[Federal Register Volume 61, Number 187 (Wednesday, September 25, 1996)]
[Notices]
[Pages 50338-50351]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-10925]


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NUCLEAR REGULATORY COMMISSION

Biweekly Notice


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189

[[Page 50339]]

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 August 30, 1996, through September 13, 1996. 
The last biweekly notice was published on September 11, 1996.

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 
Review 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 25, 1996, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The

[[Page 50340]]

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: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. 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.

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

    Date of amendment request: August 2, 1996
    Description of amendment request: The proposed amendment would 
eliminate from the licenses the requirement to conduct corrosion 
testing for the laser welded steam generator sleeves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This proposed change only involves deleting laboratory testing 
requirements designed to demonstrate service life of laser welded 
sleeved tubes in the presence of a crevice. Current inspection 
requirements ensure that premature degradation is identified and 
that tubes containing degraded sleeve joints are plugged. 
Operational primary-to-secondary leakage limits ensure that 
appropriate action is taken if sleeve degradation results in 
leakage. These actions will ensure that offsite dose will be 
maintained within a small percentage of 10 CFR 100 limits. Failure 
of a sleeve joint is bounded by the Steam Generator Tube Rupture 
event evaluated in the [Updated Final Safety Analysis Report] UFSAR. 
Therefore, the laboratory testing to determine service life of 
sleeved tube joints in the presence of a crevice does not provide 
any further useful data. The change does not result in the 
installation of any new equipment, and no existing equipment is 
modified.
    Therefore, this proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This proposed change only addresses deleting the laboratory 
testing requirements designed to demonstrate service life of sleeved 
tubes in the presence of a crevice. Sleeved tubes will continue to 
be inspected and plugged in accordance with existing requirements 
which are sufficient to ensure detection and repair of degraded 
tubes. Premature degradation of tubes is addressed through primary-
to-secondary leakage monitoring and leakage limits. No new equipment 
is being installed and no existing equipment is being modified by 
this proposed change. Also, no new system configurations will be 
introduced as a result of this proposed change. Therefore, no new or 
different failure modes are being introduced by deleting the 
laboratory testing.
    Thus, this 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 of safety.
    This proposed change only involves deleting laboratory testing 
requirements designed to demonstrate service life of sleeved tubes 
in the presence of a crevice. Sleeve integrity will be monitored 
during the operating cycle through the current primary-to-secondary 
leakage monitoring program. In the event of premature degradation of 
a sleeve joint that results in tube leakage, plant shutdown will 
occur as required by Technical Specifications and administrative 
requirements in accordance with approved plant procedures. Sleeved 
tubes will be monitored for degradation in accordance with the 
existing inservice inspection requirements which monitors a minimum 
20 percent random sleeve sample size. Any tubes with defective 
sleeve joints will be plugged as required by Technical 
Specifications. Service life of sleeved tubes in the presence of a 
crevice, as predicted by laboratory testing, does not affect the 
margin of safety of the plant. Therefore, this proposed change does 
not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: July 15, 1996
    Description of amendment request: The proposed amendments would 
revise Technical Specifications (TS) and associated Bases to relocate 
the fire protection program elements from the TS to the Fire Protection 
Program. The affected TS sections are 3/4.3.7.9, ``Fire Detection 
Instrumentation;'' 3/4.7.5, ``Fire Suppression Systems;'' 3/4.7.6, 
``Fire Rated Assemblies;'' and 6.1.C.4,

[[Page 50341]]

``Fire Brigade Staffing.'' In addition, the amendments revise the 
Operating License to replace existing fire protection license 
conditions with the NRC's standard fire protection license condition. 
These changes are made in accordance with the guidance provided in 
Generic Letter (GL) 86-10, ``Implementation of Fire Protection 
Requirements,'' and GL 88-12, ``Removal of Fire Protection Requirements 
from Technical Specifications.'' Also, the May 19, 1995, proposed 
revision to remove the fire protection requirements from the TS (60 FR 
35067) is withdrawn.
    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 because:
    This amendment request does not involve any actual changes to 
the fire protection systems at the station. It involves an 
administrative change which relocates the control of the Fire 
Protection Program from each unit's operating license and technical 
specifications to the station Fire Protection Program, as suggested 
in Generic Letters 86-10 and 88-12. Therefore, the relocation of 
these controls does not affect the assumptions for any of the 
accident analysis contained in Chapter 15 of the [Updated Final 
Safety Analysis Report] UFSAR.
    The Fire Protection Technical Specifications which are to be 
relocated to the Fire Protection Program will be controlled by the 
proposed fire protection license condition and 10CFR 50.59. These 
controls ensure that the requested changes maintain the same level 
of control for the Fire Protection Program as that which currently 
exists in the Technical Specifications. Therefore, this change is 
administrative in nature and does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    This amendment request does not involve any physical changes to 
the fire protection systems or reduce the level of control of the 
Fire Protection Program. It therefore does not create the 
possibility of a new or different type of accident than any 
previously described in the UFSAR.
    3) Involve a significant reduction in the margin of safety 
because:
    The same level of control which is currently applied to the Fire 
Protection Program by the limiting conditions for operation and the 
surveillance requirements of the technical specifications will be 
included in the controls applied by the unit licenses and the Fire 
Protection Program. Therefore, the margin of safety as defined in 
the technical specification bases will not be reduced by this 
proposed amendment.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

