[Federal Register Volume 63, Number 145 (Wednesday, July 29, 1998)]
[Notices]
[Pages 40551-40567]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-20111]


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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. Pub. L. 97-415 revised section 189 of the Atomic Energy Act of 
1954, as amended (the Act), to require the Commission to publish notice 
of any amendments issued, or proposed to be issued, under a new 
provision of section 189 of the Act. This provision grants the 
Commission the authority to issue and make immediately effective any 
amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from July 3, 1998, through July 17, 1998. The 
last biweekly notice was published on July 15, 1998 (63 FR 38198).

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

[[Page 40552]]

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 August 14, 1998, 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: 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

[[Page 40553]]

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.

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

    Date of amendment request: May 27, 1997, as supplemented by letters 
dated March 9, March 20, April 20, May 27, and June 24, 1998
    Description of amendment request: The proposed amendments would 
revise the current Technical Specifications (TS) of each unit to 
conform with NUREG-1431, Revision 1, ``Standard Technical 
Specifications--Westinghouse Plants.'' The staff had previously issued 
a Notice of Consideration of Issuance of Amendments published in the 
Federal Register on July 14, 1997 (62 FR 37628) covering all the 
proposed changes that were indeed within the scope of NUREG-1431. The 
staff subsequently published two Notices of Consideration of Issuance 
of Amendments and Proposed No Significant Hazards Determination (63 FR 
25106, dated May 6, 1998; 63 FR 27760 dated May 20, 1998) to cover 
DEC's March 9, March 20, April 20, and May 27, 1998, supplements, which 
proposed changes that are beyond the scope of NUREG-1431. On June 24, 
1998, DEC identified additional beyond-scope changes. The following 
descriptions and proposed no significant hazard analyses cover only 
those beyond-scope changes. Associated with each change are 
administrative/editorial changes such that the new or revised 
requirements would fit into the format of NUREG-1431.
    1. Current TS 4.8.1.1.2.f specifies that the fuel for the 
emergency diesel generators (EDGs) be periodically sampled for 
particulate contamination strictly in accordance with the industry 
standard ASTM-D2276-78. DEC proposed to relax this requirement, 
adopting only the guidance of the standard, but using a larger 
particulate filter for sampling (change from 0.8-to 3-micron). The 
revised requirement would show up as TS 5.5.13.c of the Improved TS. 
No changes to the design and functions of the EDGs are proposed.
    2. DEC proposed to revise current TS Table 4.3-1, Functions 16 and 
17. The revised requirements, to show up as Table 3.3.1-1, Functions 15 
and 16.b, of the Improved TS, would add an actuation logic test 
surveillance for the reactor trip system interlocks and the safety 
injection input from the engineered safety feature actuation system. No 
changes to the design and functions of these systems are involved.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), DEC has provided its 
analyses of the issue of no significant hazards consideration for each 
of the above proposed changes. The NRC staff has reviewed DEC's 
analyses against the standards of 10 CFR 50.92(c). The NRC staff's 
analysis is presented below.
    1. Will the changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    For all the changes the answer is ``no.'' The proposed changes will 
not affect the safety function of the subject systems. There will be no 
direct effect on the design or operation of any plant structures, 
systems, or components. No previously analyzed accidents were initiated 
by the functions of these systems, and the systems will continue to 
perform their functions in mitigating consequences of previously 
analyzed accidents. Therefore, the proposed changes will have no impact 
on the consequences or probabilities of any previously evaluated 
accidents.
    2. Will the changes create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    For all the changes the answer is ``no.'' The proposed changes 
would not lead to any design or operating procedure change. Hence, no 
new equipment failure modes or accidents from those previously 
evaluated will be created.
    3. Will the changes involve a significant reduction in a margin of 
safety?
    For all the changes the answer is ``no.'' Margin of safety is 
associated with confidence in the design and operation of the plant. 
The proposed changes to the TS 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 each of the proposed changes. Therefore, 
the NRC staff proposes to determine that the amendment request involves 
no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.
    Attorney for licensee: Mr. Paul R. Newton, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: September 15, 1997, as supplemented by 
letters dated March 5, April 27, and June 15, 1998.
    Description of amendment request: The staff had previously 
published a Notice of Consideration of Amendments and Proposed No 
Significant Hazards Consideration Determination on the licensee's 
September 15, 1997, application in the Federal Register on October 8, 
1997 (62 FR 52580). As a result of the staff's requests for additional 
information, DEC expanded its original amendment application by letter 
dated June 15, 1998. Specifically, the June 15, 1998, letter proposes 
requirements regarding the Low Temperature Overpressure Protection 
System to be added to the units' Technical Specifications. There is, 
however, no change to 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, addressing the three standards of 10 CFR 50.92(c):

First Standard

    Implementation of this amendment would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The Low Temperature Overpressure Protection 
System is not an accident initiating system; it is an accident 
mitigating system. Therefore, the addition of supplemental Technical 
Specification required controls pertaining to this system cannot 
impact accident initiating probabilities. The Low Temperature 
Overpressure Protection System will remain fully capable of 
performing its design accident mitigation function for the modes in 
which it is required. Therefore, no accident consequences will be 
impacted.

Second Standard

    Implementation of this amendment would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. As noted previously, the Low Temperature 
Overpressure Protection System is not an accident initiating system. 
The addition of the supplemental Technical Specification

[[Page 40554]]

controls pertaining to this system as specified will not impact any 
plant systems that are accident initiators. No other modifications 
are being proposed to the plant which would result in the creation 
of new accident mechanisms.

Third Standard

    Implementation of this amendment would not involve a significant 
reduction in a margin of safety. Margin of safety is related to the 
confidence in the ability of the fission product barriers to perform 
their design functions during and following an accident situation. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment system. The performance of the fission 
product barriers will not be impacted by implementation of this 
proposed amendment supplement. The Low Temperature Overpressure 
Protection System will remain fully capable of performing its design 
function for the modes in which it is required. Therefore, no safety 
margin will be significantly impacted.

    The staff reviewed the licensee's analysis, and agrees that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.
    Attorney for licensee: Mr. Paul R. Newton, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina.
    NRC Project Director: Herbert N. Berkow.

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: May 8, 1998.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) for the Power Range Neutron 
Flux High Trip setpoints in the event of inoperable main steam safety 
valves. The licensee has determined that the new values are more 
conservative than the values in the current TS. Also, the proposed 
changes would delete the references to the 3-loop operation. The 
proposed changes are consistent with the proposed Improved Standard TS 
submitted by the licensee on May 27, 1997.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with the 
proposed amendment involve an increase in the probability or 
consequences of an accident previously evaluated?
    The proposed amendment involves a reduction in the maximum 
allowable power range neutron flux high setpoints in case of 
inoperable main steam safety valves. All applicable UFSAR [Updated 
Final Safety Analysis Report] Chapter 15 transient acceptance 
criteria are met with the proposed change. Therefore, operation of 
the facility in accordance with the proposed amendment will not 
involve an increase in the probability or consequences of an 
accident previously evaluated.
    2. Will operation of the facility in accordance with the 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated:
    No new equipment or operating practice is involved with this 
proposed amendment. No alteration to any existing hardware is 
involved with this proposed amendment. Power Range high neutron flux 
setpoint calibration is continued to be performed by the same 
approved procedure. Therefore, operation of the facility in 
accordance with the proposed amendment will not create the 
possibility of any new or different kind of accident from any 
accident previously evaluated.
    3. Will operation of the facility in accordance with the 
proposed amendment involve a reduction in a margin of safety?
    The proposed change is in a more-conservative direction. All 
applicable UFSAR Chapter 15 transient acceptance criteria are met 
with the proposed amendment. Therefore, operation of the facility in 
accordance with the proposed amendment will not involve a reduction 
in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    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 Project Director: Herbert N. Berkow.

