[Federal Register Volume 62, Number 175 (Wednesday, September 10, 1997)]
[Notices]
[Pages 47696-47705]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-23820]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission 
(the Commission or NRC staff) is publishing this regular biweekly 
notice. Public Law 97-415 revised section 189 of the Atomic Energy Act 
of 1954, as amended (the Act), to require the Commission to publish 
notice of any amendments issued, or proposed to be issued, under a new 
provision of section 189 of the Act. This provision grants the 
Commission the authority to issue and make immediately effective any 
amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the

[[Page 47697]]

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 August 18, 1997, through August 28, 1997. 
The last biweekly notice was published on August 27, 1997 (62 FR 
45452).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By October 10, 1997, 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

[[Page 47698]]

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

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of amendment request: August 15, 1997.
    Description of amendment request: The proposed amendment would 
revise portions of the facility Technical Specifications regarding 
facility staffing and training requirements to power operations. By 
letter dated August 7, 1997, the licensee certified permanent cessation 
of power operations and permanent removal of fuel from the reactor 
vessel. By two letters both dated August 15, 1997, the licensee has 
also submitted a related ``Request for Exemption from Certain 
Requirements of 10 CFR 50.54, Conditions of License,'' and a ``Request 
for Approval of the Certified Fuel Handler Training and Retraining 
Program.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.
    The proposed change does not:

    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The purpose of the proposed change is to eliminate the 
requirements for licensed operators and a licensed operator training 
program and to replace those with certified fuel handlers and a 
certified fuel handler training and retraining program. Since the 
plant has permanently ceased operation and will be maintained in a 
defueled condition, the range of accidents for which an operator 
needs to be trained has significantly diminished such that a 
training program of the depth and breadth of that required by 10 CFR 
[Part] 55 is no longer needed. In lieu of a 10 CFR [Part] 55 
licensed operator training program, a[n] NRC-approved certified fuel 
handler training and retraining program will be implemented. Since 
this training program will adequately equip appropriate operations 
personnel for fuel handling operations, including responses to 
abnormal events/accidents, there will be no increase in the 
probability of these events occurring or in the consequences of 
these events. The proposed changes do not affect plant equipment or 
the procedures for equipment operation or response to abnormal 
events/accidents.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The purpose of this proposed change is to eliminate the 
requirements for licensed operators and a licensed operator training 
program and to replace those with certified fuel handlers and a 
certified fuel handler training and retraining program. This change 
ensures the qualifications of operations personnel are commensurate 
with the tasks to be performed and the conditions to be responded 
to. This change does not affect plant equipment or the procedures 
for operating plant equipment and, therefore, does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change is to eliminate the requirements for 
licensed operators and a licensed operator training program to 
replace those with certified fuel handlers and a certified fuel 
handler training and retraining program. This change ensures the 
qualifications of the operations personnel are commensurate with the 
tasks to be performed and the conditions to be responded to. The 
assumptions for a fuel handling accident in the Fuel Building are 
not affected by the proposed changes. Therefore, the proposed 
amendment does not involve a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578.
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, 329 Bath Road, Brunswick, ME 04011.
    NRC Acting Project Director: Ronald B. Eaton.

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

    Date of amendment request: August 1, 1997.
    Description of amendment request: The amendments would change 
Technical Specification Section 4.2.1 of Appendix B to the licenses. 
The changes include rewording of the section to generically state that 
Public Service Gas & Electric (PSE&G) will adhere to the Section 7, 
Incidental Take Statement, approved by the National Marine Fisheries 
Service (NMFS). Removing the specific requirements of this section 
enables PSE&G to utilize relief granted by the NMFS on a case-by-case 
basis.
    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 changes are administrative in nature and would in no way 
affect the initial conditions, assumptions, or conclusions of the 
Salem [Nuclear] Generating Station, Units 1 and 2, accident 
analyses. In addition, the proposed changes would not affect the 
operation or performance of any equipment assumed in the accident 
analyses. Based on the above information, we conclude that the 
proposed changes would not significantly increase 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 changes are administrative in nature and would in no way 
impact or alter the configuration or operation of the facilities and 
would create no new modes of operation. We therefore conclude that 
the proposed changes would not create the possibility of a new or 
different kind of accident.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    As indicated in the discussion of Criterion 1, the changes are 
administrative in nature and would in no way affect plant or 
equipment operation or the accident analysis. We therefore conclude 
that the proposed changes would not result in 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 47699]]

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: John F. Stolz.

Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364, 
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

    Date of amendments request: July 23, 1997.
    Description of amendments request: The proposed amendments would 
revise the Technical Specifications (TSs) by relocating the reactor 
coolant system pressure and temperature limits from the TSs to the 
proposed Pressure Temperature Limits Report in accordance with the 
guidance provided by Generic Letter 96-03, ``Relocation of the Pressure 
Temperature Limit Curves and Low Temperature Overpressure Protection 
System Limits.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed removal of the Reactor Coolant System (RCS) 
pressure temperature (P/T) limits from the Technical Specifications 
(TSs) and relocation to the proposed Pressure Temperature Limits 
Report (PTLR) in accordance with the guidance provided by Generic 
Letter (GL) 96-03 is administrative in that the requirements for the 
P/T limits are unchanged. The P/T limits proposed for inclusion in 
the PTLR are based on the fluence associated with 2775 MW 
[megawatts] thermal power and operation through 36 effective full 
power years (EFPY). GL 96-03 requires that the P/T limits be 
generated in accordance with the requirements of 10 CFR [Part] 50, 
Appendices G and H, documented in an NRC-approved topical report 
incorporated by reference in the TSs. Accordingly, the proposed 
curves have been generated using the NRC-approved methods described 
in WCAP-14040-NP-A, Revision 2, and meet the requirements of 10 CFR 
[Part] 50, Appendices G and H. TS 3.4.10.1 will continue to require 
that the RCS pressure and temperature be limited in accordance with 
the limits specified in the PTLR. The NRC-approved methodology for 
generating the P/T limit, WCAP-14040-NP-A, Revision 2, will be 
specified in TS 6.9.1.15 and NRC approval will be required in the 
form of a TS Amendment prior to changing the methodology. Use of P/T 
limit curves generated using the NRC-approved methods described in 
WCAP-14040-NP-A, Revision 2, as specified in TS 6.9.1.15, will 
provide additional protection for the integrity of the reactor 
vessel, thereby assuring that the reactor vessel is capable of 
providing its function as a radiological barrier.
    TS 3.4.10.3 for Farley Nuclear Plant (FNP) Unit 1 and Unit 2 
provides the operability requirements for RCS low temperature 
overpressure protection (LTOP). Specifically, TS 3.4.10.3 requires 
that two residual heat removal (RHR) system suction relief valves 
(RHRRVs) be operable or that the RCS be vented at RCS cold leg 
temperatures less than or equal to 310 deg.F. GL 96-03 recognizes 
that RHRRVs do not have variable pressure lift setpoints and states 
that those plants that rely on the RHRRVs for LTOP should continue 
to address the LTOP requirements in the TS. Consistent with GL 96-
03, the Farley Unit 1 and Unit 2 requirements for LTOP will be 
retained in TS 3.4.10.3.
    Based on the above evaluation, the proposed changes are 
administrative in nature and do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    As stated above, the proposed changes to remove the RCS P/T 
limits from the TSs and relocate them to the proposed PTLR is an 
administrative change. Consistent with the guidance provided by GL 
96-03, the proposed P/T limits contained in the proposed PTLR meet 
the requirements of 10 CFR [Part] 50, Appendices G and H, and were 
generated using the NRC-approved methods described in WCAP-14040-NP-
A, Revision 2. The proposed changes do not result in a physical 
change to the plant or add any new or different operating 
requirements on plant systems, structures, or components with the 
exception of limiting the number of operating RCPs [reactor coolant 
pumps] at RCS temperatures below 110 deg.F. Limiting the number of 
operating RCPs below 110 deg.F results in a reduction in the 