    Date of amendment request: July 26, 1996, and supplemented on 
September 3, 1996
    Description of amendment request: The proposed amendments would 
allow licensee control of the reactor coolant system (RCS) pressure and 
temperature (P/T) limits for heatup, cooldown, low temperature 
operation and hydrostatic testing. They would also revise the reactor 
vessel material surveillance program specimen withdrawal schedule such 
that the Unit 2 removal of capsule X is delayed until 19 Effective Full 
Power Years (EFPY). This change affects the schedule for withdrawing 
surveillance capsules from the reactor vessel for testing to measure 
the impact of neutron irradiation of the vessel material and is 
required by Section III.B.3 of 10 CFR Part 50, Appendix H, ``Reactor 
Vessel Material Surveillance Program Requirements.'' The schedule must 
be approved by the Nuclear Regulator Commission (NRC) before 
implementation.
    Based on input from the Babcock and Wilcox Owners Group Reactor 
Vessel Working Group, the data from Zion, Unit 2, capsule X would be 
more useful in the overall Master Integrated Reactor Vessel 
Surveillance Program (MIRVP) context if irradiated to the ASTM E185-82 
maximum of twice the peak End Of Life (EOL) vessel fluence, because 
data at higher fluences is needed to characterize irradiation behavior 
at the higher EOL fluences characteristic of other non-Commonwealth 
Edison MIRVP vessels. For this reason, the licensee is proposing 
withdrawing and testing Zion, Unit 2, capsule X at 19 EFPY, which is 
currently estimated to occur at refueling outage Z2R18, in the year 
2002.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change revises the 10 CFR 50, Appendix H reactor 
vessel material specimen withdrawal schedule. Neither the specimens, 
nor the process of withdrawal of the specimens, are considered as 
initiators for any previously evaluated accident. Further, data at 
all fluence levels of current interest based on ASTM E185-82 has 
already been obtained from seven Zion Unit 1 and 2 capsules which 
have been tested, and the existing evaluations show the reactor 
vessel fracture toughness properties to be as expected, and 
providing the required safety margin. Extending the time for 
withdrawal of the specimen does not adversely affect the pressure 
and temperature limit curves for the reactor vessel. Regulatory 
Guide 1.99, Rev. 2, was used to prepare the conservative pressure 
and temperature limit curves which continue to be requirements.
    Additionally, Zion Station participates in the B&W Owners Group 
Reactor Vessel Working Group designed to significantly increase the 
amount of PWR surveillance data. Under this Working Group, Zion 
Station data contributes to the overall understanding of reactor 
vessel material irradiation behavior at high EOL fluences, and 
obtains the benefit of data from other plants. This program 
complements the Zion Station program so that postponement of the 
specimen withdrawal will have minimal impact on the understanding of 
the irradiation effects on the Zion Station reactor vessel. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed revision to the specimen withdrawal schedule does 
not change the system operation or design, and therefore, does not 
change the response of any required structures, systems or 
components in the mitigation of any evaluated accident. As such, 
this change does not involve a significant increase in the 
consequences of an accident previously evaluated.
    The proposed change relocates the RCS P/T, LTOP [low-temperature 
overpressure protection] limitations, and supporting information 
from the Technical Specifications to Licensee control, specifically 
a Pressure Temperature Limits Report (PTLR). Compliance with these 
limitations will continue to be required by the Technical 
Specifications, however the limitations themselves will be relocated 
to a Licensee controlled document. Changes to these limitations will 
be controlled by Section 5.6.6 of the Technical Specifications. 
Changes to the RCS P/T limits can only be made in accordance with 
the approved methodologies listed in the Technical Specifications 
which will, in combination with the limitations that continue to be

[[Page 50342]]

imposed by the Technical Specifications, continue to assure the 
function of the reactor vessel as a pressure boundary. Revisions to 
the LTOP limits can only be made in accordance with the approved 
methodologies listed in the Technical Specifications, with any 
resulting setpoint changes controlled through a process which 
utilizes 10 CFR 50.59. Therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not necessitate a physical alteration 
of the plant (no new or different equipment will be installed). The 
proposed revision to the specimen withdrawal schedule does not 
change the system operation or design, and therefore, does not 
introduce any new failure mechanisms. The proposed specimen 
withdrawal schedule continues to provide the required data for 
subsequent reactor vessel evaluations, and previous data has 
confirmed the confidence in the integrity of the reactor vessel well 
beyond the completion of the evaluations following the proposed 
withdrawal. Therefore, this revision to the withdrawal schedule does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not necessitate a physical alteration 
of the plant (no new or different equipment will be installed). The 
Technical Specifications will continue to retain requirements to 
maintain the RCS within acceptable operational limitations and to 
assure operability of the LTOP system. As such, the Technical 
Specifications will continue to require compliance with these 
limitations. Thus, this change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change to the specimen withdrawal schedule will not 
result in a significant reduction in a margin of safety because it 
has no impact on any safety analysis assumptions. Additionally, data 
at all fluence levels of current interest based on ASTM E185-82 has 
already been obtained with the seven Zion Unit 1 and 2 capsules 
which have been tested, and the existing evaluations show the 
reactor vessel fracture toughness properties to be as expected, and 
providing the required safety margin. The current pressure and 
temperature limits are conservative and also provide sufficient 
margin to ensure the integrity of the reactor vessel. The proposed 
change to the withdrawal schedule does not adversely impact these 
curves. Therefore, this change does not involve a significant 
reduction in a margin of safety.
    The proposed change will not result in a significant reduction 
in a margin of safety because it has no impact on any safety 
analysis assumptions. Any future changes to the RCS P/T, LTOP 
limits, or supporting information must be performed in accordance 
with approved NRC methodologies, and compliance with the limitations 
relocated to the PTLR will continue to be required by the Technical 
Specifications. Additionally, any revision to the LTOP limits which 
result in setpoint changes will be controlled through a process 
which utilizes 10 CFR 50.59. 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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: September 5, 1996 (NRC-96-0075)
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) sections 2.1.2 and 3.4.1.1 to 
incorporate cycle-specific safety limit minimum critical power ratios 
(SLMCPRs) for the core that will be loaded during the upcoming 
refueling outage expected to commence in November 1996.
    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 derivation of the revised SLMCPRs for Fermi 2 for 
incorporation into the TS, and its use to determine cycle-specific 
thermal limits, have been performed using NRC-approved methods. 
Additionally, interim implementing procedures, which incorporate 
cycle-specific parameters, have been used which result in a more 
restrictive value for the SLMCPR. These calculations do not change 
the method of operating the plant and have no effect on the 
probability of an accident initiating event or transient. The basis 
of the MCPR Safety Limit is to ensure that no mechanistic fuel 
damage is calculated to occur if the limit is not violated. The new 
SLMCPRs preserve the existing margin to transition boiling and the 
probability of fuel damage is not increased. Therefore, the proposed 
TS change does not involve an increase in the probability or 
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 proposed change results from analysis of the Cycle 6 core 
reload using the same fuel types as previous cycles. These changes 
do not involve any new method for operating the facility and do not 
involve any facility modifications. No new initiating events or 
transients result from these changes. Therefore, the proposed TS 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The margin of safety as defined in the TS Bases will remain the 
same. The new SLMCPRs are calculated using NRC-approved methods 
which are in accordance with the current fuel design and licensing 
criteria. Additionally, interim implementing procedures, which 
incorporate cycle-specific parameters, have been used. The MCPR 
Safety Limit remains high enough to ensure that greater than 99.9% 
of all fuel rods in the core will avoid transition boiling if the 
limit is not violated, thereby preserving the fuel cladding 
integrity. Therefore, the proposed TS change does not involve a 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226
    NRC Project Director: John Hannon

