level instrumentation).

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendment involves no significant hazards consideration.
    Local Public Document Room location: Central Library, Government 
Documents, 828 I Street, Sacramento, California 95814.
    Attorney for licensee: Dana Appling, Esq., Sacramento Municipal 
Utility District, P.O. Box 15830, Sacramento, California 95852-1830.
    NRC Section Chief: Michael T. Masnik.

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

    Date of amendment requests: June 8, 1999 (PCN-495).
    Description of amendment requests: The licensee has re-evaluated 
its small break loss-of-coolant accident (SBLOCA) using ABB Combustion 
Engineering (ABB-CE) S2M evaluation model. Based on this re-evaluation, 
the licensee proposes to revise the Technical Specifications (TSs) for 
the San Onofre Nuclear Generating Station (SONGS) Units 2 and 3 to 
reflect that charging flow is not required to mitigate the effects of 
the SBLOCA, add a surveillance requirement to verify that each charging 
pump is operable for boration based on the Inservice Testing Program, 
increase the maximum as-found lift pressure positive tolerance of main 
steam safety valves (MSSVs) from +1% to +2% of the lift setting, and 
list the ABB-CE S2M model as an acceptable method for determining 
linear heat rate. The licensee will also revise the TS Bases and the 
Updated Final Safety Analysis Report (UFSAR) to reflect the proposed 
changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Involve a significant increase in the probability or 
consequences of any accident previously evaluated?
    Response: No.
    The new Small Break Loss Of Coolant Accident (SBLOCA) evaluation 
model (ABB Combustion Engineering (ABB-CE) S2M SBLOCA evaluation 
model, CENPD 137 Supplement 2-P-A, ``Calculative Methods of the ABB-
CE Small Break LOCA Evaluation Model,'' dated April 1998) more 
accurately models the heat transfer mechanisms that occur during a 
SBLOCA. As a result of this modeling improvement, there is no longer 
a need to credit charging flow during a SBLOCA. The reanalysis, with 
an as-found tolerance of +2%/-3% of the lift setting on Main Steam 
Safety Valves (MSSVs) 2(3)-PSV-8401 and 2(3)-PSV-8410 in Table 
3.7.1-2, determined that the peak cladding temperature (PCT) that 
occurs in a SBLOCA is within the acceptance criteria limit of 2200 
[degrees] F specified in 10CFR50.46.
    This proposed change removes the charging pump Emergency Core 
Cooling System (ECCS) surveillance requirement from the Technical 
Specifications (TS) which effectively removes the charging system 
from the ECCS. This is based on the SBLOCA reanalysis using the new 
ABB-CE S2M SBLOCA evaluation model. The reanalysis using the new 
model did not credit charging system flow to the reactor coolant 
system.
    Because this proposed change to remove the charging pump ECCS 
flow surveillance requirement is based on a reanalysis of the SBLOCA 
rather than physical changes to the plant or the way it is operated, 
the probability of the SBLOCA is not affected. The results of the 
reanalysis demonstrate the consequences of the SBLOCA without 
charging flow do not exceed the consequences of the limiting LOCA. 
This is based on the fact that the SBLOCA PCT [peak clad 
temperature] does not exceed the limiting large break LOCA PCT.
    The addition of Surveillance Requirement (SR) 3.1.9.5 to require 
the charging pump to be tested in accordance with the Inservice 
Testing (IST) program will ensure that the charging pumps remain 
capable of performing their emergency boration requirements.
    Use of the NRC approved ABB-CE S2M SBLOCA analysis methodology 
identified in TS 5.7.1.5 for calculating the core operating limits 
further assures that there is no significant increase in the 
probability or consequences of any accident.
    Therefore, the probability or consequences of any accident 
previously evaluated are not increased.
    (2) Create the possibility of a new or different kind of 
accident from any previously evaluated?
    Response: No.
    This change does not involve a physical change to the plant, or 
a change to the way the plant is operated. The as-left tolerance of 
[plus or minus] 1% on MSSVs 2(3)-PSV-8401 and 2(3)-PSV-8410 in Table 
3.7.1-2 is not being changed. The charging system will still be 
verified capable of meeting its emergency boration requirements.

[[Page 35211]]

    Use of the NRC approved ABB-CE S2M SBLOCA analysis methodology 
identified in TS 5.7.1.5 for calculating the core operating limits 
further assures that there is no increase in the possibility of a 
new or different kind of accident from any previously evaluated. 
Therefore, the possibility of a new or different kind of accident 
from any previously evaluated is not created.
    (3) Involve a significant reduction in a margin of safety?
    Response: No.
    This proposed change to remove the ECCS surveillance requirement 
for the charging pumps, and increase the as-found tolerance on MSSVs 
2(3)-PSV-8401 and 2(3)-PSV-8410, is based on a SBLOCA reanalysis 
using the new ABB-CE S2M SBLOCA evaluation model. The NRC Safety 
Evaluation for the ABB-CE S2M evaluation model determined that the 
new evaluation model contains sufficient conservatism such that an 
adequate margin of safety exists when the S21VI evaluation model is 
used. The results of the SBLOCA reanalysis are within the acceptance 
criteria specified in 10 CFR 50.46.
    Testing of the charging pumps per the Inservice Testing Program, 
combined with the existing Technical Specification 3.1.9--``Boration 
System--Operating'' surveillance requirements ensure that the 
emergency boration requirements remain met without any reduction in 
a margin of safety.
    Use of the NRC approved S2M ABB-CE SBLOCA analysis methodology 
identified in TS 5.7.1.5 for calculating the core operating limits 
further assures that there is no significant reduction in any margin 
of safety.