[delta]P between the reactor vessel beltline and the RHRRVs, thereby 
providing additional margin to limits of Appendix G. Provisions are 
made to allow the start of a second RCP at temperatures below 
110 deg.F in order to secure the pump that was originally operating 
without interrupting RCS flow. The LTOP enable temperature exceeds 
the minimum LTOP enable temperature determined using the NRC-
approved methods described in WCAP-14040-NP-A, Rev. 2, thereby 
providing additional assurance that the LTOP system will be 
available to protect the RCS in the event of an overpressure 
transient at RCS temperatures at or below 310 deg.F. Using the 
methods contained in WCAP-14040-NP-A, Rev. 2, the minimum boltup 
temperature for the reactor vessel flange region is 60 deg.F which 
is less than the design limits of the fuel cladding. Administrative 
controls require a minimum RCS temperature of 68 deg.F when fuel is 
loaded in the reactor vessel to protect against brittle failure of 
the fuel cladding, and also require that the component cooling water 
(CCW) temperature be maintained between 60 deg.F and 105 deg.F 
during refueling operations, thus reducing the potential for the RCS 
temperature to be less than the minimum boltup temperature specified 
in the proposed PTLRs.
    As stated in the above response, implementation of the proposed 
changes do not result in a significant increase in the probability 
of a new or different accident (i.e., loss of reactor vessel 
integrity). The RCS P/T limits will continue to meet the 
requirements of 10 CFR [Part] 50, Appendices G and H, and will be 
generated in accordance with the NRC approved methodology described 
in WCAP-14040-NP-A, Rev. 2. Therefore, the proposed changes do not 
result in a significant increase in the possibility of a new or 
different accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety is not affected by the removal of the RCS 
P/T limits from the TSs and relocating them to the proposed PTLR. 
The RCS P/T limits will continue to meet the requirements of 10 CFR 
[Part] 50, Appendices G and H. To provide additional assurance that 
the P/T limits continue to meet the requirements of Appendices G and 
H, TS 6.9.1.15 will require the use of the NRC-approved methodology 
described in WCAP-14040-NP-A, Rev. 2, to generate P/T limits. The 
RCS LTOP requirements will be retained in TS 3.4.10.3 due to use of 
the RHRRVs for LTOP, consistent with the guidance provided by GL 96-
03. The LTOP enable temperature exceeds the LTOP enable temperature 
determined in accordance with the NRC-approved methodology, thus 
protecting the RCS in the event of a low temperature overpressure 
transient over a broader range of temperatures than required by 
WCAP-14040-NP-A, Rev. 2. Administrative procedures preclude 
operation of the RCS at temperatures below the minimum boltup 
temperature for the reactor vessel head, thus precluding the 
possibility of tensioning the reactor vessel head at RCS 
temperatures below the minimum boltup temperature. Operation of the 
plant in accordance with the RCS P/T limits specified in the PTLR 
and continued operation of the LTOP system in accordance with TS 
3.4.10.3 will continue to meet the requirements of 10 CFR [Part] 50, 
Appendices G and H, and will therefore, assure that a margin of 
safety is not significantly decreased as the result of the proposed 
changes.
    Based on the preceding analysis, SNC [Southern Nuclear Operating 
Company, Inc.] has determined that removal of the RCS P/T limits 
from the TS and relocation to the proposed PTLR will not 
significantly increase the probability or consequences of an 
accident previously evaluated, create the possibility of a new or 
different kind of accident from any accident previously evaluated, 
or involve a significant reduction in a margin of safety. SNC 
therefore concludes that the proposed change meets the requirements 
of 10 CFR 50.92(c) and does