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf 
Nuclear Station, Unit 1, Claiborne County, Mississippi

    Date of amendment request: July 31, 1996, as supplemented by letter 
dated September 5, 1996. These letters supersede the application 
submitted in letter dated May 9, 1996, which was noticed in the Federal 
Register on June 5, 1996 (61 FR 28614).
    Description of amendment request: The amendment request would (1) 
increase the safety limit minimum critical power ratio (MCPR) for two 
loop operation and single loop operation to 1.12 and 1.14, 
respectively, and (2) add a General Electric topical report to the list 
of documents describing the analytical methods used to determine the 
core operating limits. The proposed changes are to Section 2.1.1, 
Reactor

[[Page 50343]]

Core Safety Limits, and Section 5.6.5, Core Operating Limits Report 
(COLR), respectively, of the Technical Specifications (TSs). This 
amendment would go into effect in Operating Cycle 9, at the end of the 
upcoming Refueling Outage 8, and the plant will have a mixed core of 
Siemens Power Corporation (SPS) 9x9-5 and General Electric (GE) GE11 
reload fuel. The licensee also proposed changes to the Bases of the TSs 
associated with the above proposed changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    I. The proposed change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    The Minimum Critical Power Ratio (MCPR) safety limit is defined 
in the Bases to Technical Specification 2.1.1 as that limit which 
``ensures that during normal operation and during Anticipated 
Operational Occurrences (AOOs), at least 99.9% of the fuel rods in 
the core do not experience transition boiling.'' The MCPR safety 
limit is re-evaluated for each reload and, for GGNS [Operating] 
Cycle 9, the analyses have concluded that a two-loop MCPR safety 
limit of 1.12 based on the application of the generic GE MCPR 
methodology is necessary to ensure that this acceptance criterion is 
satisfied. For single-loop operation, a MCPR safety limit of 1.14 
based on the generic GE MCPR methodology was determined to be 
necessary. Core MCPR operating limits are developed to support the 
Technical Specification 3.2 requirements and ensure these safety 
limits are maintained in the event of the worst-case transient. 
Since the MCPR safety limit will be maintained at all times, 
operation under the proposed changes will ensure at least 99.9% of 
the fuel rods in the core do not experience transition boiling. 
Therefore, The Minimum Critical Power Ratio (MCPR) safety limit 
change does not affect the probability or consequences of an 
accident.
    The implementation of GE's GESTAR-II approved methodology has no 
effect on the probability or consequences of any accidents 
previously evaluated. One exception to GESTAR is that the mis-
oriented and mis-located bundle events will continue to be analyzed 
as accidents subject to the acceptance criteria in the current 
licensing basis. The design of the GE11 fuel bundles is such that 
the bundles are not likely to be mis-oriented or mis-located and the 
normal administrative controls will be in effect for assuring proper 
orientation and location. Therefore, the probability of a fuel 
loading error is not increased. This analysis ensures that 
postulated dose releases will not exceed a small fraction (10 
percent) of 10 CFR 100 limits.
    Therefore, the consequences of accidents previously evaluated 
are unchanged.
    II. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The GE11 fuel to be used in [Operating] Cycle 9 is of a design 
compatible with fuel present in the core and used in the previous 
cycle. Therefore, the GE11 fuel will not create the possibility of a 
new or different kind of accident. The proposed changes do not 
involve any new modes of operation, any changes to setpoints, or any 
plant modifications. They introduce revised MCPR safety limits that 
have been proved to be acceptable for Cycle 9 operation. Compliance 
with the applicable criterion for incipient boiling transition 
continues to be ensured. The proposed MCPR safety limits do not 
result in the creation of any new precursors to an accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different type of accident from any accident previously 
evaluated.
    III. The proposed change does not involve a significant 
reduction in a margin of safety.
    The MCPR safety limits have been evaluated to ensure that during 
normal operation and during AOOs [abnormal operating occurrences], 
at least 99.9% of the fuel rods in the core do not experience 
transition boiling. Therefore, the implementation of the proposed 
changes in the MCPR safety limit ensure there is no reduction in the 
margin of safety.
    As with the current SPC methodology, GGNS will implement only 
the NRC-approved revisions to GE's GESTAR methodology. This GE 
methodology is similar to those SPC reports currently listed in TS 
5.6.5 and it will be applied in a similar, conservative fashion. One 
exception to GESTAR is that the mis-oriented and mis-located bundle 
events will continue to be analyzed as accidents subject to the 
acceptance criteria in the current licensing basis. This analysis 
ensures that postulated dose releases will not exceed a small 
fraction (10 percent) of 10CFR100 [10 CFR Part 100] limits. On this 
basis, the implementation of this GE methodology 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: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

Gulf States Entergy, 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: August 1, 1996
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to incorporate requirements 
for limiting the time that the hydrogen mixing isolation valves on the 
drywell are open. The requirements were contained in the old TSs and 
with the conversion to the Improved Standard Technical Specifications, 
the requirements were inadvertently changed. The proposed action is to 
restore requirements to meet the licensing basis for the River Bend 
Station. The proposed amendment would also change the time from 7 days 
to 31 days to determine the cumulative time the valves are open.
    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 changes in this submittal put the requirements that 
were in the original Technical Specifications for the Hydrogen 
Mixing System back into the current Technical Specifications. The 
changes reenstate into the Technical Specifications limitations that 
were previously agreed to between River Bend and the Nuclear 
Regulatory Commission in the FSAR Safety Evaluation Report for the 
Hydrogen Mixing System.
    The River Bend SER states in Supplement 2, Section 6.2.4, 
``Since the applicant has not demonstrated that these valves are 
capable of closing under accident conditions in the drywell, certain 
restrictions apply. Technical Specification 3.6.6.2 specifies that 
in Operating Modes 1 and 2, the total number of hours used should 
not exceed 5 hours/365 days and in Operating Mode 3 the number of 
hours should be limited to 90 hours/365 days.'' To date, the 
hydrogen mixing isolation valves have not been fully demonstrated to 
be capable of closing under accident conditions in the drywell. The 
old Standard Technical Specifications (Attachment 2) used at River 
Bend reflected this condition. When conversion to ITS was made, 
these requirements were dropped but should not have been. In 
addition, the requirement to operate the hydrogen mixing system 
every 92 days during Modes 1, 2, and 3 was added without 
consideration for the requirements in the River Bend Safety 
Evaluation Report.
    Consequently, for these proposed change, since the requirements 
already exist and are being reenstated into the Technical 
Specifications, this change is administrative in nature. The 
requirements have remained in place through the SER, but were