    Therefore, a significant reduction in margin of safety is not 
involved.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, Irvine, California 92713.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: June 7, 1999.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 2.2.1, Reactor Trip System (RTS) 
Instrumentation Setpoints, and TS 3.3.2, Engineered Safety Features 
Actuation System (ESFAS) Instrumentation, and the associated Bases, by 
removing the Total Allowance (TA), Sensor Error (S), and Z terms from 
the RTS and ESFAS Instrumentation Trip Setpoints Tables. This would 
replace the five-column methodology with a two-column methodology that 
consists of the trip setpoint and allowable value columns.
    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:

    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change eliminates the option to evaluate the 
equation (Z+R+S [is less than or equal to] TA), within 12 hours, 
from Technical Specification 2.2.1, when the trip setpoint is 
outside the allowable value limit. The equation established a 
threshold for submitting a Licensee Event Report. The change does 
not affect the probability of an accident. The evaluation of the 
equation is an administrative provision and has no relevance to the 
initiation of any analyzed event. The consequences of an accident 
are not affected. The change will not alter assumptions relative to 
the mitigation of an accident or transient event.
    The proposed amendment is a programmatic and administrative 
change that does not physically alter safety-related systems, nor 
does it affect the way in which safety-related systems perform their 
functions. Because the design of the facility and system operating 
parameters are not being changed, the proposed amendment does not 
involve an increase in the probability or consequences of any 
accident previously evaluated.
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed amendment is a programmatic and administrative 
change that does not physically alter safety-related systems, nor 
does it affect the way in which safety-related systems perform their 
functions. The changes in methods governing normal plant operation 
are consistent with current safety analysis assumptions. The 
proposed change eliminates the option to evaluate the equation 
(described above) within 12 hours, when the trip setpoint is outside 
the allowable limit. Because the design of the facility and system 
operating parameters are not being changed, the proposed amendment 
does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed amendment is a programmatic and administrative 
change that provides assurance that plant operations continue to be 
conducted in a safe manner. As stated above, the proposed amendment 
does not physically alter safety-related systems, nor does it affect 
the way in which safety-related systems perform their functions. The 
proposed change eliminates the option to evaluate the equation 
(described above) within 12 hours, when the trip setpoint is outside 
the allowable limit.
    The margin of safety is not affected by eliminating an 
administrative provision in Technical Specifications. The 
determination for submitting a Licensee Event Report when a trip 
setpoint is outside the allowable value will be performed with the 
guidelines of 10CFR50.73. The safety analysis assumptions will still 
be maintained, thus, no question of safety exists. Because the 
design of the facility and system operating parameters are not being 
changed, the proposed amendment does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 
77488.
    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, Docket Nos. 50-260, 50-296, Browns Ferry 
Nuclear Power Plant, Units 2 and 3. Limestone County, Alabama

    Date of amendment request: March 12, 1997 as supplemented by 
letters dated March 30, 1999, April 23, 1999 and June 18, 1999.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to extend, from 7 days to 14 days, 
the Allowable Outage Time (AOT) applicable to an inoperable emergency 
diesel generator.
    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:

No Significant Hazards Consideration Determination

    TVA has concluded that operation of BFN in accordance with the 
proposed change to the TS does not involve a significant hazards 
consideration. TVA's conclusion is based on it's evaluation, in 
accordance with 10 CFR 50.91(a)(1), of the three standards set forth 
in 10 CFR 50.92(c).
    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The EDGs are designed as backup AC power sources in the event of 
loss of off-site

[[Page 35212]]

power. The proposed AOT does not change the conditions, operating 
configurations, or minimum amount of operating equipment assumed in 
the safety analysis for accident mitigation. No changes are proposed 
in the manner in which the EDGs provide plant protection or which 
create new modes of plant operation. In addition, a PSA evaluation 
concluded that the risk contribution of the AOT extension is non-
risk significant. 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 proposed change does not introduce any new modes of plant 
operation or make physical changes to plant systems. Therefore, 
extension of the allowable AOT for EDGs does not create the 
possibility of a new or different accident.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    BFN's emergency AC system is designed with sufficient redundancy 
such that an EDG may be removed from service for maintenance or 
testing. The remaining EDGs are capable of carrying sufficient 
electrical loads to satisfy the UFSAR requirements for accident 
mitigation or unit safe shutdown.
    Increasing the allowable EDG AOT will likely increase EDG 
unavailability on the average since it expected that the provision 
would occasionally be used to accommodate unplanned major EDG 
maintenance. However, a conservative PSA evaluation concluded that 
the risk contribution of the AOT extension is non-risk significant. 
For the 12-year EDG PM work activity, it is expected that the 
proposed TS would actually reduce unavailability since multiple 
outages would not be necessary to accomplish the maintenance 
activity.
    The proposed change does not impact the redundancy or 
availability requirements of off-site power supplies or change the 
ability of the plant to cope with station blackout events. For these 
reasons, 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.
    Local Public Document Room Location: Athens Public Library, 405 E. 
South Street, Athens, Alabama.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Drive, ET 10H, Knoxville, Tennessee 37902,
    NRC Section Chief: Sheri R. Peterson.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station (CPSES), Units 1 and 2, Somervell County, Texas