[[Page 47700]]

not involve a significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration. Local 
Public Document Room location: Houston-Love Memorial Library, 212 W. 
Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302. Attorney 
for licensee: M. Stanford Blanton, Esq., Balch and Bingham, Post Office 
Box 306, 1710 Sixth Avenue North, Birmingham, Alabama 35201. NRC 
Project Director: Herbert N. Berkow.

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

    Date of amendment request: June 20, 1997 (TS-97-004).
    Description of amendment request: The proposed amendment would be 
an administrative change that would revise the analytical methodology 
used to determine the low temperature overpressure protection (LTOP) 
event heatup and cooldown curves. This revised methodology would be 
incorporated by reference in the Watts Bar Nuclear Plant (WBN), Unit 1 
Technical Specification (TS) 5.9, ``Reporting Requirements,'' Section 
5.9.6, ``Reactor Coolant System (RCS) Pressure and Temperature Limits 
Report (PTLR),'' upon approval for use by the U.S. Nuclear Regulatory 
Commission (NRC). The revised methodology extends the current LTOP 
requirements through the end of 7 effective full power years (EFPY). 
The only technical change being proposed is the substitution of the 7 
EFPY American Society of Mechanical Engineering (ASME), Appendix G, 
heatup and cooldown curves adjusted by ASME Code Case N-514, ``Low 
Temperature Overpressure Protection'' in place of the current 1.5 EFPY 
curves as the bounding curves for the LTOP setpoints. This change will 
not impact the current 10 CFR 50, Appendix G, pressure/temperature (P/
T) limit curves used for heatup and cooldown that are based on 7 EFPY.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The Nuclear Regulatory Commission has provided standards for 
determining whether a significant hazards consideration exists (10 
CFR 50.92). A proposed amendment to an operating license for a 
facility involves no significant hazards consideration if operation 
of the facility, in accordance with the proposed amendment, would 
not: (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated: or (2) create the 
possibility of a new or different kind of accident from any accident 
previously evaluated: or (3) involve a significant reduction in a 
margin of safety. Each standard is discussed below for the proposed 
amendment.
    (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 LTOP setpoints (identified as the cold overpressure 
mitigation system (COMS) for WBN), adjusted for instrument 
inaccuracy, pressure differential, and setpoint overshoot by the 
scaling and setpoint documents (SSDs), ensure that the 10 CFR 50, 
Appendix G P/T [pressure and temperature] limits based on 7 EFPY are 
not exceeded by more than the provisions of ASME Code Case N-514, 
and therefore, ensure that the RCS integrity is maintained.
    The change does not modify the RCS pressure boundary, nor make 
any physical changes to the facility design, material, construction 
standards, or setpoints. The LTOP enabling temperature based on TS 
3.4.12, ``Cold Overpressure Mitigation System (COMS),'' is [less 
than or equal to] 350 degrees F and is more conservative than a 
value of 271.1 degrees F (RT<INF>NDT </INF>+ 90 degrees F) based on 
7 EFPY. This temperature would be acceptable based on NRC Branch 
Technical Position-Reactor Systems Branch (BTP-RSB)-5.2, 
``Overpressurization Protection of Pressurized Water Reactors While 
Operating at Low Temperatures.'' The LTOP enabling temperature 
remains unchanged by this proposed amendment. The probability of a 
LTOP event occurring is independent of the P/T limits for the RCS 
pressure boundary; therefore, the probability of an LTOP event 
occurring remains unchanged.
    The calculation of the P/T limits in accordance with approved 
regulatory methods based on 7 EFPY provides assurance that reactor 
pressure vessel fracture toughness requirements are met and the 
integrity of the RCS pressure boundary is maintained. LTOP setpoints 
based on 1.5 EFPY P/T limits have provided margin such that a 
pressure excursion exceeding the 7 EFPY limits would not exceed the 
1.5 EFPY limits. This margin between the 7 EFPY curves and the LTOP 
setpoints is maintained by changing the bounding curves for the LTOP 
setpoints to 7 EFPY curves adjusted by the provisions of ASME Code 
Case N-514. The only technical change being made is the bounding 
curves which provide the basis for the current LTOP setpoints.
    The use of theoretical fluence for generating the P/T curves to 
be used for the first 7 EFPY is appropriate and was submitted July 
31, 1995, with the WBN Unit 1 PTLR, Revision 4 and WCAP-13829, 
Revision 2, ``Heatup and Cooldown Limit Curves for Normal Operation 
for Watts Bar Unit 1.'' The present 7 EFPY curves are generated 
using a theoretical value for fluence calculated by Westinghouse in 
accordance with NRC approved methodology since WBN had no 
surveillance capsule data available at the time of plant startup. 
This value for fluence is conservative, and the actual fluence to 
the intermediate shell forging (the controlling beltline material) 
is expected to be significantly less than the theoretical value used 
to generate the initial 7 EFPY curves since WBN is transitioning to 
a low-leakage core. The LTOP bounding curves are based on 7 EFPY 
curves adjusted in accordance with ASME Code Case N-514 which were 
generated using the same theoretical fluence as used for the P/T 
curves. The significance of using the theoretical value of fluence 
in generating these curves is the additional margin that exists 
between the 7 EFPY theoretical curves and curves that would be 
generated using actual fluence values from capsule data. This 
additional margin reduces the significance of changing the LTOP 
basis from the 1.5 EFPY curves to the 7 EFPY curves adjusted for 
ASME Code Case N-514.
    This change does not adversely affect the integrity of the RCS 
such that its function in the control of radiological consequences 
is affected. In addition, the change does not affect any fission 
barrier. The change does not degrade or prevent the LTOP power 
operated relief valves (PORVs) or other safety related systems from 
responding to accidents described in Chapter 15 of the Final Safety 
Analysis Report (FSAR). In addition, the change does not alter any 
assumptions previously made in the radiological consequences of an 
accident described in the FSAR. Therefore, the consequences of an 
accident previously evaluated in the FSAR are not increased. Thus, 
the operation of WBN Unit 1 in accordance with this proposed 
amendment does not involve a significant increase in the probability 
or consequences of any 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 Appendix G P/T limitations were prepared using methods 
derived from the ASME Boiler and Pressure Vessel Code Section III 
and the criteria set forth in NRC Regulatory Standard Review Plan 
5.3.2, ``Pressure-Temperature Limits.'' The use of ASME Code Case N-
514 and the theoretical fluence value for 7 EFPY does not modify the 
RCS pressure boundary, nor make any physical changes to the LTOP 
setpoints or system design. The proposed change was prepared in 
accordance with regulatory requirements and provides evaluation of 
LTOP events based on 7 EFPY theoretical fluence which is more 
limiting than actual expected neutron exposure for that same period.
    This proposed change is an administrative change which 
incorporates by reference the use of an NRC approved methodology; 
therefore, the change does not cause the initiation of any accident 
nor create any new creditable limiting failure for safety-related