[[Page 50344]]

inadvertently removed from the Technical Specifications. This change 
places the requirements from the SER back into the Technical 
Specifications.
    In addition, changing the requirement from the old Technical 
Specifications for determining the cumulative time that the hydrogen 
mixing inlet and outlet valves are open from every 7 days to every 
31 days is again administrative in nature, since this only changes 
the frequency with which a given requirement is tracked 
administratively. It does not change the actual requirement in any 
way.
    Consequently, since both of these changes are administrative in 
nature and only incorporate requirements into the Technical 
Specifications that already existed in the RBS FSAR Safety 
Evaluation Report, the changes proposed in this amendment request do 
not change the probability or consequences of an accident previously 
evaluated.
    This proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any of the parameters or conditions that could contribute to 
the initiation of any accidents.
    The changes proposed in this amendment request are 
administrative in nature and merely add requirements back into the 
Technical Specifications that were inadvertently deleted during the 
conversion to ITS. Because of the administrative nature of the 
proposed changes, it is not possible to create a new or different 
kind of accident from any accident previously evaluated.
    The proposed changes in this amendment request reenstate 
requirements into the Technical specifications that are contained 
present in the RBS FSAR Safety Evaluation Report. These requirements 
were inadvertently deleted during the conversion to ITS.
    Because of the administrative nature of these Technical 
Specification changes, there is no change to 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: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005
    NRC Project Director: William D. Beckner

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: August 15, 1996.
    Description of amendment request: The proposed amendments would 
remove a requirement for performance of a surveillance incorporating a 
high toxic gas test signal.
    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.
    Analyses were performed to evaluated postulated releases of 
potentially hazardous chemicals for their impact on Control Room 
habitability. The latest revision of these analyses shows that none 
of the potentially hazardous chemicals utilized onsite or in the 
surrounding 5-mile radius around the South Texas Project pose a 
credible hazard to the Control Room. Consequently, there is no need 
to ensure that the Control Room Makeup and Cleanup Filtration System 
can automatically switch into a recirculation mode of operation by 
isolating the normal supply and exhaust flow in response to a High 
Toxic Gas test signal. Therefore, elimination of the unnecessary 
surveillance has no effect on the probability of an accident or its 
consequences.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The Toxic Gas Monitoring System was provided to protect against 
hazardous toxic gas releases only. Verifying automatic switch into 
the recirculation mode of operation is no longer necessary since the 
Toxic Gas Analyzers have been removed. This change does not affect 
other tests for verification of automatic switching into the 
recirculation mode of operation. 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 of safety.
    Analyses have shown that none of the chemicals onsite and within 
a 5-mile radius of the South Texas Project pose a credible hazard to 
the facility. Automatic switching of the Control Room Makeup and 
Cleanup Filtration System will continue to be verified using test 
signals from other sources.
    Based upon this evaluation, the South Texas Project has 
concluded that these changes do not involve any significant hazards 
considerations.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
    NRC Project Director: William D. Beckner

Illinois Power Company and Soyland Power Cooperative, Inc., Docket 
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, 
Illinois

    Date of amendment request: August 15, 1996
    Description of amendment request: A Federal Register Notice on May 
22, 1996 (61 FR 25707), stated that revisions were being proposed to 
Clinton Power Station Technical Specification (TS) 3.3.6.2, ``Secondary 
Containment Isolation Instrumentation;'' TS 3.3.7.1, ``Control Room 
Ventilation System Instrumentation;'' TS 3.6.1.2, ``Primary Containment 
Air Locks;'' TS 3.6.1.3, ``Primary Containment Isolation Valves;'' TS 
3.6.4.1, ``Secondary Containment;'' TS 3.6.4.2, ``Secondary Containment 
Isolation Dampers;'' TS 3.6.4.3, ``Standby Gas Treatment;'' TS 3.7.3, 
``Control Room Ventilation;'' and TS 3.7.4, ``Control Room AC System.'' 
By letter dated August 15, 1996, the licensee revised their proposal to 
consolidate the above changes under a newly proposed Special Operations 
LCO (i.e., LCO 3.10.10, ``Single Control Rod Withdrawal - Refueling''). 
Therefore, the Description of Amendment Request to the TSs has changed 
as described herein. The Basis for No Significant Hazards Consideration 
has not changed and is repeated below.
    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 eliminate CORE ALTERATIONS as an 
applicable condition requiring operability of the primary and 
secondary containment and control room ventilation system. As stated 
in the BASES for the associated Technical Specifications, 
operability of these systems is primarily required for mitigation of 
the design basis accident - fuel handling accident (DBA-FHA) and 
design basis accident - loss of coolant accident (DBA-LOCA). The 
performance of CORE ALTERATIONS alone is neither a