    Date of amendment request: May 4, 1999, as supplemented by letter 
dated June 4, 1999.
    Brief description of amendments: The proposed license amendments 
would revise the Technical Specifications for CPSES, Units 1 and 2. 
Specifically, the changes would revise the surveillance requirements 
associated with the plant battery and emergency diesel generators, and 
correct miscellaneous editorial errors that resulted from the issuance 
of Amendment No. 64. The original application was noticed and published 
in the Federal Register on June 2, 1999 (64 FR 29715). The June 4, 
1999, supplement provided proposed additional editorial corrections. 
The supplemental information is being noticed herein to address the 
issue of no significant hazards consideration.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequence of an accident previously evaluated?
    (1) Batteries are used to support mitigation of the consequences 
of an accident, and are not considered to be an initiator of any 
previously analyzed accident. The proposed change would not effect 
the design or performance of the batteries. The allowance to perform 
the modified performance discharge test in lieu of the service test 
at any time is permissible since the test's discharge rate envelopes 
the duty cycle of the service test. Therefore, the allowance for 
unrestricted substitution of the modified performance discharge test 
in lieu of the service discharge test does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) The diesel generators are used to support mitigation of the 
consequences of an accident, and are not considered to be an 
initiator of any previously analyzed accident. The proposed change 
does not affect the accident analysis assumption that the DG reaches 
minimum conditions to accept load within 10 seconds. The ability of 
the DG to maintain steady state operation within 10 seconds is not 
an accident analysis assumption and is primarily used to identify 
degradation of governor and voltage regulator performance. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (3) The editorial changes are non-technical and therefore do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    (1) The allowance for unrestricted substitution of the modified 
performance discharge test in lieu of the service discharge test 
does not involve any physical alteration to the plant. No new 
failure mechanisms will be introduced and the change does not affect 
the ability of the batteries to fulfill their safety-related 
function. Therefore, this change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    (2) The separation of the DG start surveillance criteria into 
those criteria required to be met within 10 seconds, and those 
criteria required to be met following achievement of steady state 
conditions, does not involve any physical alteration to the plant. 
No new failure mechanisms will be introduced and the change does not 
affect the ability of the DGs to fulfill their safety-related 
function. Therefore, this change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    (3) The editorial changes are non-technical and therefore do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    (1) The allowance for unrestricted substitution of the modified 
performance discharge test in lieu of the service discharge test 
will not alter any accident analysis assumptions, initial 
conditions, or results. Consequently, it does not have any effect on 
the margin of safety. Therefore, this change does not involve a 
significant reduction in a margin of safety.
    (2) The proposed change to delete the requirement to demonstrate 
that the DG can achieve and maintain steady state operation within 
10 seconds is not an accident analysis assumption. The accident 
analysis assumption that the DG reaches minimum conditions to accept 
load within 10 seconds is preserved. Consequently, it does not have 
any effect on the margin of safety. Therefore, this change does not 
involve a significant reduction in a margin of safety.
    (3) The editorial changes are non-technical and therefore 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.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

[[Page 35213]]

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station (CPSES), Units 1 and 2, Somervell County, Texas

    Date of amendment request: May 14, 1999.
    Brief description of amendments: The proposed license amendments 
would change the name of the CPSES licensee from ``Texas Utilities 
Electric Company'' to ``TXU Electric Company'' in the Facility 
Operating Licenses of CPSES, Units 1 and 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequence of an accident previously evaluated?
    No. This request involves an administrative change only. The 
Operating Licenses (OLs) are being changed to reference the new 
corporate name of the licensee. No actual plant equipment or 
accident analyses will be affected by the proposed change. 
Therefore, TU [Texas Utilities] Electric concludes that this request 
will have no impact on the possibility of any type of accident, 
whether new, different or previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. This request involves an administrative change only. The OLs 
are being changed to reference the new corporate name of the 
licensee. No actual plant equipment or accident analyses will be 
affected by the proposed change and no failure modes not bounded by 
previously evaluated accidents will be created. Therefore, TU 
Electric concludes that this request will have no impact on the 
possibility of any type of accident, whether new, different or 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. Margin of safety is associated with confidence in the 
ability of the fission product barriers (i.e., fuel and fuel 
cladding, Reactor Coolant System pressure boundary, and containment 
structure) to limit the level of radiation dose to the public. This 
request involves an administrative change only. The OLs are being 
changed to reference the new corporate name of the licensee. No 
actual plant equipment or accident analyses will be affected by the 
proposed change. Additionally, the proposed change will not relax 
any criteria used to establish safety limits, will not relax any 
safety systems settings, or will not relax the bases for any 
limiting conditions of operation. Therefore, this request will not 
impact margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station (CPSES), Units 1 and 2, Somervell County, Texas