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

    Date of amendment request: May 27, 1997, as supplemented by letters 
dated March 9, March 20, April 20, May 27, June 3, June 24, and July 7, 
1998.
    Description of amendment request: The proposed amendments would 
revise the current Technical Specifications (TS) of each unit to 
conform with NUREG-1431, Revision 1, ``Standard Technical 
Specifications--Westinghouse Plants.'' The staff had previously issued 
a Notice of Consideration of Issuance of Amendments published in the 
Federal Register on July 15, 1997 (62 FR 37940) covering all the 
proposed changes that were indeed within the scope of NUREG-1431. The 
staff subsequently published additional Notices of Consideration of 
Issuance of Amendments and Proposed No Significant Hazards 
Determination on May 6, 1998 (63 FR 25107 and 63 FR 25108 (two 
notices)) and on May 20, 1998 (63 FR 27761) to cover DEC's March 9, 
March 20, April 20, and May 27, 1998, supplements, which proposed 
changes that are beyond the scope of NUREG-1431.
    On June 24, 1998, DEC identified additional beyond-scope changes. 
The following descriptions and proposed no significant hazard analyses 
cover only those beyond-scope changes. Associated with each change are 
administrative/editorial changes such that the new or revised 
requirements would fit into the format of NUREG-1431.
    1. Current TS 4.8.1.1.2.f specifies that the fuel for the emergency 
diesel generators (EDGs) be periodically sampled for particulate 
contamination in accordance with ASTM-D2276-78. DEC proposed to relax 
this requirement, adopting instead the guidance of ASTM-D2276, Method 
A. The revised requirement would show up as TS 5.5.13.c of the Improved 
TS. No changes to the design and functions of the EDGs are proposed.
    2. DEC proposed to change the required action due to inoperable 
channels of the containment pressure control system as currently 
contained in Table 3.3-3, Item 7. The revised requirement would show up 
as Action Item 16b in Table 3.3.2-1 of the Improved TS. No changes to 
the design and functions of the containment pressure control system are 
involved.
    3. DEC proposed to revise current TS Table 4.3-1, Functions 16 and 
17. The revised requirements, to show up as Table 3.3.1-1 Functions 15 
and 16.b, would add an actuation logic test surveillance for the 
reactor trip system interlocks and the safety injection input from the 
engineered safety feature actuation system. No changes to the design 
and functions of these systems are involved.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 40555]]

As required by 10 CFR 50.91(a), DEC has provided its analyses of the 
issue of no significant hazards consideration for each of the above 
proposed changes. The NRC staff has reviewed DEC's analyses against the 
standards of 10 CFR 50.92(c). The NRC staff's analysis is presented 
below.
    1. Will the changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    For all the changes the answer is ``no.'' The proposed changes will 
not affect the safety function of the subject systems. There will be no 
direct effect on the design or operation of any plant structures, 
systems, or components. No previously analyzed accidents were initiated 
by the functions of these systems, and the systems will continue to 
perform their functions in mitigating consequences of previously 
analyzed accidents. Therefore, the proposed changes will have no impact 
on the consequences or probabilities of any previously evaluated 
accidents.
    2. Will the changes create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    For all the changes the answer is ``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.
    3. Will the changes involve a significant reduction in a margin of 
safety?
    For all the changes the answer is ``no.'' Margin of safety is 
associated with confidence in the design and operation of the plant. 
The proposed changes to the TS 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 each of the proposed changes. Therefore, 
the NRC staff proposes to determine that the amendment request involves 
no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.
    Attorney for licensee: Mr. Paul R. Newton, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina.
    NRC Project Director: Herbert N. Berkow.

Duke Energy Corporation, Docket No. 50-287, Oconee Nuclear Station, 
Unit 3, Oconee County, South Carolina

    Date of amendment request: July 16, 1998.
    Description of amendment request: The proposed change would extend, 
on a one-time basis, certain specified Technical Specification 
surveillances that are required to be performed at a frequency of 18 
months from the maximum allowed frequency of 22 months, 15 days, to a 
maximum of 24 months. The following surveillances are involved: (a) 
Standby Shutdown Facility (SSF) Reactor Coolant System (RCS) Pressure 
Instrument Calibration; (b) SSF RCS Pressurizer Level Instrument 
Calibration; (c) SSF RCS Makeup Pump Flow Instrument Calibration; (d) 
Reactor Protective System (RPS) RCS Flow Instrument Calibration; (e) 
RPS RCS Pressure Instrument Calibration; and (f) Low Pressure Injection 
System Pump Discharge Valves LP-17 and LP-18 Manual Cycle.
    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:

    This proposed change has been evaluated against the standards in 
10 CFR 50.92 and has been determined to involve no significant 
hazards, in 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?
    No. A review of the previous two instrument channel tests and 
calibrations, and two manual valve cycle tests discussed in this 
amendment request concluded that no adverse effects should occur as 
a result of the one-time extension.
    There is a high level of confidence that the instruments and 
valves should be available to perform their intended function during 
the requested extension period. Thus, the probability and 
consequences of an accident previously evaluated will not be 
significantly increased.
    (2) Create the possibility of a new or different kind of 
accident from the accidents previously evaluated?
    No. Since the one-time extension should not cause any adverse 
effects on Standby Shutdown Facility, Reactor Protective System or 
the Low Pressure Injection system, a new or different kind of 
accident from the accidents which were previously evaluated will not 
occur. The Standby Shutdown Facility, Reactor Protective System or 
the Low Pressure Injection system should be available to perform 
their intended function during the requested extension period.
    (3) Involve a significant reduction in a margin of safety?
    No. The margin of safety will not be significantly reduced by 
this amendment request because the Standby Shutdown Facility, 
Reactor Protective System or the Low Pressure Injection system 
should be available to perform their intended function during the 
requested extension period. In addition, the review of the previous 
tests and calibrations which are discussed in the amendment request 
concluded that no adverse effects should occur as a result of the 
one-time extension.
    Duke [Energy Corporation] has concluded, based on the above 
information, that there are no significant hazards involved in this 
amendment request.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: April 28, 1998.
    Description of amendment request: The proposed amendment would 
change the scope and frequency of volumetric and surface inspections 
for the reactor coolant pump motor flywheels. The current prescribed 
frequency and scope are contained in U.S. NRC Regulatory Guide 1.14, 
Regulatory Positions C.4.b.1 and C.4.b.2. The proposed revision 
reflects the frequency and scope of volumetric and surface 
examinations, which has been reviewed and approved by the NRC, as 
stated in the Safety Evaluation for Topical Report WCAP-14535A.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The CR-3 [Crystal River Unit 3] components addressed by this 
proposed change are the Reactor Coolant Pumps (RCPs), identified by 
plant tagging procedures as RCP-1A, RCP-1B, RCP-1C,

[[Page 40556]]

and RCP-1D. The RCPs are vertical, single stage, single suction, 
shaft seal, centrifugal pumps. The RCPs ensure that adequate cooling 
water is circulated through the reactor coolant system. Following 
loss of power to the RCP motor, the flywheel, in conjunction with 
the impeller and motor rotating assembly, provide sufficient 
rotational inertia to assure adequate coolant flow during RCP 
coastdown, thus providing adequate core cooling. The maximum loading 
on the RCP motor flywheel results from overspeed following a large 
loss of coolant accident (LOCA). The estimated maximum speed in the 
event of a LOCA was established conservatively. The proposed change 
does not affect that analysis. Reduced coastdown times due to a 
single failed flywheel is bounded by the locked rotor analysis, 
therefore it will not place the plant in an unanalyzed condition.
    Reducing the frequency of inspection, as proposed, will not 
significantly increase the probability of an accident previously 
evaluated. CR-3 is not specifically analyzed for a flywheel failure 
accident. The design, fabrication, and testing of the flywheels in 
accordance with the guidance found in Regulatory Guide 1.14 
minimizes the potential for flywheel failure. Nevertheless, the 
topical report indicates that the flywheels could be operated for 
forty years without inspection, and there would be no significant 
increase in the probability of failure of the flywheel. However, 
inspections are proposed to continue at a frequency of once every 
ten years as a conservative measure. Therefore, these 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.
    The purpose of the RCP motor flywheel inspection is to identify 
flaws that could lead to failure of the flywheel. The design, 
fabrication, and testing of the flywheels in accordance with the 
guidance found in Regulatory Guide 1.14 minimizes the potential for 
flywheel failure. No new failure mode is introduced due to the 
change in flywheel inspection frequency since the proposed changes 
do not involve the addition or modification of equipment, nor alter 
the design or operation of affected plant systems, structures or 
components. Therefore, these changes do not create a possibility of 
a new or different kind of accident from any previously evaluated.
    (3) Involve a significant reduction in a margin of safety.
    As shown in the topical report, RCP motor flywheels have been 
inspected for twenty years without any service induced flaws being 
identified. Additionally, the analyses demonstrated that the 
flywheels are manufactured from excellent quality steel, have a high 
fracture toughness, and have a very high flaw tolerance. The topical 
report indicates that the flywheels could be operated for forty 
years without inspection, and there would be no significant increase 
in the probability of failure of the flywheels. However, inspections 
are proposed to continue at a frequency of once every ten years as a 
conservative measure. The non-destructive examination acceptance 
criteria is not changing as a result of the proposed LAR. Thus, the 
margin of safety is not reduced significantly by the proposed change 
in inspection frequency.

    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: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC--A5A, P. O. Box 14042, St. Petersburg, Florida 
33733-4042.
    NRC Project Director: Frederick J. Hebdon.