[[Page 50345]]

precursor to, nor a condition during which these DBAs are postulated 
to occur. The proposed changes only delete CORE ALTERATIONS as an 
applicable condition for the affected Technical Specifications. All 
other applicable MODES or specified conditions, including operations 
with the potential for draining the reactor vessels (OPDRVs) and the 
movement of irradiated fuel assemblies within the primary or 
secondary containment, remain unchanged. Further, the limitations 
placed on the handling of light loads are also unchanged. The 
Technical Specifications (and the separate requirements imposed on 
the handling of light loads) will thus continue to require that 
systems or functions designed to mitigate design-basis/previously 
evaluated accidents are OPERABLE during the relevant operating MODES 
or conditions. On the basis of the above, it is concluded that the 
requested amendment will not increase the probability or 
consequences of any accident previously evaluated.
    2. The proposed changes do not involve any modification to the 
plant design or to the operation of plant systems (except to 
determine when certain analyzed accident-mitigating systems or 
features are required to be OPERABLE). The failure modes considered 
for the proposed changes are the same as those previously 
considered, therefore, it can be concluded that no new failure modes 
will be created. On this basis, the proposed amendment will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The changes being made to eliminate CORE ALTERATIONS as an 
applicable condition for which certain LCOs must be met, do not 
eliminate the requirements for operability of those systems or 
features assumed to mitigate design-basis or analyzed accidents 
during the applicable MODES when such systems or features are 
assumed to be available for performing their mitigating function. 
The safety margins assumed or established by the accident analyses 
for those design-basis events (as described in the accident analyses 
of the Clinton Power Station Updated Final Safety Analysis Report) 
therefore remain unchanged. Further, the proposed changes do not 
impact the controls imposed on the handling of light loads 
(including unirradiated fuel assemblies) for ensuring that such 
activities cannot result in an event that yields consequences more 
severe than those calculated for the DBA-FHA. With respect to 
reactivity concerns during refueling operations (MODE 5), all 
systems or features required to be OPERABLE for precluding 
inadvertent criticality and monitoring reactivity changes will 
continue to be required OPERABLE as per the current Technical 
Specification requirements. The deletion of CORE ALTERATIONS as an 
applicable condition only applies to the noted systems which do not 
contribute to precluding reactivity events. Based on the above, the 
proposed changes do 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: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727
    Attorney for licensee: Leah Manning Stetzner, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th Street, 
Decatur, Illinois 62525
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: August 12, 1996
    Description of amendment request: The proposed amendment would add 
an additional circumstance to Exception 2 of Technical Specification 
(TS) 3.6, Emergency Core Cooling and Containment Spray Systems, during 
which operation of a service water/component cooling pump subsystem is 
permitted at reduced flow to flush the service water header or inlet 
strainer. The Bases for this TS would be augmented to support the 
additional circumstance of reduced service water flow.
    The proposed amendment would also modify the valve surveillance 
requirements of TS 4.6.A.1.b, Periodic Testing of ECCS Valves, to 
provide an exception to surveillance requirements for those locked 
valves that are inaccessible during power operations or located in a 
locked high radiation area. The Bases for this TS would be augmented to 
support the change in surveillance requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff's analysis is presented below.
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Invocation of the proposed addition to Exception 2 to TS 3.6 would 
not alter any associated Remedial Action completion time, nor those of 
TS 3.0.A, Nonconformance with a Limiting Condition for Operation. The 
evolutions for which this amendment is intended (flushing a heat 
exchanger inlet strainer or cleaning a service water header that has 
become fouled)are administratively controlled by procedures that 
require review and approval by the Plant Operation Review Committee.
    The proposed change to TS 4.6.A.1.b would revise the surveillance 
requirements for a very limited number of locked manual valves in the 
emergency core cooling system (ECCS). The purpose of the surveillance 
requirements is unchanged and is intended to verify that locked valves 
remain in their correct position. The position of the valves is not 
changed and the revised surveillance requirements will continue to 
demonstrate ECCS valve operability.
    Thus, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed addition to Exception 2 to TS 3.6 recognizes that 
service water cleaning and flushing are operations that are required to 
maintain heat transfer capability and equipment reliability. The 
proposed amendment does not affect the design of the plant and do not 
permit operation of the plant outside the currently allowed modes of 
operation.
    The proposed change to TS 4.6.A.1.b maintains verification of ECCS 
valve operability, while requiring no changes in system configuration 
to perform surveillance testing. System functional performance is not 
adversely affected.
    Thus, the proposed amendment 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 change to TS 3.6 does not significantly alter the 
availability or condition of applicable equipment and therefore does 
not alter the accident analyses or the conclusions associated with 
that equipment. The proposed change permits service water flow to be 
reduced below that required for operation of the ECCS in the 
recirculation mode, for a short time. The time during which flow is 
reduced and both the mussel control and flushing evolutions are 
administratively controlled by procedures reviewed and approved by 
the Plant Operation Review Committee.
    The proposed change to TS 4.6.A.1.b maintains verification of valve 
operability. Valve position surveillances will continue to be conducted 
in accordance with plant Technical Specifications to ensure valve 
operational readiness.
    Thus, there is no significant reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to

[[Page 50346]]

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 Deputy Director: 
John A. Zwolinski

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 requests: June 7, 1996
    Description of amendment requests: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant, Unit Nos. 1 and 2 by revising Technical Specifications 
(TS) 3/4.9.14.1, ``Spent Fuel Assembly Storage - Spent Fuel Pool Region 
2,'' and 3/4.9.14.3, ``Spent Fuel Assembly Storage - Spent Fuel Pool 
Region 1,'' to allow storage of fuel assemblies in a checkerboard 
pattern in region 2 of the spent fuel pool.
    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.
    Analysis indicates that allowing fuel storage in a checkerboard
    pattern with empty storage cells in region 2 of the spent fuel
    pool will not result in an inadvertent criticality event. The 
k<INF>eff will continue to remain below 0.95 as required to meet the 
acceptance criteria in the NRC Standard Review Plan, Section 9.1.1.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The change to allow fuel storage in a checkerboard pattern with 
no minimum burnup requirements in region 2 of the spent fuel pool 
would designate locations where a fuel assembly could be incorrectly 
placed. However, the incorrect placement of a fuel assembly has been 
analyzed and would not cause an inadvertent criticality or any other 
accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The NRC Standard Review Plan, Section 9.1.1, acceptance 
criterion of a k<INF>eff of 0.95 provides the margin to criticality. 
An analysis was performed that concluded that the proposed change to 
allow fuel storage in spent fuel pool region 2 in a checkerboard 
pattern meets the acceptance criterion.
    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 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 and 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Project Director: William H. Bateman