    Date of amendment request: May 24, 1999.
    Brief description of amendments: The proposed license amendments 
would remove several cycle-specific parameter limits from the Technical 
Specifications (TSs) and add parameter limits to the Core Operating 
Limits Report. In addition, the core safety limit curves would be 
replaced with safety limits more directly applicable to the fuel and 
fuel cladding fission product barriers. The affected TSs are: (1) TS 
2.0, ``Safety Limits (SLs)''; (2) TS 3.3.1, ``Reactor Trip System 
Instrumentation Setpoints''; (3) TS 3.4.1, ``RCS pressure temperature 
and flow from Nucleate Boiling (DNB) Limits''; and (4) TS 5.6.5, ``Core 
Operating Limits 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 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes remove cycle-specific parameter limits from 
the Technical Specifications, add them to the list of limits 
contained in the Core Operating Limits Report (COLR), and revise the 
Administrative Controls section of the Technical Specifications. The 
proposed changes also insert the original minimum RCS [reactor 
coolant system] flow limits into the Technical Specifications. The 
changes do not, by themselves, alter any of the parameter limits. 
The changes are administrative in nature and have no adverse effect 
on the probability of an accident or on the consequences of an 
accident previously evaluated. The removal of parameter limits from 
the Technical Specifications does not eliminate the requirement to 
comply with the parameter limits.
    The parameter limits in the COLR may be revised without prior 
NRC approval. However, [Technical] Specification 5.6.5c continues to 
ensure that the parameter limits are developed using NRC-approved 
methodologies and that applicable limits of the safety analyses are 
met. While future changes to the COLR parameter limits could result 
in event consequences which are either slightly less or slightly 
more severe than the consequences for the same event using the 
present parameter limits, the differences would not be significant 
and would be bounded by the requirement of specification 5.6.5c to 
meet the applicable limits of the safety analysis.
    Based on the above, addition of the minimum RCS flow limit into 
the Technical Specifications, removal of the parameter limits from 
the Technical Specifications and the addition of the described 
limits in the COLR, thus allowing revision of the parameter limits 
without prior NRC approval, has no significant effect on the 
probability or consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes add the minimum RCS flow limit into the 
Technical Specifications, remove certain parameter limits from the 
Technical Specifications and add these limits to the list of limits 
in the COLR, thus removing the requirement for prior NRC approval of 
revisions to those parameters. The changes do not add new hardware 
or change plant operations and therefore cannot initiate an event 
nor cause an analyzed event to progress differently. Thus, the 
possibility of a new or different kind of accident is not created.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The margin of safety is the difference between the acceptance 
criteria and the associated failure values. The proposed changes do 
not affect the failure values for any parameter. Through the 
accident analyses, all applicable limits (i.e., relevant event 
acceptance criteria as described in the NRC-approved analysis 
methodologies) are shown to be satisfied; therefore, there is no 
impact on event acceptance criteria. Because neither the failure 
values nor the acceptance criteria are affected, the proposed change 
has no effect on 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.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

[[Page 35214]]

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

    Date of amendment request: May 5, 1999.
    Description of amendment request: The proposed change modifies the 
Technical Specifications (TS) to enhance limiting conditions for 
operation and surveillance requirements relating to the Standby Liquid 
Control (SLC) system and incorporates certain provisions of NRC's rule 
on anticipated transients without scram (ATWS) (10CFR50.62). The change 
involves the use of enriched boron in the SLC system and improves upon 
other aspects of the TS for this 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 operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment, will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change deletes the requirement for standby liquid 
control (SLC) system operability during refueling and modifies the 
conditions for allowing the system to be inoperable when shutdown.
    This change also permits changing the reactor mode switch to the 
``Run'' or ``Startup/Hot Standby'' position to test mode switch 
interlock functions while the SLC system is inoperable. To allow 
testing of instrumentation associated with the reactor mode switch 
interlock functions, compensatory measures are provided for assuring 
that no core alterations are in progress and that all control rods 
remain fully inserted in core cells containing one or more fuel 
assemblies. These compensatory measures ensure that no credible 
mechanisms for an inadvertent criticality are introduced by 
administratively controlling the required functions of the reactor 
mode switch interlocks. Control rods are not required to be inserted 
in empty core cells (i.e., those containing no fuel) because, with 
one or more cells in this configuration, the overall shutdown margin 
is actually greater than when all control rods and all fuel 
assemblies are inserted.
    The SLC system is not assumed in the initiation of any 
previously evaluated events and therefore the proposed change will 
not significantly increase the probability or consequences of a 
previously analyzed accident. The SLC system is not assumed to 
operate in the mitigation of any previously analyzed accidents which 
are assumed to occur during shutdown or refueling conditions. This 
change will not result in operation that will significantly increase 
the probability of initiating an analyzed event. This change will 
not alter assumptions relative to mitigation of an accident or alter 
the operation of process variables, structures, systems, or 
components as described in the final safety analysis report.
    VY has determined that the proposed change to increase the 
standby liquid control system reactivity control capacity using a 
borated water solution enriched in the boron-10 isotope effectively 
increases the rate of injection of neutron absorber and does not 
alter the function of the system, method of operation or dual train 
configuration. The system response time to an anticipated transient 
without scram (ATWS) event has been reduced as the increased boron-
10 enrichment of the solution provides faster negative reactivity 
insertion, thus reducing the consequences of the ATWS event. The SLC 
system is not credited in any of the design basis accident analyses 
and, as such, is considered to provide only an additional mitigative 
feature in the event of an accident. The SLC system sodium 
pentaborate solution concentration and flow rate required by the 
ATWS rule (10CFR50.62) for reactivity control independent of the 
control rods are not reduced from the values previously evaluated 
and presented in the Vermont Yankee Technical Specifications. The 
addition of enriched boron provides a shutdown margin greater than 
the previously calculated shutdown reactivity control capacity, and 
the change does not affect the probability of an ATWS event.
    Therefore, this change will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment, will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change modifies the modes of applicability for the 
SLC system. Included in this change is allowance to permit changing 
the reactor mode switch to the ``Run'' or ``Startup/Hot Standby'' 
position to test mode switch interlock functions while the SLC 
system is inoperable. Precautions are taken when manipulating the 
mode switch to one of these positions to maintain all control rods 
fully inserted in core cells containing at least one fuel assembly 
and to not allow any core alterations. These two provisions 
eliminate the possibility of introducing any credible mechanisms for 
inadvertent criticality. The proposed change will not involve a 
physical alteration of the plant (no new or different type of 
equipment will be installed) or changes in methods governing normal 
plant operation. The proposed change will not eliminate any valid 
requirements necessary for safe operation.
    VY has determined that the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated because the proposed change involves a system 
whose function is to provide an additional (backup) mitigative 
shutdown capability and no system modifications are made.
    The addition of enriched boron does not affect any system or 
component that could initiate an accident. Thus, no new or different 
type of accident is created.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    VY has determined that the proposed change does not involve a 
significant reduction in a margin of safety. The proposed change 
would remove the backup to the available reactivity control systems 
when the reactor is in a shutdown or refueling condition. However, 
this backup is not considered in the margin of safety when 
determining the required reactivity for shutdown and refueling 
events. This change will have no impact on any safety analysis 
assumptions.
    Included in this change is allowance to permit changing the 
reactor mode switch to the ``Run'' or ``Startup/Hot Standby'' 
position to test mode switch interlock functions while the SLC 
system is inoperable. The margin of safety will not be reduced 
during such testing of interlock functions with the SLC system 
inoperable because compensatory measures have been added to ensure 
that no credible mechanisms for inadvertent criticality exist with 
the reactor mode switch in other than the ``Shutdown'' or ``Refuel'' 
positions.
    The use of enriched boron in the SLC system sodium pentaborate 
solution actually increases the capability of the SLC system to 
achieve cold shutdown; thus, no margin of safety is reduced.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington

    Date of amendment request: June 3, 1999.
    Description of amendment request: The request is to amend the 
operating license such that the name of the licensee is changed from 
Washington Public Power Supply System to Energy Northwest. The name of 
the facility will be changed from WPPS Nuclear Project No. 2 to WNP-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:


[[Page 35215]]


    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This request involves an administrative change only. The 
Operating License (OL) is being changed to reference the new name of 
the licensee. No actual plant equipment or accident analyses will be 
affected by the proposed change. Therefore, this request will have 
no impact on the probability or consequence of any type of accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This request involves an administrative change only. The OL is 
being changed to reference the new name of the licensee. No actual 
plant equipment or accident analyses will be affected by the 
proposed change and no failure modes not bounded by previously 
evaluated accidents will be created. Therefore, this request will 
have no impact on the possibility of any new type of accident: new, 
different, or previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Margin of safety is associated with the confidence in the 
ability of the fission product barriers (i.e., fuel and fuel 
cladding, Reactor Coolant System pressure boundary, and containment 
structure) to limit the level of radiation dose to the public. This 
request involves an administrative change only. The OL is being 
changed to reference the new name of the licensee.
    No actual plant equipment or accident analyses will be affected 
by the proposed change. Additionally, the proposed change will not 
relax any criteria used to establish safety limits, will not relax 
any safety system settings, or will not relax the bases for any 
limiting conditions of operation. Therefore, this request will not 
impact the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352.
    Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: June 10, 1999.
    Description of amendment request: The amendment would revise 
Technical Specification Table 3.3-4, Functional Unit 7.b., Automatic 
Switchover to Containment Sump (Refueling Water Storage Tank Level--
Low-Low) to reflect the results of calculations that were performed for 
the associated instrumentation setpoints to consider the density 
variations due to temperature and boric acid concentration.
    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 protection system performance will remain within the bounds 
of the previously performed accident analysis. The protection 
systems will continue to function in a manner consistent with the 
plant design basis. The proposed changes will not affect any of the 
analysis assumptions for any of the accidents previously evaluated, 
since the changes are consistent with the setpoint methodology and 
ensure adequate margin to the Safety Analysis Limit. The proposed 
changes will not affect any event initiators nor will the proposed 
changes affect the ability of any safety related equipment to 
perform its intended function. There will be no degradation in the 
performance of nor an increase in the number of challenges imposed 
on safety related equipment assumed to function during an accident 
situation. There will be no change to normal plant operating 
parameters or accident mitigation capabilities.
    Therefore these changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no changes in the method by which any safety related 
plant system performs its safety function. The normal manner of 
plant operation remains unchanged, and no new equipment is being 
introduced. The increase in the RWST [refueling water storage tank] 
Level Low-Low Allowable Value still provides acceptable margin 
between the nominal Trip Setpoint and Allowable Value while taking 
into account a temperature and boric acid density correction. The 
change in Allowable Value does not impact the systems capability to 
perform an ECCS [emergency core cooling system] switchover from 
injection to cold leg recirculation since the nominal Trip Setpoint 
remains the same. The change in Allowable Value also will not affect 
injection or recirculation of the Containment Spray System.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed changes. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do not affect the acceptance criteria for 
any analyzed event nor is there a change in any Safety Analysis 
Limit. There will be no effect on the manner in which safety limits 
or Engineered Safety Features Actuation System settings are 
determined nor will there be any affect on those plant systems 
necessary to assure the accomplishment of protection functions. 
Therefore, there will be no impact on any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: June 11, 1999.
    Description of amendment request: The amendment would revise 
Technical Specification 3.7.1.6, ``Steam Generator Atmospheric Relief 
Valves,'' and its associated Bases to (1) require four atmospheric 
relief valves (ARVs) to be operable; (2) eliminate the use of 
``required'' in the action statements; (3) provide action statements to 
address inoperability of two ARVs and three or more ARVs due to causes 
other than excessive leakage; and (4) limit the Limiting Condition for 
Operation (LCO) 3.0.4 exception to one inoperable ARV.
    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.
    Revising the LCO to require four ARVs to be OPERABLE rather than 
three; eliminating