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 27, 1998.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TS) to remove the requirement for 
safety injection tanks (SITs) to be operable in reactor operational 
Mode 4.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) 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 amendment does not involve changes to previously 
evaluated accident initiators. The proposed TS changes related to 
removal of the requirement for safety injection tanks to be operable 
in MODE 4 do not impact the results of existing accident analyses, 
and have no adverse impact on any plant system performance.
    The function of each SIT is to provide early reactor core 
reflood in the event of a LBLOCA [large break loss-of-coolant 
accident]. Safety injection tanks are not required for mitigating 
the consequences of large RCS pipe ruptures in MODE 4, and the 
proposed change to TS 3.5.1 will delete the requirement for SIT 
operability when in this mode. Due to the reduced initial stored 
energy and decay heat generation rate consistent with operation in 
the shutdown modes, the required operable HPSI [high-pressure safety 
injection] pump is sufficient to perform the function of reactor 
vessel reflood and coolant inventory make-up. Therefore, operation 
of the facility in accordance with the proposed amendment will 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 amendment will not change the physical plant or the 
modes of operation defined in the facility license. The changes do 
not involve the addition of new equipment or the modification of 
existing equipment, nor do they alter the design of St. Lucie plant 
systems described in the Updated Final Safety Analysis Report 
(UFSAR). There are no adverse effects on any system performance due 
to the proposed TS changes, and the plant configuration will 
continue to remain consistent with assumptions used in the existing 
accident analyses. Therefore, 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.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed TS changes have been evaluated with respect to the 
applicable safety analyses. FPL [Florida Power and Light Co.] 
determined from this new evaluation that safety injection tanks are 
not required to prevent core uncovery during a loss of coolant 
accident initiated in MODE 4. Due to the reduced core heat removal 
requirements in this lower mode and in the absence of substantial 
core uncovery, fuel cladding temperatures and clad oxidation will 
remain at low levels, long term cooling will be maintained, and 10 
CFR 50.46 acceptance criteria will be satisfied. Therefore, 
operation of the facility in accordance with the proposed amendment 
would not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Indian River Community College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: June 29, 1998.
    Description of amendment request: This Technical Specification 
change

[[Page 40557]]

request replaces in its entirety, a previously submitted request dated 
February 22, 1996, and published in the Federal Register on March 27, 
1996 (61 FR 13525). This request greatly reduces the scope of the 
previous request. It retains the provision to delete the requirement 
that the biennial inspection of the Emergency Diesel Generators (EDGs) 
be performed during shutdown, permits skipping diesel starting battery 
capacity test for recently installed batteries, and increases the 
minimum loading during diesel testing from 20% to 80%. In addition, 
there are wording changes to enhance clarity, and a typographical error 
is corrected.
    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. State the basis for the determination that the proposed 
activity will or will not increase the probability of occurrence or 
the consequences of an accident.
    The proposed activity deletes the requirement to inspect EDGs 
during shut down from the Technical Specifications and permits 
skipping diesel starting battery capacity tests of recently 
installed batteries. The minimum loading during the testing of the 
diesels has been increased from 20% to 80%. In addition, wording 
changes were made to enhance clarity and a minor typographical error 
was corrected. During reactor operations other power sources are 
available to compensate for one diesel being out of service. The 
inspections and testing will continue to be done with the same 
intervals and the 80% loading is a more stringent requirement. 
Therefore, these changes do not affect the design or performance of 
the EDGs or their ability to perform their design function.
    2. State the basis for the determination that the activity does 
or does not create a possibility of an accident or malfunction of a 
different type than any previously identified in the [safety 
analysis report] SAR.
    The EDGs are not the source of any accident described in the 
SAR. These changes do not modify the design or performance of the 
EDGs and do not affect plant functions or actions. Current 
specifications permit one diesel generator to be inoperable for up 
to 7 days and this change will not impact that time frame. 
Therefore, the proposed change does not create the possibility of an 
accident or malfunction of a different type than those previously 
identified.
    3. State the basis for the determination that the margin of 
safety is not reduced.
    The proposed changes are designed to improve EDG reliability and 
availability during shutdown periods by providing flexibility in the 
scheduling and performance of maintenance. The surveillance 
intervals are unchanged and operability requirements are not 
modified. The proposed activity does not alter the basis of any 
technical specification that is related to the establishment or 
maintenance of a nuclear safety margin. Therefore, the margin of 
safety is not significantly reduced by this action.

    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: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cecil O. Thomas.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point 
Nuclear Station, Unit 1 (NMP1), Oswego County, New York

    Date of amendment request: June 19, 1998.
    Description of amendment request: The proposed amendment would 
update Technical Specification (TS) 3.2.2, ``Minimum Reactor Vessel 
Temperature for Pressurization,'' and the associated TS Bases pages. TS 
3.2.2 contains tables and figures that limit the minimum reactor vessel 
temperature for a given pressure. The limits are based upon the number 
of Effective Full Power Years (EFPY) of core operation. The current 
tables and figures are valid for up to 18 EFPYs of core operation. The 
proposed amendment will substitute new tables and figures that are 
valid for 20, 24 and 28 EFPYs. The word ``leakage'' would be added to 
clarify that this TS applies to both leakage and hydrostatic tests.
    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 Nine Mile Point Unit 1, in accordance with 
the proposed amendment, will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The changes to the P-T [pressure and temperature] curves are 
being proposed to preclude brittle fracture of RPV [reactor pressure 
vessel] materials for up to 28 EFPYs. In addition to the leakage/
hydrostatic test curve for 28 EFPYs, leakage/hydrostatic test curves 
have been prepared for exposures up to 20 EFPYs and up to 24 EFPYs 
to shorten outage time for startups conducted prior to these 
exposures. Safety margins specified in 10 CFR Part 50, Appendix G 
and Appendix G to Section III of the ASME [American Society of 
Mechanical Engineers] Code will continue to be met for each of these 
curves. Also, the proposed changes do not affect the probability of 
any accident precursors. Therefore, operation in accordance with the 
proposed change will not involve a significant increase in the 
probability of an accident previously evaluated.
    The RPV, as part of the reactor coolant system, provides a 
barrier to the release of reactor coolant and subsequent 
radiological consequences. Operation in accordance with the proposed 
amendment will preclude brittle fracture of the RPV consistent with 
current requirements, and consequently, not affect the consequences 
of any accidents. Therefore, operation of NMP1 [Nine Mile Point Unit 
1] in accordance with the proposed amendment will not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    2. The operation of Nine Mile Point Unit 1, 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 does not involve any physical alterations to 
plant configurations or introduce any new accident precursors which 
could initiate a new or different kind of accident. The proposed 
change does not affect the intended function of the RPV nor does it 
affect the operation of the RPV in a way which would create a new or 
different kind of accident. The changes to the P-T curves are being 
proposed to preclude brittle fracture of RPV materials for up to 28 
EFPYs. Safety margins specified in 10 CFR Part 50, Appendix G and 
Appendix G to Section III of the ASME Code will continue to be met. 
Therefore, operation of NMPI in accordance with the proposed change 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The operation of Nine Mile Point Unit 1, in accordance with 
the proposed amendment, will not involve a significant reduction in 
a margin of safety.
    The existing NMP1 P-T curves were developed using safety margins 
for brittle fracture found in 10 CFR PART 50 Appendix G and Appendix 
G to Section III of the ASME Code. The proposed NMP1 P-T operation 
curves, which are valid for up to 28 EFPYs of operation, were also 
developed using the safety margins for brittle fracture found in 10 
CFR PART 50, Appendix G and Appendix G to Section III of the ASME 
Code. Accordingly, operation of NMPI in accordance with the revised 
P-T operating limits will continue to preclude brittle fracture of 
the RPV materials during plant heatup, cooldown, and leakage/
hydrostatic test conditions with the same margin of safety that 
currently exists. Therefore, operation of NMP1 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

[[Page 40558]]