[[Page 47701]]

systems and components. The change does not result in an event 
previously deemed incredible being made credible. As such, it does 
not create the possibility of an accident different than any 
evaluated in the FSAR.
    The change does not have any effect on the ability of the 
safety-related systems to perform their intended safety functions. 
The change does not create failure modes that could adversely impact 
safety-related equipment. Therefore, it will not create the 
possibility of a malfunction of equipment important to safety 
different than previously evaluated in the FSAR. Thus, the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different type 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 margin of 
safety.
    The 10 CFR 50, Appendix G P/T limitations were prepared using 
methods derived from ASME Section III and criteria set forth in NRC 
Regulatory Standard Review Plan 5.3.2. These documents along with 
the calculational limitations specified in 10 CFR 50.61 are an 
acceptable method for implementing the requirements of 10 CFR 50 
Appendices G and H. Inherent conservatisms in the P/T limits 
resulting from these documents include:
    a. An assumed defect in the reactor vessel wall with a depth 
equal to \1/4\ of the thickness (T) of the vessel wall and a length 
equal to 1\1/2\ times the thickness of the vessel wall.
    b. Assumed reference flaw oriented in both longitudinal and 
circumferential directions and limiting material property. At WBN, 
the only weld in the core region is oriented in the circumferential 
direction.
    c. A factor of safety of 2 is applied to the membrane stress 
intensity factor.
    d. The limiting toughness is based upon a reference value 
(K<INF>IM</INF>) which is the lower bound of the dynamic crack 
initiation and arrest toughness.
    e. A 2-sigma margin term is applied in determining the adjusted 
reference temperature (ART) that is used in calculating the limiting 
toughness.
    Beyond the conservatisms described above, WBN has the following 
additional margin:
    a. The value of fluence used in the calculation of the WBN Unit 
1 Appendix G P/T limits is a theoretical value calculated by NRC 
approved methodology.
    b. The ART for 7 EFPY is based on the theoretical value for 
fluence and therefore is conservative. The LTOP enabling temperature 
of [less than or equal to] 350 degrees F in accordance with TS 
3.4.12 is conservative with respect to (RT<INF>NDT</INF> + 90 
degrees F) which based on an ART of 181.1 degrees F would equal 
271.1 degrees F. An enabling temperature of (RT<INF>NDT</INF> + 90 
degrees F) is based on NRC BTP-RSB 5.2.
    The ASME Working Group for Operating Plant Criteria developed 
Code Case N-514 as an alternative methodology to the safety margin 
requirements of Appendix G to 10 CFR 50. The Code Case provides 
criteria to determine pressure limits during LTOP events that avoid 
certain operational restrictions, provide adequate margins against 
failure of the reactor vessel, and reduce the potential for 
unnecessary activation of the relief valves used for LTOP. 
Specifically, the N-514 Code Case allows determination of the LTOP 
setpoints such that for LTOP events the maximum pressure in the 
reactor vessel would not exceed 110% of the P/T limits of the 
existing ASME Appendix G curves, and redefines the enabling 
temperature as a coolant temperature less than 200 degrees F or a 
reactor vessel metal temperature less than RT<INF>NDT</INF> + 50 
degrees F. Code Case N-514 has been approved by the ASME Code 
Committee and its content has been incorporated in Appendix G of 
ASME Section XI and published in the 1993 Addenda and 1995 Edition. 
Code Case N-514 has not been approved for use in Regulatory Guide 
1.147, ``Inservice Inspection Code Case Acceptability, ASME Section 
XI;'' however, it has been included in the Draft Regulatory Guide 
1.147 (Task DG-1050) which is currently out for public review and 
comment. As stated above, WBN Unit 1 uses Appendix G for the P/T 
limits for plant operation and an LTOP enabling temperature greater 
than RT<INF>NDT</INF> + 90 degrees F which is more conservative than 
the alternative methodology contained in Code Case N-514.
    The need for implementation of Code Case N-514 at WBN involves 
the avoidance of certain operational restrictions associated with 
low temperature operation of the plant. Use of Appendix G P/T limits 
to determine the PORV setpoints would result in pressure setpoints 
within the operating window; consequently, no margin would be 
available for normal operating pressure surges. Therefore, operating 
with these limits could result an unnecessary challenge to the PORVs 
and cavitation of the reactor coolant pumps (RCP) during normal 
operation. Additionally, the need to raise the RCS inventory by 
external heating methods to a temperature high enough to avoid PORV 
activation when starting a RCP from a RCS cold shutdown condition 
could result in undesirable thermal transients in the RCS.
    Utilizing the methodology set forth in the ASME Boiler and 
Pressure Vessel Code Section XI, Appendix G, which includes the 
provisions of Code Case N-514, NRC Regulatory Standard Review Plan 
5.3.2, 10 CFR 50.61, and 10 CFR 50, Appendices G and H with the 
above additional margins ensures that proper limits and conservative 
safety factors are maintained. Thus the proposed change does not 
significantly reduce the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    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 10H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendments: March 18, 1997.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) to increase the High Pressure Coolant 
Injection (HPCI)

[[Page 47702]]

system low pressure isolation setpoint from greater than 80 psig to 
greater than 100 psig.
    Date of issuance: August 21, 1997.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 161, 156.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17228).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 21, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.