[[Page 35216]]

``required'' from the Actions; adding a new ACTION for three or more 
ARVs inoperable; and limiting the LCO 3.0.4 exception to one ARV 
inoperable constitute more restrictive changes from the current 
Technical Specifications. The proposed changes do not affect 
initiating mechanisms or mitigation capabilities associated with 
SGTR [steam generator tube rupture] events analyzed in Chapter 15 of 
the Updated Safety Analysis Report. The proposed changes impose more 
stringent requirements to ensure that ARV OPERABILITY is maintained 
consistent with the safety analysis and licensing basis, and also to 
address all potential single failure scenarios. Therefore these 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    With two ARVs inoperable, the allowed outage time for 
restoration of all but one ARV to OPERABLE status is changed from 24 
hours to 72 hours. The existing specification allows one valve to be 
inoperable indefinitely and with one required ARV inoperable, the 
allowed outage time for restoration is seven days. By modifying the 
LCO to require four ARVs to be OPERABLE, an allowed outage time of 
72 hours is more restrictive than the existing specification. 
Therefore, revising the allowed outage time from 24 hours to 72 
hours is acceptable based on a more restrictive allowed outage time 
from the existing specification and the low probability of an event 
requiring decay heat removal occurring during the restoration period 
that would require the ARVs. With respect to Reactor Coolant System 
cooldown for SGTR accident mitigation, the increase in time is 
acceptable based on the low probability of a SGTR event occurring 
during the restoration period and the low probability of a SGTR 
event in conjunction with the failure of the turbine bypass system 
(i.e., loss of offsite power). Therefore, this change in allowed 
outage time does not result in a significant increase in the 
probability or consequences of previously analyzed accidents.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no hardware changes nor are there any changes in the 
method by which any safety related plant system performs its safety 
function. Revising the LCO to require four ARVs to be OPERABLE 
rather than three; eliminating ``required'' from the Actions; adding 
a new ACTION for three or more ARVs inoperable; and limiting the LCO 
3.0.4 exception to one ARV inoperable will not impact the normal 
method of plant operation. The proposed changes ensure operation of 
the plant remains consistent with analysis assumptions. No new 
accident scenarios, transient precursors, failure mechanisms, or 
limiting single failures are introduced as a result of the proposed 
changes. Based on the above discussion, the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.

    The proposed changes do not affect the acceptance criteria for 
any analyzed event. There will be no effect on the manner in which 
safety limits or limiting safety system settings are determined nor 
will there be any affect on those plant systems necessary to assure 
the accomplishment of protection functions. The proposed changes 
ensure operation of the plant consistent with the analysis 
assumptions. Therefore, there will be no impact on any margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Section Chief: Stephen Dembek.

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

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

    Date of amendment request: May 24, 1999.
    Description of amendment request: Clarify nonconservative wording 
of Technical Specification (TS) 3/4,5,1, ``Safety Injection Tanks,'' 
and revise TS 3/4.5.2, ``ECCS Subsystems--Tavg Greater Than or Equal to 
325 degrees F,'' to align their associated surveillance requirements 
with the intent and design bases requirements intended to be verified.
    Date of publication of individual notice in the Federal Register: 
June 10, 1999 (64 FR 31322).
    Expiration date of individual notice: June 25, 1999.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

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

    Date of application for amendment: April 12, 1999.
    Brief description of amendment: The amendment is a temporary 
amendment change effective until September 30, 1999, which revises 
Technical Specification 3.7.8, ``Ultimate Heat Sink (UHS),'' to permit 
an 8-hour delay in the UHS temperature restoration period prior to 
entering the plant shutdown required actions.
    Date of issuance: June 4, 1999.
    Effective date: June 4, 1999.
    Amendment No.: 183.
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24193).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 4, 1999.

[[Page 35217]]

    No significant hazards consideration comments received: No.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.

Commonwealth Edison Company, Docket No. 50-249, Dresden Nuclear Power 
Station, Unit 3, Grundy County, Illinois

    Date of application for amendment: May 5, 1999.
    Brief description of amendment: The amendment removes the safety 
valve function of the Target Rock safety/relief valve from Technical 
Specifications (TS) Section 3.6.E and moves the reactor coolant system 
safety valve lift pressure setpoints from TS Section 3.6.E to TS 
Section 4.6.E.
    Date of issuance: June 4, 1999.
    Effective date: As of the date of issuance and shall be effective 
within 30 days from the date of issuance.
    Amendment No.: 168.
    Facility Operating License No. DPR-25: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 21, 1999 (64 FR 
27824).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 4, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.