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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: S. Singh Bajwa.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of amendment request: May 22, 1997, as supplemented by letters 
dated June 12, 1997, August 28, 1997 and January 29, 1998.
    Description of amendment request: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant Unit Nos. 1 and 2 TS 3/4.7.3.1, ``Plant Systems--Vital 
Component Cooling Water System,'' to add new action statements and 
surveillance requirements for the component cooling water (CCW) surge 
tank pressurization system. CCW surge tank pressurization system 
requirements currently exist in an equipment control guideline, but are 
proposed for inclusion in TS because the CCW surge tank pressurization 
system is required to support licensing basis assumptions for a design 
basis loss-of-coolant accident.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, 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 component cooling water (CCW) surge tank pressurization 
system is designed to mitigate the consequences of an accident, and 
cannot initiate an accident.
    The proposed changes to the Technical Specifications (TS) 
incorporate requirements for the CCW surge tank pressurization 
system to assure that the consequences of an accident are not 
increased. The CCW surge tank pressurization system was installed to 
restore the component cooling water system to its original design 
and licensing basis. The design of the CCW surge tank pressurization 
system ensures that a minimum pressure of 17 psig is maintained in 
the surge tank at the initiation of a design basis loss of coolant 
accident. This minimum pressure is sufficient to ensure that boiling 
will not occur in the containment fan cooler units (CFCUs), assuming 
the worst case accident conditions with a concurrent loss of offsite 
power (LOOP).
    Therefore, the addition of these new requirements does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The CCW surge tank pressurization system is designed to mitigate 
the consequences of an accident, and cannot initiate an accident.
    The proposed TS changes incorporate requirements for the CCW 
surge tank pressurization system. Installation of the CCW surge tank 
pressurization system provides assurance that boiling in the CFCUs 
will not occur, assuming the worst case accident, with a concurrent 
LOOP.
    Therefore, addition of these requirements does 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.
    The proposed changes to the TS incorporate requirements for the 
CCW surge tank pressurization system to assure that the consequences 
of an accident are not increased. The design of the CCW surge tank 
pressurization system ensures that a minimum pressure of 17 psig is 
maintained in the surge tank at the initiation of a design basis 
accident. The minimum pressure is sufficient to ensure that boiling 
will not occur in the CFCUs, assuming the worst case accident 
conditions with a concurrent LOOP.
    Therefore, the proposed changes do not involve a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.
    Attorney for Licensee: Christopher J. Warner, Esq., Pacific Gas & 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Project Director: William H. Bateman.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Docket 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: March 20, 1998, as revised by 
letter dated June 26, 1998.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to permit incorporation of an 
End-of-Cycle Recirculation Pump Trip.(EOC-RPT) System at Peach Bottom 
Atomic Power Station (PBAPS), Units 2 and 3.
    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 TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The addition of the EOC-RPT System will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. The EOC-RPT System has been designed 
to appropriate standards and specifications to ensure that the 
ability of the plant to mitigate the effects of accidents is 
maintained. Each division is electrically, mechanically, and 
physically independent to meet the single failure criterion.
    The EOC-RPT System will improve the reactor core thermal 
response following a turbine trip transient caused by either a 
turbine control valve fast closure or a turbine stop valve closure. 
The EOC-RPT will be relied upon to reduce the fuel thermal 
mechanical transient excursion such that fuel thermal limits are not 
violated. Under conditions when the system is inoperable, more 
conservative thermal limits will be enforced.
    The new system will utilize existing RPS [Reactor Protection 
System] logic to initiate the Reactor Recirculation System (RRS) 
pump trips on a turbine generator trip and a generator load 
rejection event. The inputs to RPS used by EOC-RPT will be from 
turbine stop valve (TSV) limit switches and turbine control valve 
(TCV) oil pressure switches. There will be no direct interface 
between the EOC-RPT System and the main turbine control system. Thus 
the new system can not initiate a turbine trip or generator load 
rejection event. This change does not result in significant increase 
in the probability of events described in the UFSAR [Updated Final 
Safety Analysis Report]. Additionally, the probability of 
inadvertent single or dual recirculation pump trips due to the 
addition of the EOC-RPT components will not be significantly 
increased by this modification.
    No new challenges to the reactor coolant pressure boundary will 
result from the incorporation of the EOC-RPT System which

[[Page 40559]]

could result in a significant increase in the consequences of an 
accident.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The ECO-RPT System has been designed to appropriate standards 
and specifications to ensure that no new sequence of events or 
failure modes will occur such that a transient event will escalate 
into a new or different type of accident.
    The PBAPS UFSAR evaluates several recirculation pump trip 
events, including the limiting case of a pump seizure. A spurious 
dual EOC-RPT pump trip is similar to other RRS pump trip events 
evaluated in the UFSAR and does not represent a different type of 
accident.
    Additionally, this modification will not create any new failure 
mode or sequences of events that could lead to a different type of 
accident than previously evaluated. The new EOC-RPT System will not 
involve any new challenges to a fission product barrier. The EOC-RPT 
System does not make any changes to the design function of the RRS. 
Therefore, the new equipment installed by this modification cannot 
create the possibility of a different type than previously evaluated 
in the SAR [Safety Analysis Report].
    The EOC-RPT System is classified as important-to-safety. Failure 
or malfunction of the new equipment will not prevent or affect the 
ability of safety-related or important-to-safety systems to respond 
to the design basis accidents described in the FSAR [Final Safety 
Analysis Report].
    There will be no software used in the EOC-RPT System. The system 
logic consists of two electrically and physically separated trip 
systems; one will be used to trip one EOC-RPT System breaker, and 
the other will be used to trip the second EOC-RPT System breaker for 
each pump.
    The design of this modification assures that the new system is 
not susceptible to electromagnetic (EM) emissions and will not cause 
inadvertent operation of existing plant equipment due to EM 
emissions.
    Based on the previous discussion, the possibility of a new or 
different kind of accident from any accident previously evaluated 
will not be created.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    There are no significant reductions in any margin of safety 
previously approved by the USNRC as a result of this change to the 
TS. The EOC-RPT System will ensure that fuel thermal limits are not 
exceeded during the limiting transient. In the event that the EOC-
RPT System is determined to be inoperable, specific operating limits 
are provided in the COLR. In all cases, thermal limits are not 
exceeded and the margin of safety is not significantly reduced.
    The plant LOCA response will not change for present core 
configurations (i.e., 9 x 9 fuel) with the EOC-RPT System installed. 
For GE 8 x 8 fuel, which could be used at a future time, there could 
be a small increase in Peak Cladding Temperature (PCT). This 
increase would still be well below the 2200 deg. F acceptance limit 
defined in 10 CFR 50.46.
    There will be no significant reduction in the margin of safety 
as previously approved by the USNRC, since the calculated increase 
in peak cladding temperature for a core containing limit 8x8 fuel 
design (BP/P8 x 8R) is a small increase above the previously 
analyzed peak cladding temperature. Additionally, this modification 
does not impact the safety function of the RRS piping, thus reactor 
coolant pressure boundary safety limits are not affected.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.
    Attorney for Licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101.
    NRC Project Director: Robert A. Capra.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Dockets 
Nos. 50-277 and 50-278, PeachBottom Atomic Power Station, Units Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: May 1, 1998
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to delete requirements for the 
functional testing of the safety relief valves (SRVs) during each unit 
startup at Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3.
    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 TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed Technical Specification changes to the requirement 
for functional testing of the SRVs during each unit startup will not 
significantly increase the probability or consequences of an 
accident previously evaluated. Elimination of the functional test 
will not prevent the SRVs from performing their intended safety 
function. The proposed change to delete the SRV functional test at 
power should delete a potential initiator of SRV leakage. The 
remaining testing and inspections will continue to adequately 
demonstrate the operability of the SRVs for both the safety and 
depressurization modes.
    As a result of deleting the requirement for functional testing 
of the SRVs during each unit startup and replacing these 
requirements with the proposed tests contained TS SR [surveillance 
requirement] 3.4.3.2 and 3.5.1.12, the only change in the frequency 
of testing of the SRV components is that the main valve disc of the 
SRVs will be tested every two cycles (approximately four years) as 
compared to the current one cycle (approximately two years) 
frequency. As described above, the lift test of the main valve disc 
is currently performed at an offsite facility. A review of offsite 
testing data for the years 1987 through 1998 was performed for the 
PBAPS, Units 2 and 3 SRVs. Since the design of the SRVs is to ensure 
operation of the overpressurization protection and the ADS 
[Automatic Depressurization System] function is to reduce reactor 
pressure during a small break LOCA [loss-of-coolant accident], the 
review consisted of looking for any failures of the main valve disc 
to stroke open during setpoint actuation. This review consisted of 
reviewing ``as-found'' test data since any failures following a 
rebuild would be found during the final certification testing. Based 
on a review of as-found data, it was concluded that there were no 
reported cases of the main disc failing to open during setpoint 
pressure testing. Therefore, deleting the requirement for functional 
testing of the SRVs during each unit startup is not expected to 
negatively impact these test results.
    Therefore, eliminating the functional test is not expected to 
negatively impact these test results or involve a significant 
increase in the probability of an accident previously evaluated.
    As discussed in the PBAPS, Units 2 and 3 Updated Final Safety 
Analyses Report (UFSAR), analyzed events resulting in a nuclear 
system pressure increase, such as MSIV [main steam line isolation 
valve] closure, generator load rejection, turbine trip, failure of 
the turbine bypass valves to open, and loss of main condenser 
vacuum, take credit for the SRVs opening to mitigate the 
consequences of these events. The proposed changes will not increase 
the consequences of these events, since a series of remaining tests 
will ensure all SRV components will function. The SRVs will 
therefore be capable of performing their design functions.
    SRV second stage valve leakage can be increased as a result of 
corrosion/debris introduced on the seating area surface. Second 
stage leakage, if allowed to continually increase, will eventually 
start to depressurize the volume above the SRV main valve piston to 
the extent that sufficient differential pressure will lift the main 
valve disc. Reactor vessel coolant inventory decrease due to an 
inadvertent opening of a Safety Relief Valve is an abnormal 
operating transient event. This event can be a precursor to fuel 
failure due to gradual loss of coolant, and the mitigation is 
similar to the small break LOCA. Under the proposed change, it is 
expected that the probability of SRV leakage will decrease, thus the 
probability of occurrences of an inadvertent SRV actuation is 
reduced, therefore reducing the probability or