Duke Power Company, 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: June 12, 1997.
    Brief description of amendments: The amendments change the name 
``Duke Power Company'' to ``Duke Energy Corporation'' in the Catawba 
operating licenses and appendices as a result of Duke Power Company's 
recent name change.
    Date of issuance: August 22, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 161 and 153.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: July 2, 1997 (62 FR 
35848).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 22, 1997, and an Environmental 
Assessment dated July 31, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.

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

    Date of application for amendments: June 12, 1997.
    Brief description of amendments: The amendments change the name 
``Duke Power Company'' to ``Duke Energy Corporation'' in the McGuire 
operating licenses and appendices as a result of Duke Power Company's 
recent name change.
    Date of issuance: August 26, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 176 and 158.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Licenses.
    Date of initial notice in Federal Register: July 2, 1997 (62 FR 
35848).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 26, 1997. An Environmental 
Assessment was issued and dated August 15, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, Charlotte, North Carolina.

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

    Date of amendment request: October 26, 1995 and supplemented by 
letters dated April 7 and July 30, 1997.
    Brief description of amendment: The amendment revised the technical 
specifications for 16 editorial changes and deletes the reuirement for 
a program to prevent and detect Asiatic Clams (Corbicula) in the 
service water system (SWS). The Corbicula program is no longer needed 
because the facility has been modified and SWS no longer takes water 
from the Mississippi River; source of the larvae and infestation.
    Date of issuance: August 26, 1997.
    Effective date: August 26, 1997.
    Amendment No.: 95.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 6, 1995 (60 FR 
62492).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 26, 1997.
    No significant hazards consideration comments received. No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803.

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

    Date of amendment request: November 15, 1996, as supplemented May 9 
and August 15, 1997.
    Brief description of amendment: The amendment revises the technical 
specifications to increase the two recirculation loop Minimum Critical 
Power Ratio (MCPR) from 1.07 to 1.10 and the single recirculation loop 
MCPR limit from 1.08 to 1.12.
    Date of issuance: August 26, 1997.
    Effective date: August 26, 1997.
    Amendment No.: 96.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 2, 1997 (62 FR 
127).
    The May 9 and August 15, 1997, submittal provided clarifying 
information that did not change the initial no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 26, 1997.
    No significant hazards consideration comments received. No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803.

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

    Date of amendment request: January 20, 1997 as supplemented by 
letter dated July 7, 1997.
    Brief description of amendment: The amendment revises the technical 
specifications to allow the use of flow control spectral shift 
strategies to increase cycle energy. The revision is based on a Maximum 
Extended Load Line Limit (MELLL) analysis for the River Bend Station.
    Date of issuance: August 26, 1997.
    Effective date: August 26, 1997.
    Amendment No.: 97.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications/operating license.
    Date of initial notice in Federal Register: February 26, 1997 (62 
CFR 8799).
    The July 7, 1997 submittal provided clarifying information and did 
not change the initial no significant hazards consideration 
determination.

[[Page 47703]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 26, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803.

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

    Date of amendment request: November 6, 1996, as supplemented by 
letter dated July 31, 1997.
    Brief description of amendment: The amendment revises the Technical 
Specifications to delete the requirement for the Penetration Valve 
Leakage Control System. The licensee requested deferal of the proposal 
to increase the allowed leakage by main steam isolation valves and to 
delete the requirement for the Main Steam Positive Leakage Control 
System.
    Date of issuance: August 26, 1997.
    Effective date: August 26, 1997.
    Amendment No.: 98.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 2, 1997 (62 FR 
125).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 26, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803.

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

    Date of application for amendment: October 22, 1996, as 
supplemented by letter dated June 26, 1997.
    Brief description of amendment: The amendment revises Figure 
3.4.11-1, ``Minimum Reactor Vessel Metal Temperature vs. Reactor Vessel 
Pressure,'' in Limiting Condition for Operation 3.4.11, ``RCS [Reactor 
Coolant System] Pressure and Temperature (P/T) Limits,'' of the 
Technical Specifications. The previous figure was only up to 10 
Effective Full Power Years (EFPYs) and this amendment revises the 
figure up to 32 EFPYs. There are now five curves of Figure 3.4.11-1 for 
five different EFPY periods: up to 16, 16 to 20, 20 to 24, 24 to 28, 
and 28 to 32. The licensee submitted two sets of curves. The first set 
replaced TS Figure 3.4.11-1. The second set were duplicates of the 
first set except the second set also contained detailed information 
used in development of the curves and would be included in the next 
update of the Updated Final Safety Analysis Report. There were also 
minor additions to Surveillance Requirements (SRs) 3.4.11.1 and 
3.4.11.2 to have the SRs reference the ``applicable Figure 3.4.11-1 
based on the current effective full power year (EFPY).''
    Date of issuance: August 27, 1997.
    Effective date: August 27, 1997.
    Amendment No: 132.
    Facility Operating License No. NPF-29: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 26, 1997 (62 
FR 8797).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 27, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120.