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

    Date of application for amendment: December 7, 1998, as 
supplemented May 12, 1999.
    Brief description of amendment: The amendment revises Technical 
Specification 4.13A.2.a. to allow a one-time extension of the steam 
generator (SG) inspection interval. In addition, the amendment would 
remove the requirement of receiving NRC concurrence on the proposed SG 
examination program in TS 4.13C.1.
    Date of issuance: June 9, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 201.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 10, 1999 (64 
FR 6694).
    The May 12, 1999, supplemental letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 9, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: August 29, 1996, as supplemented January 
8, 1998.
    Brief description of amendment: The proposed changes revise 
requirements prescribed in Technical Specification Surveillance 
Requirement 3.3.1.1.8 and allow River Bend to increase the interval 
between whole core traversing in-core probe to local power range 
monitor calibrations from 1,000 megawatt days per ton (MWD/T) to 2,000 
MWD/T.
    Date of issuance: June 11, 1999.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment No.: 107.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 23, 1996 (61 FR 
55032).
    The January 8, 1998, letter provided additional information that 
did not change the scope of the original application and the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 11, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803.

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

    Date of amendment request: April 30, 1998.
    Brief description of amendment: The amendment revises the 
definition of quadrant power tilt to clearly allow the use of either 
the incore detectors or the excore detectors for determining quadrant 
power tilt.
    Date of issuance: June 10, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 197.
    Facility Operating License No. DPR-51: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 10, 1999 (64 
FR 6694).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 10, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

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

    Date of application for amendment: April 9, 1999.
    Brief description of amendment: The proposed amendment modifies the 
Technical Specifications (TSs) to add Limiting Condition for Operation 
3.0.6 and its associated Bases. This change allows equipment that has 
been removed from service or declared inoperable in compliance with the 
TS Action statement to be returned to service under administrative 
controls solely to perform testing required to demonstrate its 
operability or the operability of other equipment. The proposed change 
is consistent with TS 3.0.5 as discussed in NUREG-1432, Revision 1, 
``Standard Technical Specifications for Combustion Engineering 
Plants.'' TS 3.0.2 is also modified to reflect that TS 3.0.6 is an 
exception to TS 3.0.2.
    Date of issuance: June 7, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance: June 7, 1999.
    Amendment No.: 207.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24196).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 7, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

[[Page 35218]]

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3 (Waterford 3), St. Charles Parish, Louisiana

    Date of amendment request: October 1, 1998, as supplemented by 
letters dated March 25 and May 6, 1999.
    Brief description of amendment: The amendment modifies Technical 
Specification (TS) 3.3.3.7.3 and Surveillance Requirement 4.3.3.7.3 for 
the broad range gas detection system at Waterford 3. In addition, TS 
Bases 3/4.3.3.7 has been changed to reflect the new system.
    Date of issuance: June 3, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 151.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 64114).
    The March 25 and May 6, 1999, letters provided clarifying 
information that did not change the scope of the original application 
and the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: January 25, 1999, as supplemented by 
letter dated April 16, 1999.
    Brief description of amendment: The amendment removes certain 
administrative controls from the Waterford 3 Technical Specifications 
and instead relies on the requirements of the new Entergy common 
Quality Assurance Program Manual and the change controls of Title 10 of 
the Code of Federal Regulations, Section 50.54(a).
    Date of issuance: June 16, 1999.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 152.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 26, 1999 (64 
FR 9192).
    The April 16, 1999, letter provided clarifying information that did 
not change the scope of the original application and expand the initial 
proposed no significant hazards consideration determination as 
published in the Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 16, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

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

    Date of application for amendment: March 9, 1999.
    Brief description of amendment: This amendment modifies the 
Technical Specifications to increase the inservice inspection interval, 
and reduces the scope of volumetric and surface examinations for the 
reactor coolant pump flywheels.
    Date of issuance: June 8, 1999.
    Effective date: June 8, 1999.
    Amendment No.: 232.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24196).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.

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

    Date of application for amendment: September 30, 1998.
    Brief description of amendment: The amendment corrected the 
description of the reactor coolant system leakage detection capability 
of the reactor building atmosphere gaseous radioactivity monitor in the 
Improved Technical Specification Bases and the Final Safety Analysis 
Report.
    Date of issuance: June 14, 1999.
    Effective date: June 14, 1999.
    Amendment No.: 179.
    Facility Operating License No. DPR-31: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 64116).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 14, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal River, Florida 34428.

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

    Date of application for amendment: December 3, 1996.
    Brief description of amendment: The amendment incorporates certain 
improvements from the Standard Technical Specifications for Babcock and 
Wilcox plants (NUREG-1430).
    Date of issuance: June 15, 1999.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 211.
    Facility Operating License No. DPR-50: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 18, 1996 (61 
FR 66708).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 15, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (Regional Depository) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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

    Date of application for amendment: January 22, 1999.
    Brief description of amendment: Revises Technical Specification 
(TS) Section 4.3, ``Fuel Storage,'' by updating the criticality 
requirements (k-infinity and U-235 enrichment limits) for storage of 
fuel assemblies in the spent fuel racks. This change would allow for 
storage of nuclear fuel assemblies with new designs, including GE-12 
with a 10X10 pin array.
    Date of issuance: June 8, 1999.
    Effective date: June 8, 1999.
    Amendment No.: 226.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1999 (64 
FR 9192).

[[Page 35219]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, IA 52401.