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 amendment request: June 6, 1996 (TS 372)
    Description of amendment request: The proposed amendment revises 
Section 6 of the Browns Ferry Nuclear Plant Units 1, 2, and 3 technical 
specifications. Administrative controls associated with quality 
assurance are relocated to the licensee's Nuclear Quality Assurance 
Plan, consistent with Administrative Letter 95-06, and provides 
revisions that make Section 6 more consistent with the improved 
Standard Technical Specifications. Additional administrative changes 
are included to ensure consistent terminology within the 
specifications, and to update obsolete items such as titles and 
addresses. The proposed amendment also includes minor editorial 
changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed TS change to revise items 1 through 28 above 
(Section I, Description of the Proposed Change) was evaluated and 
the proposed TS changes were determined to be administrative in 
nature. The changes [items 2 through 9, 11, 17 through 21, 23, 26, 
and 27] involve administrative title changes of TVA management 
positions, the updating of an NRC mailing address and an NRC 
regional office title. In addition, certain sections [items 1, 10, 
12, 13, 24, and 25] are being relocated into other licensee 
documents for which those provisions are adequately controlled by 
regulatory requirements. [Items 14, 15, 16, 22, and 28 are editorial 
changes.] These changes do not affect any of the design basis 
accidents. They do not involve an increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS change to revise items 1 through 28 above 
(Section I, Description of the Proposed Change) was evaluated and 
the proposed TS changes were determined to be administrative in 
nature. The changes involve administrative title changes of TVA 
management positions, the updating of an NRC mailing address and an 
NRC regional office title. In addition, certain sections are being 
relocated into other licensee documents for which those provisions 
are adequately controlled by regulatory requirements. These changes 
do not affect any of the design basis accidents. No modifications to 
any plant equipment are involved. There are no effects on system 
interactions made by these changes. They do not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed TS change to revise items 1 through 28 above 
(Section I, Description of the Proposed Change) was evaluated and 
the proposed TS changes were determined to be administrative in 
nature. The changes involve administrative title changes of TVA 
management positions, the updating of an NRC mailing address and an 
NRC regional office title. In addition, certain sections are being 
relocated into other licensee documents for which those provisions 
are adequately controlled by regulatory requirements. The margin of 
safety as reported in the basis for the TSs is not reduced. The 
proposed change is administrative and does not impact any technical 
information contained in the bases of the TS.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611

[[Page 50347]]

    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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 amendment request: August 30, 1996 (TS 380)
    Description of amendment request: The proposed amendment deletes 
License Condition 2.C.(3) regarding thermal water quality standards 
from the licenses for the Browns Ferry Nuclear Plant Units 1, 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 amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed License Condition change is an adminstrative change 
and has no relationship to plant safety analyses. Therefore, this 
change does not increase the frequency of the precursors to design 
basis events or operational transients analyzed in the BFN [Browns 
Ferry Nuclear Plant] Final Safety Analysis Report. Likewise, the 
proposed changes will not increase the consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed License Condition change is an administrative 
change and has no relationship to plant safety analyses. Thus, the 
change does not create any type of new accident sequences. Likewise, 
the proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed License Condition change is an administrative 
change and has no relationship to plant safety analyses. Therefore, 
the proposed amendment does not involve a reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

    Date of amendment request: August 16, 1996
    Description of amendment request: This notice relates to your 
submittal to remove the uncertainty term from the specified distance 
and remove the footnote which specifies the time frame it is 
applicable.
    Date of publication of individual notice in Federal Register: 
September 11, 1996 (61 FR 47968)
    Expiration date of individual notice: October 11, 1996
    Local Public Document Room location: location: Waukegan Public 
Library, 128 N. County Street, Waukegan, Illinois 60085.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

    Date of amendment request: September 3, 1996
    Description of amendment request: This notice relates to your 
submittal to modify Technical Specification Section 4.3.1.B.4.A.10.a 
which provides the acceptance criteria for steam generator tube repairs 
by adding a footnote which references the cleanliness and 
nondestructive examination requirements as described in CEN-629-P, 
Revision 00, ``Repair of Westinghouse Series 44 and 51 Steam Generator 
Tubes Using Leak Tight Sleeves.'' Date of publication of individual 
notice in Federal Register: September 11, 1996 (61 FR 47966)
    Expiration date of individual notice: October 11, 1996
    Local Public Document Room location: location: Waukegan Public 
Library, 128 N. County Street, Waukegan, Illinois 60085.

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 amendment request: March 25, 1996, as supplemented by 
letter dated August 23, 1996
    Brief description of amendment request: The proposed amendment 
would revise the safety limit minimum critical power ratios (SLMCPRs) 
to support use of GE-13 fuel at PBAPS, Units 2 and 3. Date of 
publication of individual notice in Federal Register: August 30, 1996 
(61 FR 45997)
    Expiration date of individual notice: September 30, 1996
    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.

Pennsylvania Power and Light Company, Docket No. 50-387 Susquehanna 
Steam Electric Station, Unit 1, Luzerne County, Pennsylvania

    Date of amendment request: May 28, 1996, as supplemented by letter 
dated July 25, 1996
    Brief description of amendment request: The proposed amendment 
would revise the Minimum Critical Power Ratio safety limit values, 
adding two references to reflect the use of the ANF-B Critical Power 
Ratio Correlation and to reflect the use of the ABB Combustion 
Engineering licensing methodology, with a modification to the 
associated Bases.
    Date of publication of individual notice in Federal Register: 
September 9, 1996 (61 FR 47529)
    Expiration date of individual notice: October 9, 1996
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

[[Page 50348]]

Notice Of Issuance Of Amendments To Facility Operating Licenses

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

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

    Date of application for amendment: January 30, 1996, as 
supplemented May 20, 1996
    Brief description of amendment: This amendment revises the 
Technical Specifications (TS) to: (1) add TS 4.6.1.5 to provide 
criteria for 24-hour full-load testing of the emergency diesel 
generators (EDGs) to be performed during each refueling outage; (2) 
revise TS 4.6.1.2 to allow testing of the EDG protective bypasses 
listed in TS 3.7.1.d to be done independent of the safety injection or 
loss of offsite power testing; and (3) revise TS 4.6.1.3 to include the 
EDG protective bypass inspection.
    Date of issuance: September 11, 1996
    Effective date: September 11, 1996
    Amendment No. 174
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7546) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 11, 1996. The May 20, 
1996, letter provided clarifying information that did not change the 
initial proposed no significant hazards consideration determination. No 
significant hazards consideration comments received: No
    Local Public Document Room location: location: Hartsville Memorial 
Library, 147 West College Avenue, Hartsville, South Carolina 29550