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

    Date of amendment request: April 11, 1997.
    Brief description of amendment: The amendment modifies Technical 
Specifications 3.3.3.7.3, and Surveillance Requirements (SR) 4.3.3.7.3 
for the broad range gas detection system. Also it includes some changes 
to the Bases in Section 3/4.3.3.7 to incorporate information associated 
with the proposed modifications. The licensee is planning to replace 
the existing toxic gas monitors in the system with a new, more advanced 
gas monitors.
    Date of issuance: August 19, 1997.
    Effective date: August 19, 1997, to be implemented within 90 days.
    Amendment No.: 133.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 7, 1997 (62 FR 
24987)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 19, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric 
Authority of Georgia, City of Dalton, Georgia, Docket No. 50-321, Edwin 
I. Hatch Nuclear Plant, Unit 1, Appling County, Georgia

    Date of application for amendment: April 29, 1997, as supplemented 
by letter dated May 28, 1997.
    Brief description of amendment: The amendment revises Hatch Unit 1 
reactor vessel pressure and temperature limits to reflect data 
collected from the material sample recovered during the March 1996 Unit 
1 outage.
    Date of issuance: August 19, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 207.
    Facility Operating License No. DPR-57: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 16, 1997 (62 FR 
38138).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 19, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513.

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

    Date of application for amendment: October 4, 1996, as supplemented 
June 10 and August 15, 1997 (TSCR 250).
    Brief description of amendment: The amendment changes the Safety 
Limit Minimum Critical Power Ratio and as a result, the operating 
Minimum Critical Power Ratio. The amendment also capitalized certain 
definitions and provided a uniform type font for Sections 2.1 and 3.10.
    Date of Issuance: August 26, 1997.
    Effective date: August 26, 1997, with full implementation within 30 
days.
    Amendment No.: 192.
    Facility Operating License No. DPR-16. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 6, 1996 (61 FR 
57484).
    The Commission's related evaluation of this amendment is contained 
in a

[[Page 47704]]

Safety Evaluation dated August 26, 1997.
    The June 10 and August 15, 1997, submittals provided clarifying 
information that did not alter the staff's initial proposed no 
significant hazards considerations determination.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

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

    Date of application for amendment: April 21, 1997, as supplemented 
July 17, 1997.
    Brief description of amendment: The amendment reduces the required 
volume of borated water in each core flood tank from 1040 ft \3\ to 940 
ft \3\, reduces the required high pressure injection pump flowrate from 
500 gallons per minute (gpm) to 431 gpm, and deletes the local manual 
valve operability option for decay heat system valves DH-V-6A and DH-V-
6B.
    Date of issuance: August 27, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 203.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 21, 1997 (62 FR 
27795).
    The July 17, 1997, submittal provided clarifying information that 
did not alter the initial no significant hazards determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated August 27, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, Docket 
Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda 
County, Texas

    Date of amendment request: April 22, 1997.
    Brief description of amendments: The proposed amendment revised 
Technical Specifications 5.3.1, Fuel Assemblies, and 6.9.1.6, Core 
Operating Limits Report, to allow use of an alternate zirconium-based 
fuel cladding, ZIRLO, and limited substitution of fuel rods by ZIRLO 
filler rods.
    Date of issuance: August 19, 1997.
    Effective date: August 19, 1997.
    Amendment Nos.: Unit 1--Amendment No. 89; Unit 2--Amendment No. 76.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 21, 1997 (62 FR 
27795).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 19, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

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

    Date of application for amendment: November 20, 1995.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TSs) by providing clarifications to the applicability 
and action statements in TS Table 3.3-12 relating to the Steam 
Generator Blowdown Monitor and the Condensate Polishing Facility Waste 
Neutralizing Sump radiation monitor.
    Date of issuance: August 26, 1997.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 207.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65683).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 26, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385.

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

    Date of application for amendment: May 1, 1997.
    Brief description of amendment: Technical Specifications 3/4.8.2.2 
and 3/4.8.3.2 specify which electrical power systems are required to be 
operable in Modes 5 and 6. The amendment clarifies the requirements by 
identifying the specific equipment required and their alignments in 
Modes 5 and 6.
    Date of issuance: August 21, 1997.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 146.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 4, 1997 (62 FR 
30637).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 21, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385.