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

    Date of application for amendment: October 15, 1998, as 
supplemented on December 21, 1998.
    Brief description of amendment: Revise the Technical Specifications 
(TS) by adding a new TS 3.7.9, ``Control Building/Standby Gas Treatment 
System Instrument Air System,'' and revises (TS) 3.6.1.3, ``Primary 
Containment Isolation Valves,'' Condition E.
    Date of issuance: June 9, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 227.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1999 
(64FR9193).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 9, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, IA 52401.

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

    Date of application for amendments: April 19, 1999.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3/4.8.1.2, ``Electrical Power Systems, Shutdown,'' 
and its associated bases to provide a one-time extension of the 18-
month surveillance interval for specific surveillance requirements 
associated with the emergency diesel generators for Units 1 and 2. The 
surveillances will be performed prior to the first entry into Mode 4 
following the current plant shutdown. In addition, for Unit 2 only, a 
minor administrative change is included to delete a reference to TS 
4.0.8, which is no longer applicable. For Unit 1 only, an editorial 
change is made to add the word ``or'' to action statement 3.8.1.2.
    Date of issuance: June 8, 1999.
    Effective date: June 8, 1999, with full implementation within 45 
days.
    Amendment Nos.: 228 and 211.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 29, 1999 (64 FR 
23129).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, MI 49085.

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

    Date of application for amendment: May 15, 1998, as supplemented by 
letters dated September 25, October 13, December 9 (two letters), 1998; 
January 11, April 1, and April 22, 1999.
    Brief description of amendment: This amendment changes Technical 
Specification (TS) 5.5, ``Storage of Unirradiated and Spent Fuel,'' to 
reflect a planned modification to increase the storage capacity of the 
spent fuel pool from 2776 to 4086 fuel assemblies. It also deletes an 
inappropriate statement and reference within TS 5.5.
    Date of issuance: June 17, 1999.
    Effective date: This license amendment is effective as of the date 
of its issuance to be implemented before spent fuel is stored within 
the new high-density spent fuel rack modules authorized for 
installation and use by this amendment.
    Amendment No.: 167.
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1998 (63 
FR 64973).
    The September 25, October 13, December 9 (two letters) 1998, 
January 11, April 1, and April 22, 1999, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 17, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendments: January 21, 1999, which 
superseded application dated July 22, 1998.
    Brief description of amendments: The amendments revise the 
Technical Specifications high radiation trip setpoints for the reactor 
building and the refueling floor ventilation exhaust monitors.
    Date of issuance: June 9, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--216; Unit 2--157.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24200); this supersedes the original notice dated August 26, 1998 (63 
FR 45529).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 9, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia.

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: March 30, 1999.
    Brief description of amendments: The amendments deleted Technical 
Specification 3/4.3.3.4, ``Meteorological Instrumentation,'' and its 
associated Bases. These requirements have already been relocated to the 
Technical Requirements Manual (TRM). Because the TRM is incorporated 
within the South Texas Project updated final safety analysis report for 
the units, changes to the relocated requirements will be controlled by 
10 CFR 50.59.
    Date of issuance: June 16, 1999.
    Effective date: June 16, 1999, to be implemented within 30 days.
    Amendment Nos.: Unit 1--111; Unit 2--98.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24201).

[[Page 35220]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 16, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.

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

    Date of application for amendments: February 16, 1999.
    Brief Description of amendments: The amendments revise Technical 
Specifications (TS) Sections 3.6, 3.9, and 3.16 and the associated 
Bases for those sections for Units 1 and 2. The changes consolidate the 
auxiliary feedwater cross-connect requirements by relocating the 
electrical power requirements from Section 3.16 to Section 3.6. The TS 
are also clarified with regard to permitting simultaneous entry into 
certain conditions of operation on Units 1 and 2.
    Date of issuance: June 7, 1999.
    Effective date: June 7, 1999.
    Amendment Nos.: 220 and 220.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24203).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 7, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By July 30, 1999, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the

[[Page 35221]]

results of the proceeding. The petition should specifically explain the 
reasons why intervention should be permitted with particular reference 
to the following factors: (1) the nature of the petitioner's right 
under the Act to be made a party to the proceeding; (2) the nature and 
extent of the petitioner's property, financial, or other interest in 
the proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

    Date of application for amendments: June 10, 1999.
    Brief description of amendments: The amendments revised the 
Technical Specifications TS 3.7.9, ``Control Room Area Ventilation 
System (CRAVS),'' to establish actions to be taken for an inoperable 
control room ventilation system due to a degraded control room pressure 
boundary. This revision approves a one-time-only action for two CRAVS 
trains inoperable due to a degraded control room boundary in Modes 1, 
2, 3, and 4, that is to be completed within 24 hours. The applicable TS 
Bases have been revised to document the TS changes and to provide 
supporting information.
    Date of issuance: June 11, 1999.
    Effective date: As of the date of issuance and shall be implemented 
upon receipt.
    Amendment Nos.: Unit 1--185; Unit 2--167.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    The Commission's related evaluation and the amendment, finding of 
emergency circumstances, consultation with the State of North Carolina, 
and final no significant hazards consideration determination are 
contained in a Safety Evaluation dated June 11, 1999.
    Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina.
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, Charlotte, North Carolina.
    NRC Section Chief: Richard L. Emch, Jr.

    Dated at Rockville, Maryland, this 23rd day of June 1999.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 99-16489 Filed 6-29-99; 8:45 am]
BILLING CODE 7590-01-P