[[Page 40560]]

consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The SRVs will not be operated or tested in a manner contrary to 
their design. As a result, no new mode of operation is introduced. 
Therefore, the revised testing will not create a new failure mode of 
the SRVs which could create the possibility of a new or different 
kind of accident from any previously evaluated. Since other tests, 
taken together, confirm the entire SRV assembly functions 
adequately, this proposed change is justified. The proposed change 
to delete the SRV functional test at power will not impact the 
ability of the SRV to open and provide their intended safety 
function.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    By removing the Technical Specification requirements to perform 
the in-situ functional testing during startup, the probability of 
inadvertently opening of a SRV should be reduced through the 
elimination of a potential initiator of SRV second stage disc 
leakage and subsequent erosion. This Technical Specification change 
will aid in decreasing SRV leakage and improve SRV reliability at 
power operations. Eliminating the SRV in-situ functional test during 
startup will increase the margin of safety during operations, 
transients, or accidents. Remaining surveillance testing and 
inspections assure each component necessary for successful opening 
of the SRV function properly as designed.
    Removal of the functional test will not negatively impact the 
Technical Specifications lift setpoints of the SRVs necessary for 
the function of the safety mode. The functional test does not 
completely test the safety mode of the SRV which is based on the 
Technical Specifications lift setpoints.
    Offsite testing at operating steam pressure ensures the 
operability of the SRV pilot, second stage, and main valve function. 
The valves are refurbished and post maintenance testing is performed 
at a steam pressure of 1040 psig. Upon successful test completion, 
the valve receives written certification from the lab and is 
returned to PBAPS for reinstallation. To receive certification, the 
valve must have zero main seat leakage and meet the acceptance 
criteria for setpoint pressure. These tests satisfy the requirements 
of the PBAPS IST [Inservice Testing] Program and TS. The tests 
contained in the proposed TS SR 3.4.3.2 and 3.5.1.12 will verify the 
operation of the solenoid and second stage disc movement of all 11 
SRVs in the depressurization mode.
    The remaining segments of the SRV tests verify the ability of 
the SRV logic. In summary, this change will not involve a 
significant reduction in the margin of safety, because of the 
reduction in SRV degradation, and the remaining tests confirm the 
valves will function properly when required.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, this 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: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.
    Attorney for Licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: June 16, 1998.
    Description of amendment request: The proposed change would 
relocate the Safety Review Committee (SRC) review, audit, and related 
record keeping requirements from the Technical Specifications (TSs) to 
Chapter 17 of the Final Safety Analysis Report (FSAR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously analyzed?
    Response: This amendment application does not involve a 
significant increase in the probability or consequences of an 
accident previously analyzed. The relocation of the SRC review, 
audit, and related record keeping requirements from the TS to the 
FSAR does not alter the performance or frequency of these 
activities. Future changes to the QA [Quality Assurance] program, 
located in Chapter 17 of the FSAR, which constitute a reduction in 
commitments, are governed by 10 CFR 50.54(a). Therefore, sufficient 
controls for these requirements exist and these changes do not 
involve a significant increase in the probability or consequences of 
an accident previously analyzed.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response: This amendment application does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. The proposed changes involve the relocation of 
SRC requirements from the TS to the FSAR. Relocation of these 
requirements does not affect plant equipment or the way the plant 
operates. The reviews, audits, and record keeping will continue to 
be performed in the identical manner as they are currently being 
performed. Therefore, the proposed revisions cannot create a new or 
different kind of accident.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: This amendment application does not involve a 
significant reduction in a margin of safety. The requested Technical 
Specification revisions relocate SRC review, audit and related 
record keeping requirements from the TS to the FSAR. These 
requirements are not being altered by this relocation. The reviews, 
audits, and record keeping will continue to be performed in the 
identical manner as they are currently being performed. Any changes 
to these requirements which constitute a reduction in commitments 
will be processed in accordance with 10 CFR 50.54(a). 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 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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Director.

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

    Date of amendment request: July 10, 1998.
    Description of amendment request: The proposed changes would 
relocate portions of reactor coolant chemistry requirements from the 
technical specifications (TSs) to licensee-controlled procedures. 
Changes to the relocated requirements will then be controlled by the 
provisions of 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 TS amendment will not significantly increase the 
probability or consequences of any previously evaluated accidents.
    The proposed changes simplify the TS, meet regulatory 
requirements for relocated TS, and implement the recommendations of 
the Commission's Final Policy Statement on TS improvements. Future 
changes to these requirements will be controlled by 10 CFR

[[Page 40561]]

50.59. The proposed changes are administrative in nature and do not 
involve any modification to any plant equipment or affect plant 
operation. Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of any 
previously evaluated accident.
    2. The proposed TS amendment will not create the possibility of 
a new or different kind of accident.
    The proposed changes are administrative in nature, do not 
involve any physical alterations to any plant equipment, and cause 
no change in the method by which any safety related system performs 
its function. Therefore, this proposed TS amendment will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The proposed TS amendment will not involve a significant 
reduction in a margin of safety.
    The proposed changes are administrative in nature, will not 
alter the basic regulatory requirements, and do not affect any 
safety analyses. Therefore, no margin of safety is reduced as a 
result of these changes.

    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 Project Director: S. Singh Bajwa, Director.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of amendment request: June 5, 1997, and supplemented April 21, 
1998.
    Description of amendment request:
    Part 1--DG Online Testing:
    The proposed amendment involves the testing of the standby diesel 
generators (DGs) and revises the Watts Bar Unit 1 (WBN) Technical 
Specifications (TSs) to allow additional testing of the DGs on-line 
during MODES 1 and 2. The proposed changes affect Surveillance 
Requirement (SR) 3.8.1.14. The testing performed for this surveillance 
fulfills the requirements of Regulatory Guide 1.9, ``Selection, Design, 
Qualification, and Testing of Emergency Diesel Generator Units Used as 
Class 1E Onsite Electric Power Systems at Nuclear Power Plants.'' This 
testing is performed once every 18 months to ensure that the DGs can 
start and run continuously for an interval of not less than 24 hours. 
Specifically, the proposed amendment revises SR 3.8.1.14 and its 
associated Bases to delete the note which prohibits the performance of 
the on-line 24 hour test during MODES 1 or 2.
    Part 2--DG Battery Testing:
    As currently written, the TSs permit testing of the DG batteries 
and chargers only during MODES 5 and 6 when operability of all four DGs 
is not required. The proposed amendment would revise the Watts Bar Unit 
1 TSs to allow testing of the DG batteries and battery chargers during 
MODES 1, 2, 3, and 4 as well. Implementation of these changes will 
require entry into Action B.4 of TS 3.8.1 for the affected diesel.
    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:

Part 1--DG Online Testing

    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 amendment to allow the 24-hour DG endurance run to 
be conducted during any mode of operation does not significantly 
increase the probability or consequences of an accident previously 
evaluated in Chapter 15 of the FSAR [Final Safety Analysis Report] 
since the capability to safely shutdown the plant following a LOOP 
[loss of offsite power], LOCA [loss-of-coolant accident] or LOCA/
LOOP coincident with a single failure is maintained throughout the 
surveillance test. The 24-hour endurance test does not disable any 
of the automatic actuations and interlocks of the DG control 
functions, nor prevent the satisfactory completion of the LOOP or 
LOCA/LOOP loading sequence if a LOOP or LOCA signal is received at 
any time during the test. Required Class-1E onsite power operability 
during normal operation, shutdown cooling, loss of offsite power, 
and accident conditions will be the same.
    In addition, the performance of proposed Surveillance 
Requirement 3.8.1.14 during MODES 1 or 2 will not significantly 
increase the consequences of perturbations to any of the electrical 
distribution systems that could result in a challenge to steady 
state operation or to plant safety systems. Performance of proposed 
Surveillance Requirement 3.8.1.14 during MODES 1 or 2, or failure of 
the surveillance, will not cause, or result in, an anticipated 
operational occurrence with attendant challenges to plant safety 
systems that has not been previously analyzed for the existing 
monthly surveillances.
    Therefore, TVA concludes that the above change 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.
    The requested changes do not result in a new or different kind 
of accident from that previously analyzed in WBN's Final Safety 
Analysis Report. The changes propose to eliminate restrictions of 
the plant operating modes in which standby DG system testing may be 
performed but does not change the type of testing performed and are 
not due to modification of the system design. NRC's assessment of 
the testing of the DGs in the configuration proposed is documented 
in Section 8.3.1.12 of Supplements 13 and 14 of the Safety 
Evaluation Report and in letters dated June 20, 1991, and March 28, 
1994.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in the margin of 
safety.
    As previously stated, performance of proposed Surveillance 
Requirement 3.8.1.14 during Modes 1 or 2 will not cause, or result 
in, an anticipated operational occurrence with attendant challenges 
to plant safety systems that has not been previously analyzed for 
the existing monthly surveillances. Therefore, implementation of the 
proposed amendment will not reduce the margin of safety for this 
system.