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

    Date of application for amendment: May 5, 1997.
    Brief description of amendment: Technical Specification 
Surveillance 4.5.2.b.1 requires that the emergency core cooling system 
piping be verified full of water at least once per 31 days. The 
amendment revises the surveillance to exempt the operating charging 
pump(s) and associated piping from the requirement to be verified full 
of water and moves the description of the verification method from the 
surveillance to the Bases section.
    Date of issuance: August 28, 1997.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 147.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 4, 1997 (62 FR 
30638).
    The Commission's related evaluation of the amendment is contained 
in a

[[Page 47705]]

Safety Evaluation dated August 28, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope.

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 31, 1997, as supplemented 
by letter dated June 25, 1997.
    Brief description of amendments: These amendments extend the APRM 
flow bias instrumentation surveillance interval from 18 months to 24 
months. This will eliminate the need to perform on-line APRM 
surveillance testing, which requires plant operators to place an 
operating unit in a half scram configuration.
    Date of issuance: August 19, 1997.
    Effective date: Units 2 and 3 effective as of date of issuance.
    Amendments Nos.: 219 and 222.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 7, 1997 (62 FR 
24988).
    The supplemental 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 August 19, 1997.
    No significant hazards consideration comments received: No.
    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.

Southern Nuclear Power Company, Inc., Georgia Power Company, Oglethorpe 
Power Corporation, Municipal Electric Authority of Georgia, City of 
Dalton, Georgia, Docket Nos. 50-424 and 50-425, Vogtle Electric 
Generating Plant, Units 1 and 2, Burke County, Georgia

    Date of application for amendments: June 13, 1997, as supplemented 
by letter dated July 18, 1997.
    Brief description of amendments: The amendments revise the 
pressurizer safety relief valve setpoint specified in Technical 
Specification 3.4.10.
    Date of issuance: August 26, 1997.
    Effective date: As of the date of issuance to be implemented for 
Unit 1 prior to or after initial entry into Mode 3 (in accordance with 
the provisions of the note to the Applicability for LCO 3.4.10) 
following the fall 1997 refueling outage; for Unit 2 prior to or after 
initial entry into Mode 3 (in accordance with the provisions of the 
note to the Applicability for LCO 3.4.10) following the spring 1998 
refueling outage.
    Amendment Nos.: 98 and 76.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 16, 1997 (62 FR 
38139).
    The supplemental material did not change the no significant hazards 
finding or expand the scope of the Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 26, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830.

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

    Date of amendment request: December 7, 1994 (TXX-94326), as 
supplemented by letter dated June 21, 1996 (TXX-96384).
    Brief description of amendments: These changes revised Section 
3.7.1.5 of the Technical Specification to increase the Allowed Outage 
Time for one inoperable Main Steam Isolation Valve (MSIV) while in Mode 
1, and to clarify requirements related to inoperable MSIVs while in 
Modes 2 and 3.
    Date of issuance: August 18, 1997.
    Effective date: August 18, 1997, to be implemented within 60 days.
    Amendment Nos.: 54 and 40.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 1995 (60 FR 
6312).
    The additional information contained in the supplemental letter 
dated June 21, 1996, was clarifying in nature and thus, within the 
scope of the initial notice and did not affect the staff's proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 18, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019.

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

    Date of application for amendments: April 14, 1997 (TSCR 197), as 
supplemented on August 11, 1997.
    Brief description of amendments: These amendments revise Technical 
Specifications (TS) Sections 15.6.2, ``Organization,'' TS 15.6.5.1, 
``Manager's Supervisory Staff,'' TS 15.6.6, ``Reportable Event 
Action,'' TS 15.6.7, ``Actions To Be Taken If A Safety Limit Is 
Exceeded,'' and TS 15.7.8, ``Administrative Controls,'' by changing the 
title of the corporate officer responsible for nuclear operations from 
the ``Vice President-Nuclear Power,'' to the ``Chief Nuclear Officer.''
    Date of issuance: August 25, 1997.
    Effective date: August 25, 1997, with full implementation within 45 
days.
    Amendment Nos.: 177 and 181.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 21, 1997 (62 FR 
27802), as corrected May 29, 1997 (62 FR 29163) The August 11, 1997, 
submittal provided a corrected TS page. This information was within the 
scope of the action noticed and did not change the staff's initial 
proposed no significant hazards considerations determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 25, 1997.
    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 3rd day of September 1997.
    For The Nuclear Regulatory Commission.
Bruce E. Boger,
Director, Division of Reactor Projects--I/II Office of Nuclear Reactor 
Regulation.
[FR Doc. 97-23820 Filed 9-9-97; 8:45 am]
BILLING CODE 7590-01-P