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

    Date of application for amendment: December 10, 1995, as 
supplemented August 1, 1996, and September 4, 1996.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) Section 3.5.1 and Tables 3.5-2, 3, and 4 concerning 
the reactor trip system, engineering safety feature actuation system, 
and isolation function.
    Date of issuance: September 12, 1996Effective date: September 12, 
1996
    Amendment No. 175
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 14, 1996 (61 
FR 5812). The August 1, 1996, and September 4, 1996, submittals 
provided administrative changes to the TS pages that did not change the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated September 12, 1996.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550

Duke Power Company, et al., Docket No. 50-413, Catawba Nuclear 
Station, Unit 1, York County, South Carolina

    Date of amendment request: September 30, 1994, as supplemented 
September 18, 1995, January 19, March 15, May 16, and August 27, 1996
    Description of amendment: The amendment revises the Technical 
Specifications to reflect the new setpoints, operational parameters, 
and approved analysis methodologies associated with replacement of the 
Unit 1 steam generators. The amendment also deletes references to steam 
generator tube repair methods, which will no longer be applicable after 
the replacement, and clarifies initial surveillances.
    Date of issuance: August 29, 1996
    Effective date: As of the date of issuance, to be implemented 
within 30 days
    Amendment No.: 151
    Facility Operating License No. NPF-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 10, 1996 (61 FR 
15986) The May 16 and August 27, 1996, letters provided clarifying 
information that did not change the scope of the September 30, 1994, 
application and the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated August 29, 1996. 
No significant hazards consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

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

    Date of application for amendment: July 17, 1996, as supplemented 
August 28, 1996 (TSCR 242, Rev. 2). This application supersedes 
applications dated February 23 (TSCR 242) and June 19, 1996 (TSCR 242, 
Rev. 1).
    Brief description of amendment: The amendment changes the Technical 
Specifications (TS) to allow the implementation of 10 CFR Part 50, 
Appendix J, Option B.
    Date of Issuance: September 3, 1996
    Effective date: September 3, 1996, to be implemented within 30 days 
of issuance
    Amendment No.: 186
    Facility Operating License No. DPR-16. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
40019) Supersedes notice dated March 27, 1996 (61 FR 13526). The August 
28, 1996, supplement provided updated and corrected TS and bases pages. 
These

[[Page 50349]]

revisions were within the scope of the original application and did not 
change the staff's initial proposed no significant hazards 
consideration determination. Therefore renoticing was not warranted. 
The Commission's related evaluation of this amendment is contained in a 
Safety Evaluation dated September 3, 1996. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: location: Ocean County 
Library, Reference Department, 101 Washington Street, Toms River, NJ 
08753
    Illinois Power Company and Soyland Power Cooperative, Inc., Docket 
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois
    Date of application for amendment: February 22, 1996, as 
supplemented by letter dated July 3, 1996
    Brief description of amendment: The amendment revises the Clinton 
Power Station Technical Specifications for the drywell to permit bypass 
testing on a 10-year frequency with increased testing if performance 
degrades, changes the drywell air lock testing and surveillance 
requirements, deletes action notes for the drywell air lock and drywell 
isolation valves when the bypass leakage limit is not met, and deletes 
the specific leakage limits for the drywell air lock seal.
    Date of issuance: September 4, 1996
    Effective date: September 4, 1996
    Amendment No.: 106
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 24, 1996 (61 FR 
18170) The July 3, 1996, submittal consisted of supporting technical 
information which did not change the staff's initial proposed no 
significant hazards consideration determination or expand the scope of 
the original notice. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated September 4, 1996. 
No significant hazards consideration comments received: No
    Local Public Document Room location: location: The Vespasian Warner 
Public Library, 120 West Johnson Street, Clinton, Illinois 61727

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

    Date of amendment request: June 20, 1996
    Description of amendment request: The proposed amendment modifies 
the Seabrook Station Appendix A Technical Specifications (TSs) for the 
Electrical Power Systems, Onsite Power Distribution. Specifically, the 
proposed amendment changes TS 3.8.3.1, Action a., to increase from 8 
hours to 7 days the allowable time that 480-volt Emergency Bus 
<greek-i>E64 may be less than fully energized.
    Date of issuance: August 30, 1996
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment No.: 48
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 26, 1996 (61 FR 
33142) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 30, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location:  location: Exeter Public 
Library, Founders Park, Exeter, NH 03833

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

    Date of application for amendment: April 25, 1996
    Brief description of amendment: The amendment modifies the 
calibration requirement for the source range monitors and intermediate 
range monitors by noting that the sensors are excluded.
    Date of issuance: August 19, 1996
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 96
    Facility Operating License No. DPR-21. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 19, 1996 (61 FR 
31183) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 19, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: 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. 50-336, 
Millstone Nuclear Power Station, Unit No. 2, New London County, 
Connecticut

    Date of application for amendment: March 28, 1996
    Brief description of amendment: The amendment changes Technical 
Specification 3.7.7, ``Sealed Source Contamination,'' and its Bases 
that modify the criteria for testing sealed sources for contamination 
and leakage. The approved changes are consistent with the testing 
criteria currently used at the Millstone Nuclear Power Station, Unit 
No. 3, the Haddam Neck Plant, and the Seabrook Station.
    Date of issuance: September 4, 1996
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 202
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20853) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 4, 1996 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

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay 
Power Plant, Unit 3, Humboldt County, California

    Date of application for amendment: March 13, 1996
    Brief description of amendment: This amendment revised the 
Technical Specification by incorporating position changes to reflect a 
proposed plant staff reorganization.
    Date of issuance: September 6, 1996
    Effective date: This license amendment is effective as of the date 
of its issuance and must be fully implemented no later than 30 days 
from the date of issuance.
    Amendment No.: 31Facility License No. DPR-7: This amendment revised 
the Technical Specifications
    Date of initial notice in Federal Register: April 24, 1996 (61 FR 
18174) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 6, 1996. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Humboldt County Library, 1313 
3rd Street, Eureka, California 95501

[[Page 50350]]