Part 2--DG Battery Testing

    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 changes to the TSs apply only to the DG battery 
system and do not in any way affect the vital battery system or 
safety system loads supplied by the vital battery system. The 
changes do not result in a condition where the design or function of 
the DGs or DG battery systems would be modified. The DG battery 
subsystems supply only the control and field flashing power to 
support a single DG and do not supply any other unrelated system 
loads or functions. Therefore, manipulation of the DG battery system 
is not a credible means of perturbing the vital power distribution 
system and challenging safety systems. In addition, the 
surveillances for the DG batteries are required to be performed only 
once every 18 months.
    A DG declared inoperable due to the testing must be returned to 
operable status within 72 hours in accordance with Action B.4 of TS 
3.8.1. To ensure this could be achieved, the results of previous 
performances of the SRs were reviewed. From this review, it was 
established that in accordance with LCO [limiting condition for 
operation] 3.8.6, Table 3.8.6-1, Note c, the batteries can be 
restored within 72 hours to a condition where the charging current 
is less than 1 ampere. Achieving this charging current for the DG 
batteries is acceptable for meeting specific gravity limits 
following a battery recharge for a maximum of 31 days. In addition, 
the DG sets are occasionally removed from the standby condition to 
perform preventative and/or corrective maintenance. The intent is to 
perform this

[[Page 40562]]

testing in conjunction with other required maintenance activities 
such that adverse effects on diesel unavailability are minimized. 
Compliance with the 10 CFR 50.65 Maintenance Rule program 
requirements for diesel unavailability ensures that any diesel 
inoperability incurred by this change is minimized.
    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 requested changes do not result in a new or different kind 
of accident from that previously analyzed in WBN's Final Safety 
Analysis Report. The changes propose to eliminate restrictions of 
the plant operating modes in which DG battery system testing may be 
performed but does not change the type of testing performed and are 
not due to modification of the system design. The requested changes 
will result in a DG being declared inoperable in accordance with 
Action B.4 of TS 3.8.1 for the duration of the testing, but does not 
impact the existing time limitations for the LCO. This change does 
not alter system performance and does not introduce any new accident 
initiators or scenarios.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in the margin of 
safety.
    The proposed amendment concerns only the conduct of testing but 
does not in any way affect the performance parameters of the safety 
system or in any way affect the ability of the system to perform its 
safety function of providing control and field flashing power for 
the DGs. Consequently, operation of the facility in accordance with 
the requested changes would 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: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of amendment request: June 26, 1998
    Description of amendment request: The proposed amendment would 
revise the Watts Bar Nuclear Plant (WBN) Technical Specifications (TS) 
and associated Bases to delete the power range neutron flux high 
negative rate reactor trip function based on the analysis provided in 
Westinghouse Electric Corporation topical report WCAP-11394-P-A, 
``Methodology for the Analysis of the Dropped Rod Event.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The negative flux rate trip deletion does not increase the 
probability or consequences of core damage accidents resulting from 
dropped RCCA [rod cluster control assembly] events previously 
analyzed. The safety functions of other safety related systems and 
components, which are related to accident mitigation, have not been 
altered. All other primary protection (reactor trip and ESF) 
functions are not impacted by the elimination of the negative flux 
rate trip function. The consequences of accidents previously 
evaluated in the FSAR [final safety analysis report] are unaffected 
by this proposed change because no change to any equipment response 
or accident mitigation scenario has resulted. There are no 
additional challenges to fission product barrier integrity. No new 
radiological analyses are required. Therefore the proposed change 
will have no effect on the probability or consequences of accidents 
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 negative flux rate trip deletion does not create the 
possibility of a new or different kind of accident than any accident 
already evaluated in the FSAR. No new accident scenarios, failure 
mechanisms, or limiting single failures are introduced as a result 
of this proposed change. The proposed modification does not 
challenge the performance or integrity of any safety-related 
systems.
    It has been demonstrated that the function of the negative flux 
rate trip can be eliminated by the approved methodology described in 
WCAP 11394-P-A. A Watts bar specific analysis has confirmed that for 
the dropped RCCA and dropped RCCA bank event, no direct reactor trip 
or automatic power reduction is required to meet the DNB [departure 
from nucleate boiling] licensing basis for this Condition II event. 
The negative flux rate trip function is not credited as a backup for 
any other Chapter 15 event. Thus, this change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The margin of safety associated with the acceptance criteria for 
any postulated WBN accident is unchanged. It has been demonstrated 
that the function of the negative flux rate trip can be eliminated 
by the approved methodology described in WCAP 11394-P-A. Watts Bar 
specific analysis has confirmed that the dropped RCCA and dropped 
RCCA bank acceptance criteria (DNB) continue to be met. Conformance 
to the regulatory criteria for plant operation with the negative 
flux rate trip deletion is demonstrated, and regulatory limits (DNB) 
are not exceeded. The modification will have no effect on the 
availability, operability, or performance of the safety-related 
systems and components. Therefore, the proposed license 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: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

The Cleveland Electric Illuminating Company, Centerior Service Company, 
Duquesne Light Company, Ohio Edison Company, Pennsylvania Power 
Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power 
Plant, Unit 1, Lake County, Ohio

    Date of amendment request: June 30, 1998.
    Description of amendment request: The proposed license amendment 
revises Technical Specification 3.1.7, ``Standby Liquid Control 
System.'' The purpose of the proposed change is to increase the boron 
concentration in the Standby Liquid Control System for the Perry 
Nuclear Power Plant Cycle 8 fuel design, and to provide margin for 
future cycles as required.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change will not significantly increase the 
probability or consequences of an accident previously evaluated. The 
change will only vary the ratio of borax to boric acid that resides 
within the Standby Liquid

[[Page 40563]]

Control System (SLCS) as the neutron absorber.
    Changing the definition of the solution from a mixture of Sodium 
Pentaborate having a molar ratio of 0.200, to a mixture of borax and 
boric acid having a nominal molar ratio of 0.229, does not degrade 
the stability of the solution, change the mixing accuracy 
requirements, or reduce the temperature margins that might add to 
the risk of solution crystallization. For each cycle, the reload 
safety analysis confirms that the SLCS boron concentration will 
satisfy the Technical Specification requirements for the Perry 
Nuclear Power Plant (PNPP).
    The 5 deg.F margin of safety for solution solubility will 
continue to be maintained and supported by the Containment Building 
ambient temperatures and additionally supplemented by auto initiated 
heating on the SLCS tank and piping. The chosen borax and boric acid 
molar ratio will continue to maintain a limiting chemical addition 
mass, to the plus 5 deg.F solubility limit, greater than or equal to 
the current 0.200 mixture. Any inaccuracies associated with tank 
temperature, tank volume, chemical analysis, and initial and 
subsequent chemical additions to the tank will also remain the same.
    The primary reactivity control system for postulated accident 
conditions is the control rod system. The SLCS is a redundant 
reactivity control system to the control rod system and is used in 
special plant capability demonstration events cited in Appendix A of 
the Updated Safety Analysis Report (USAR), Chapter 15, which are 
extremely low probability non-design basis postulated incidents. 
There are no postulated accidents evaluated in USAR Chapter 15 that 
take credit for two or more reactivity control systems preventing or 
mitigating each accident. There is no increase to the radiological 
consequences of postulated incidents with the proposed change.
    With the implementation of this proposed change, the SLCS will 
continue to operate and perform to all of its current requirements 
for providing shutdown margin under operating and ATWS conditions 
per 10 CFR 50, Appendix A, General Design Criterion (GDC) 26 and 10 
CFR 50.62. The proposed change will not alter the operation of any 
plant equipment assumed to function in response to an analyzed event 
or otherwise increase its failure probability. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not impact the operation of the SLCS or 
the function of any of the active components. No new system 
interactions are created by this change and any parameters or 
conditions that could contribute to the initiation of accidents 
different than those already evaluated in the USAR are not impacted. 
The change will only vary the ratio of borax to boric acid that 
resides within the SLCS as the neutron absorber. The proposed values 
for solution molar ratio and boron concentration ensure that 
solution temperature margins are maintained greater than or equal to 
the current required margin to prevent solution crystallization. As 
a result, no new failure modes are being introduced.
    Changing the definition of the solution from Sodium Pentaborate 
having a nominal molar ratio of 0.200, to a mixture of borax and 
boric acid having a nominal molar ratio of 0.229 does not degrade 
the stability of the solution, change the mixing accuracy 
requirements, or reduce the temperature margins that might add to 
the risk of solution crystallization.
    Sufficient margin will be maintained to allow for expected 
deviations in the molar ratio and boron weight as the result of 
variations in product composition, test measurement inaccuracies, 
and for chemical addition inaccuracies. The boron concentration 
required within the SLC system to meet the required shutdown margin, 
will continue to be determined for each fuel cycle as part of the 
reload safety analysis per Technical Specifications. The borax and 
boric acid concentration will remain controlled via the Technical 
Specification Surveillance Requirements and the associated 
administrative procedures, USAR text, and existing licensing 
commitments.
    The SLC system will meet its design basis requirements for the 
weight of boron injected and for maintaining the required 
temperature margin for system operation. As the result of the 
proposed change to increase the minimum boron concentration, a new 
minimum required SLC pump flow rate was determined for compliance 
with the NRC ATWS Rule 10 CFR 50.62.
    The proposed change meets current regulations, maintains the 
fundamental safety principles of plant design, and the associated 
margins of safety. With the implementation of the proposed change, 
the SLCS will continue to operate and perform to all of its current 
requirements for providing shutdown margin under operating and ATWS 
conditions per 10 CFR 50, Appendix A, GDC 26 and 10 CFR 50.62. As a 
result, no new failure modes are being introduced. There are no 
changes in the methods governing normal plant operations, nor are 
the methods used to respond to plant transients altered. 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 change does not involve a significant reduction 
in a margin to safety.
    The required margin of safety for the SLCS solution ensures an 
adequate margin of solubility such that no precipitation will occur 
in the SLC storage tank. The current margin is provided by 
maintaining a minimum solution temperature that is no less than the 
saturation temperature corresponding to the concentration of the 
solution in the storage tank plus 5 deg.F.
    This 5 deg.F provides the adequate margin for inaccuracies 
associated with tank temperature, tank volume, chemical analysis, 
and initial and subsequent chemical additions to the tank. The 
proposed change does not impact the inaccuracies associated with 
tank temperature, tank volume, chemical analysis, and initial and 
subsequent chemical additions to the tank. The new analytical design 
values for the molar ratio and boron concentration will continue to 
maintain the solution temperature margins in excess of the current 
minimum specified to prevent solution crystallization.
    Ambient temperatures within the building that houses the SLC 
storage tank, the Containment Building, will maintain the solution 
temperature. Additionally, the solution temperature is maintained by 
the presence of auto initiated tank heaters and pipe heat tracing. 
The 5 deg.F margin will be maintained with the new SLCS mole ratio 
and higher boron concentration with the existing instrument 
setpoints and administrative controls.
    The proposed change maintains the same reactor shutdown margin 
for the next fuel cycle and does not reduce the margin of safety for 
any system parameter as defined in the bases for the Technical 
Specifications. The proposed change will not physically alter the 
SLCS's physical configuration or components or introduce new system 
interactions that could produce any parameters or conditions that 
could contribute to a reduction of safety for any other system or 
scenario. The change will only vary the ratio of borax to boric acid 
that resides within the SLCS as the neutron absorber.
    Therefore, with the implementation of this proposed change, the 
SLCS will continue to operate and perform to all of its current 
requirements for providing shutdown margin under operating and ATWS 
conditions per 10 CFR 50, Appendix A, GDC 26 and 10 CFR 50.62 and 
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: Perry Public Library, 3753 
Main Street, Perry, OH 44081.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Ronald R. Bellamy (Acting).

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

    Date of amendment request: June 30, 1998.
    Description of amendment request: The licensee proposes to delete 
the calibration requirements for emergency core cooling actuation 
instrumentation--core spray (CS) subsystem and low pressure coolant 
injection (LPCI) system auxiliary power

[[Page 40564]]

monitor since the relays operate from a switched input and functional 
testing is sufficient to demonstrate the relay pickup/dropout 
capability.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated:
    The proposed change does not involve a change to the plant 
design or operation. The Auxiliary Power Monitor logic relays 
installed are tested to fully demonstrate operability without 
performance of a calibration on the pickup voltage value. The design 
intent of the relays is to start LPCI and CS pumps as soon as 
possible without causing loss of the normal or emergency power 
supplies and within the time frames specified in the LOCA analysis 
of record. The proposed change does not affect any of the parameters 
or conditions that contribute to initiation of any accidents 
previously evaluated. Thus, the proposed change cannot increase the 
probability of an accident previously evaluated.
    The proposed change does not involve a change in the operation 
of the relays controlling [Residual Heal Removal] RHR and CS Pump 
start with normal power available nor the initial RHR pump start on 
a LOCA with normal power not available or the time delay start of 
the remaining RHR or CS pumps with normal power not available. 
Failure of the relays to pickup would still result in the start 
sequence for normal power not available. The logic for both start 
sequences is verified independent of an instrument calibration and 
is consistent with the LOCA analysis and the EDG load analysis, 
therefore, the proposed change does not significantly increase the 
consequences of any accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any previously evaluated:
    This proposed change will not involve any physical changes to 
plant systems, structures or components (SSC), or the manner in 
which these SSCs are operated or maintained. The calibration 
requirement has previously been considered to be met by performance 
of the Simulated Automatic Actuation Test. Deletion of the 
calibration requirement will not affect the RHR or CS Pumps starting 
on a LOCA signal, with or without an [Loss of Normal Power] LNP. The 
operability of the Auxiliary Power Monitor relays will still be 
tested under the Functional test and Trip System Logic and Simulated 
Automatic Actuation tests at the frequencies specified. Therefore, 
this change will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety:
    This proposed change to delete the calibration requirement for 
the CS and LPCI Auxiliary Power Monitor functions will not change 
operation of the RHR or CS Pump start sequences on a LOCA signal, 
with or without normal power available. The instantaneous logic 
sequence relays and time delay relays will function to initiate RHR 
and CS Pump start as designed. RHR and CS Pump start times will 
remain within the LOCA Safety Evaluation of record. Operability of 
the relays and associated circuitry are still demonstrated by the 
Functional test and associated Trip System Logic and Simulated 
Automatic Actuation tests. Therefore, this change does not involve a 
significant reduction in a margin of safety.

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

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

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

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

    Date of amendment request: June 26, 1998.
    Brief Description of amendment: The proposed amendment would revise 
Technical Specification (TS) 3.7.8, ``Ultimate Heat Sink (UHS),'' to 
permit an 8-hour delay in UHS temperature restoration period prior to 
entering the plant shutdown required actions. Also, for the duration of 
the restoration, service water system (SWS) temperature will be 
monitored hourly, and should the temperature exceed 99 degrees F, the 
plant will enter TS 3.7.8 required action A.1, and be in MODE 3 within 
6 hours.
    Date of publication of individual notice in the Federal Register: 
July 8, 1998 (63 FR 36967).
    Expiration date of individual notice: July 22, 1998, for comments; 
August 7, 1998, for hearings.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.

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: March 3, 1998, as supplemented by 
letters dated April 24 and May 7, 1998.
    Description of amendment request: The proposed amendments would 
revise Figure 5.1-1 of the Technical Specifications (TS) to show the 
new location of the meteorological tower. The meteorological tower will 
be relocated to a new location to facilitate use of the current 
location as a construction site. The proposed TS change does not change 
the related TS Section 5.1.1.
    Date of publication of individual notice in Federal Register: June 
29, 1998 (63 FR 35293).
    Expiration date of individual notice: July 29, 1998.
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, Charlotte, North Carolina.

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

[[Page 40565]]

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.

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

    Date of application for amendment: January 28, 1998 (NRC-98-0002).
    Brief description of amendment: The amendment revises technical 
specification surveillance requirements 4.8.2.1.a.2, 4.8.2.1.b, and 
4.8.2.1.c.4 to accommodate new limits associated with the design of the 
replacement Division II 130/260-volt dc battery.
    Date of issuance: July 9, 1998.
    Effective date: July 9, 1998, with full implementation prior to 
restart from the sixth refueling outage.
    Amendment No.: 121.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 25, 1998 (63 
FR 9597).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 9, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
Ellis Reference and Information Center, 3700 South Custer Road, Monroe, 
Michigan 48161.