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of application for amendments: February 23, 1996, as 
supplemented by letter dated June 28, 1996
    Brief description of amendments: These amendments change the 
Technical Specification Requirement 4.6.2.1d concerning drywell-to-
suppression chamber bypass testing interval to correspond with the 
interval for Primary Containment Integrated Leak Rate Testing under 10 
CFR Part 50, Appendix J, Option B.
    Date of issuance: September 6, 1996
    Effective date: September 6, 1996
    Amendment Nos.: 160 and 131
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 10, 1996 (61 FR 
15992) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 6, 1996. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

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

    Date of application for amendment: June 21, 1996, as supplemented 
August 19, 1996, and August 21, 1996.
    Brief description of amendment: The amendment extends the 
surveillance interval on certain instruments from 18 to 24 months.
    Date of issuance: September 5, 1996
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 168
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
49027) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 5, 1996. No 
significant hazards consideration comments received: No
    Local Public Document Room location: location: White Plains Public 
Library, 100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: March 6, 1996, as supplemented 
by letter dated May 30, 1996.
    Brief description of amendment: The amendment changes Technical 
Specification (TS) 3.8.1, ``A.C. Sources - Operating,'' to decrease the 
minimum fuel oil storage capacity of the Emergency Diesel Generator 
Fuel Oil Storage Tanks, from 48,800 to 44,800 gallons. In addition, 
footnote ** is deleted from TS 3.8.1.1.b.2. The TS change also adds an 
Action Statement to address remedial action when a fuel oil transfer 
pump becomes inoperable.
    Date of issuance: September 10, 1996
    Effective date: As of date of issuance, to be implemented within 90 
days.
    Amendment No.: 96
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34897) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 10, 1996. No 
significant hazards consideration comments received: No
    Local Public Document Room location: location: Pennsville Public 
Library, 190 S. Broadway, Pennsville, New Jersey 08070

Southern California Edison Company, et al, Docket No. 50-206, San 
Onofre Nuclear Generating Station, Unit No. 1, San Diego County, 
California

    Date of application for amendment: December 22, 1995
    Brief description of amendment: The change revises the San Onofre 
Unit 1 License Condition to delete a reference to License Condition 
2.C(4) from License Condition 2.D. This change eliminates a reporting 
requirement for violations of the physical protection plans that is 
redundant to reporting requirements in 10 CFR 73.71 and 10 CFR Part 73 
Appendix G.
    Date of issuance: August 30, 1996
    Effective date: August 30, 1996 and shall be implemented no later 
than 30 days from August 30, 1996.
    Amendment No.: 157
    Facility Operating License No. DPR-13: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
40028) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 30, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Science Library, University of 
California, Irvine, California 92713

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: June 12, 1996
    Brief description of amendments: The amendments revise the reactor 
core safety limits, Overtemperature delta T (OTDT) and Overpressure 
delta T (OPDT) reactor trip setpoints and allowable values, and the 
power distribution limits associated with implementation of Relaxed 
Axial Offset Control (RAOC) and F<INF>Q surveillance. The amendments 
also include changes to the Bases associated with these specifications 
and surveillances.
    Date of issuance: September 3, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 121 and 113
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
40029) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 3, 1996. No 
significant hazards consideration comments received: No
    Local Public Document Room location: location: Houston-Love 
Memorial Library, 212 W. Burdeshaw Street, Post Office Box 1369, 
Dothan, Alabama 36302

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: June 20, 1996
    Brief description of amendments: The amendments revise the 
Technical Specifications to reflect the implementation of 10 CFR Part 
50, Appendix J, Option B.
    Date of issuance: September 3, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 122 and 114
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
40030) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 3, 1996. No 
significant hazards consideration comments received: No

[[Page 50351]]

    Local Public Document Room location: location: Houston-Love 
Memorial Library, 212 W. Burdeshaw Street, Post Office Box 1369, 
Dothan, Alabama 36302

Tennessee Valley Authority, Docket Nos. 50-390 Watts Bar Nuclear 
Plant, Unit 1, Rhea County, Tennessee

    Date of application for amendment: July 31, 1996
    Brief description of amendment: The amendment revises Technical 
Specification 3.6.12 to allow a one-time extension of the 3-month 
surveillance requirement for the ice condenser lower inlet doors.
    Date of issuance: September 9, 1996
    Effective date: As of the date of issuance, to be implemented 
within 30 days
    Amendment No.: 3
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 8, 1996 (61 FR 
41431) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 9, 1996. No 
significant hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: February 19, 1996, as 
supplemented on July 3 and August 26, 1996
    Brief description of amendment: The amendment revises Kewaunee 
Nuclear Power Plant Technical Specification Section 4.2 and its 
associated basis by allowing the application of a voltage-based repair 
limit for the steam generator tube support plate intersections 
experiencing outside diameter stress corrosion cracking. The repair 
criteria are based on guidance provided in Generic Letter 95-05, 
``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes 
affected by Outside Diameter Stress Corrosion Cracking,'' dated August 
3, 1995, and on associated industry guidance.
    Date of issuance: September 11, 1996
    Effective date: September 11, 1996, and is to be implemented within 
30 days of the date of issuance.
    Amendment No.: 126
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 10, 1996 (61 FR 
15999) The July 3 and August 26, 1996, submittals 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 
September 11, 1996. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001

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

    Date of amendment request: October 24, 1995, and superseded by 
letter dated May 16, 1996.
    Brief description of amendment: The amendment adopts ASTM D3803-
1989 as the laboratory testing standard for charcoal samples from the 
charcoal absorbers in the control room filtration system, control 
building pressurization system, and the auxiliary/fuel building 
emergency exhaust system. The output of the heaters in the control 
building pressurization system is reduced from a nominal 15kW to a 
nominal 5kW and the acceptance criterion for the testing of the 
charcoal absorbers is changed.
    Date of issuance: September 4, 1996
    Effective date: September 4, 1996, to be implemented within 120 
days of the date of issuance.
    Amendment No.: 102
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28622) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 4, 1996. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: locations: Emporia State 
University, William Allen White Library, 1200 Commercial Street, 
Emporia, Kansas 66801 and Washburn University School of Law Library, 
Topeka, Kansas 66621 Dated at Rockville, Maryland, this 18th day of 
September 1996.
    For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II Office of Nuclear Reactor 
Regulation
[Doc. 96-24413 Filed 9-24-96; 8:45 am]
BILLING CODE 7590-01-F