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

    Date of application for amendment: December 10, 1997 (NRC-97-0105), 
as supplemented January 28 and April 9, 1998.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 2.2.1, ``Reactor Protection System Instrumentation 
Setpoints,'' TS 3.3.1, ``Reactor Protection System Instrumentation,'' 
TS 3.3.6, ``Control Rod Block Instrumentation,'' TS 3.4.1.1, 
``Recirculation Loops,'' and the associated Bases to accommodate an 
upgrade of the power range neutron monitoring system. The amendment 
also revises the first page of Table 3.3.6-2 to correct a typographical 
error in the title.
    NRC has also granted the request of Detroit Edison Company to 
withdraw a portion of its December 10, 1997, application. The proposed 
change would have revised TS Surveillance Requirement 4.3.1.3 and its 
associated Bases to indicate response time testing is performed only on 
applicable channels. However, following discussions with the NRC staff, 
the licensee withdrew the proposed change in a letter dated April 9, 
1998 (NRC-98-0037). For further details with respect to this action, 
see the application for amendment dated December 10, 1997, as 
supplemented above, and the licensee's letter dated April 9, 1998, 
which withdrew this portion of the application for license amendment, 
and the staff's Safety Evaluation enclosed with the amendment. The 
above documents 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 listed below.
    Date of issuance: July 13, 1998.
    Effective date: July 13, 1998, with full implementation prior to 
restart from the sixth refueling outage. Implementation of this 
amendment shall include preparation of Design Calculation DC-5721, 
Volume I, and performance of a human factors review for the 
installation of the plant modification as described in the licensee's 
application dated December 10, 1997, as supplemented January 28 and 
April 9, 1998, and as evaluated in the staff's safety evaluation 
attached to this amendment.
    Amendment No.: 122.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: January 14, 1998 (63 FR 
2279). The January 28 and April 9, 1998, letters provided clarifying 
information and updated TS pages that were within the scope of the 
original Federal Register notice and did not change the staff's initial 
proposed no significant hazards considerations determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 13, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
Ellis Reference and Information Center, 3700 South Custer Road, Monroe, 
Michigan 48161.

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

    Date of application for amendments: April 20, 1998.
    Brief description of amendments: The amendments revise Tables 3.3-3 
and 4.3-2 of the Technical Specifications of each unit, correcting the 
operation mode applicability of the control room area ventilation 
actuation logic and relays from ``All'' to ``1, 2, 3, 4.''
    Date of issuance: July 9, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 167--Unit 1; 159--Unit 2.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 20, 1998 (63 FR 
27761).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 9, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

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

    Date of amendment request: October 2, 1996.
    Brief description of amendments: The amendments revise the ANO-1&2 
TSs by relocating selected TS requirements related to instrumentation 
from the TS to the Updated Final Safety Analysis Report. The NRC 
provided guidance to all holders of operating licenses or construction 
permits for nuclear power reactors on the proposed TS changes in 
Generic Letter 95-10, ``Relocation of

[[Page 40566]]

Selected Technical Specifications Requirements Related to 
Instrumentation,'' dated December 15, 1995.
    Date of issuance: July 13, 1998.
    Effective date: July 13, 1998, to be implemented within 30 days.
    Amendment Nos.: 192 and 191.
    Facility Operating License Nos. DPR-51 and NPF-6: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 15, 1997 (62 FR 
2188).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 13, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

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

    Date of amendment request: March 27, 1997, as supplemented by 
letters dated April 3, July 21, October 23, November 13, and December 
12, 1997, January 21, January 29, March 23, May 1, May 19, and May 21, 
May 28, and June 12, 1998.
    Brief description of amendment: The amendment changes Appendix A 
Technical Specification by increasing the Spent Fuel Pool storage 
capacity from 1088 to 2398 fuel assemblies and by increasing the 
maximum fuel enrichment from 4.9 w/o (weight percent) to 5.0 w/o U-235.
    Date of issuance: July 10, 1998.
    Effective date: July 10, 1998.
    Amendment No.: 144.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 2, 1997 (62 FR 
63732).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 10, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

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

    Date of application for amendments: March 12, 1998.
    Brief description of amendments: The amendments revised Turkey 
Point Units 3 and 4 Facility Operating Licenses and Technical 
Specifications to remove certain license conditions and oudated 
references, and to incorporate an organizational change.
    Date of issuance: July 9, 1998.
    Effective date: July 9, 1998.
    Amendment Nos.: 198 and 192.
    Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments 
revised Turkey Point Units 3 and 4 Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: April 8, 1998 (67 FR 
17225).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 9, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

Niagara Mohawk Power Corporation, Docket Nos. 50-220 and 50-410, Nine 
Mile Point Nuclear Station Unit Nos. 1 and 2, Oswego County, New York

    Date of applications for amendments: May 15, 1998 (two letters, one 
for each unit).
    Brief description of amendment: The amendments change 
administrative sections of the Technical Specifications to reflect a 
restructuring of licensee's Nuclear Division upper management 
organization.
    Date of issuance: July 7, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 162 and 83.
    Facility Operating License Nos. DPR-63 and NPF-69: Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: June 2, 1998 (63 FR 
30026).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 7, 1998.
    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.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Stawn, 1400 L Street, NW, Washington, DC 20005-3502.
    NRC Project Director: S. Singh Bajwa.

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

    Date of application for amendments: September 28, 1995, and April 
23, 1998.
    Brief description of amendments: The amendments revise Technical 
Specification 3/4.8.1.2, ``Electrical Power Sources--Shutdown,'' by 
adding a note to surveillance requirement 4.8.1.2 that identifies those 
surveillances which are required to be performed during Modes 5 and 6 
(cold shutdown and refueling, respectively).
    Date of issuance: July 14, 1998.
    Effective date: July 14, 1998.
    Amendment Nos.: 212 and 192.
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 8, 1995 (60 FR 
56369).
    The April 23, 1998, letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 14,1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

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

    Date of application for amendment: September 29, 1998, as 
supplemented February 6, 1998, April 17, 1998, and June 4, 1998.
    Brief description of amendment: This amendment revises the 
allowable value and trip setpoint for the main steam isolation high 
steam flow input into limiting condition for operation.
    Table 3.3.2-1, function 4.d.
    Date of issuance: July 14, 1998.
    Effective date: July 14, 1998.
    Amendment No.: 71.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 22, 1997 (62 FR 
54876).
    The February 6, 1998, April 17, 1998, and June 4, 1998, letters 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 14, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.

[[Page 40567]]

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

    Date of application for amendments: June 6, 1996, as supplemented 
September 26, 1997, January 23, 1998, and May 19, 1998 (TS-372).
    Brief description of amendments: Changes to the technical 
specifications administrative controls related to quality assurance, 
and other administrative and editorial changes.
    Date of issuance: July 9, 1998.
    Effective date: July 9, 1998.
    Amendment Nos.: 233, 252, and 211.
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 25, 1996 (61 
FR 50346).
    The supplemental letters dated September 26, 1997, January 23, and 
May 19, 1998 did not change the original no significant hazards 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 9, 1998.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611.

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear 
Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: April 24, 1998.
    Brief description of amendment: This amendment changed Technical 
Specification (TS) Section 3/4.3.1.1, ``Reactor Protection System 
Instrumentation,'' TS Section 3/4.3.2.1, ``Safety Features Actuation 
System Instrumentation,'' TS Section 3/4.3.2.2, ``Steam and Feedwater 
Rupture Control System Instrumentation,'' and the associated TS bases. 
The TS tables of response time limits were relocated to the Davis-Besse 
Technical Requirements Manual. Other changes in these TS sections were 
also made consistent with the relocation.
    Date of issuance: July 7, 1998.
    Effective date: July 7, 1998.
    Amendment No.: 225.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated July 7, 1998.
    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.

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

    Date of application for amendments: April 24, 1996, as supplemented 
December 15, 1997, and June 22, 1998.
    Brief description of amendments: These amendments revise Technical 
Specifications (TS) Section 15.7, ``Radiological Effluent Technical 
Specifications (RETS).'' Portions of the RETS are moved to licensee-
controlled documents consistent with Nuclear Regulatory Commission 
guidance on TS improvements. Other sections of the TSs have also been 
revised consistent with the removal of portions of the RETS.
    Date of issuance: July 13, 1998.
    Effective date: July 13, 1998, with full implementation within 45 
days. Implementation shall include relocation of certain Technical 
Specification requirements to licensee-controlled documents, as 
described in the licensee's application dated April 24, 1996, as 
supplemented by letter dated December 15, 1997, and June 22, 1998, and 
evaluated in the staff's safety evaluation attached to the amendments.
    Amendment Nos.: 184 and 188.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28620) The December 15, 1997, and June 22, 1998, submittals provided 
additional clarifying information and updated TS pages. This 
information was within the scope of the original Federal Register 
notice and did not change the staff's initial no significant hazards 
considerations determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 13, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.

    Dated at Rockville, Maryland, this 22nd day of July 1998.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 98-20111 Filed 7-28-98; 8:45 am]
BILLING CODE 7590-01-P