[Federal Register Volume 61, Number 138 (Wednesday, July 17, 1996)]
[Notices]
[Pages 37295-37307]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-20717]


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NUCLEAR REGULATORY COMMISSION

Biweekly Notice

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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the

[[Page 37296]]

Commission the authority to issue and make immediately effective any 
amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from June 22, 1996, through July 5, 1996. The 
last biweekly notice was published on July 3, 1996 (61 FR 34884).

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 August 16, 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 
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,

[[Page 37297]]

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.

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

    Date of amendments request: April 4, 1996
    Description of amendments request: The proposed amendments would 
revise the Technical Specifications (TS) to add an allowance to 
complete a TS-required surveillance within 24 hours of discovery of a 
missed surveillance in accordance with the guidance of Generic Letter 
(GL) 87-09, ``Sections 3.0 and 4.0 of the Standard Technical 
Specifications (STS) on the Applicability of Limiting Conditions for 
Operation and Surveillance Requirements'' and NUREG-1433, ``Standard 
Technical Specifications, General Electric Plants, BWR/4,'' Revision 1, 
April 1995. Typographical errors are being corrected and wording 
adjustments are being incorporated for consistency between plant TS 
terminology and the associated Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    . The proposed amendments do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The operational flexibility resulting from the proposed 
revision to Technical Specification 3.0.4 is consistent with that 
allowed by the existing individual LCO [limiting condition for 
operation] and their associated ACTION requirements, which provide 
an acceptable level of safety for continued operation. A delay of up 
to 24 hours or the time of the surveillance interval, whichever is 
less, provided by Technical Specification 4.0.3 to complete a missed 
surveillance reduces the probability of a transient occurring when 
the affected system or component is either out of service to allow 
performance of the surveillance test, or there is a lower level of 
confidence in the operability because the normal surveillance was 
exceeded. The revision to Technical Specification 4.0.4 makes it 
clear that Technical Specification 4.0.4 does not prevent passage 
through or to OPERATIONAL CONDITIONS as required to comply with 
ACTION requirements. The revision to the wording in Unit 2 Technical 
Specification Table 3.12.1-1, Notation (h), revisions to the Bases 
of the Technical Specifications, and the elimination of specific 
exemptions to Technical Specifications 3.0.4 are administrative in 
nature.
    Based on the above, the proposed license amendments do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed amendments do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The proposed license amendments do not introduce any new 
equipment nor do they require any existing equipment or systems to 
perform a different type of function than they are presently 
designed to perform. The proposed changes result in improved 
Technical Specifications by removing unnecessary restrictions on 
changes in OPERATIONAL CONDITIONS and facility operation, removing 
unnecessary shutdowns caused by inadvertently exceeding surveillance 
intervals, and removing conflicts between various Technical 
Specifications. The revision to the wording in Unit 2 Technical 
Specification Table 3.12.1-1, Notation (h), revisions to the Bases 
of the Technical Specifications, and the elimination of specific 
exemptions to Technical Specification 3.0.4 are administrative in 
nature.
    Based on the above, the proposed license amendments do not 
create a new or different kind of accident from any previously 
evaluated.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety. The operational flexibility that 
results from the proposed revision to Technical Specification 3.0.4 
is consistent with that allowed by the existing individual LCO and 
associated ACTION requirements, which provide an acceptable level of 
safety for continued operation. Therefore, there is no change in the 
margin of safety associated with this change. A delay of up to 24 
hours or the length of the surveillance interval, whichever is less, 
provided by Technical Specification 4.0.3 to complete a missed 
surveillance reduces the probability of a transient occurring when 
the affected system or component is either out of service to allow 
performance of the surveillance test, or there is a lower level of 
confidence in the operability because the normal surveillance was 
exceeded. In addition, the proposed change acknowledges that the 
most common outcome of the performance of a surveillance is the 
successful demonstration that acceptance criteria are met. The 
proposed change provides the potential benefit of avoiding a 
shutdown transient when required equipment is still capable of 
performing its function, and variables are still within limits. The 
revision to Technical Specification 4.0.4 makes it clear that 
Technical Specification 4.0.4 does not prevent passage through or to 
OPERATIONAL CONDITIONS as required to comply with ACTION 
requirements. This change is considered to be a clarification to 
achieve consistency with existing Technical Specification 
requirements. The revision to the wording in Unit 2 Technical 
Specification Table 3.12.1-1, Notation (h), revisions to the Bases 
of the Technical Specifications, and the elimination of specific 
exemptions to Technical Specification 3.0.4 are administrative in 
nature.
    The proposed changes would result in improved Technical 
Specifications and eliminate unnecessary plant challenges. Based on 
the above, the proposed license amendments 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: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297

[[Page 37298]]

    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: Eugene V. Imbro

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van 
Buren County, Michigan

    Date of amendment request: December 6, 1995
    Description of amendment request: The proposed amendment would 
relocate the crane operation and movement of heavy loads requirements 
and their bases from the Technical Specifications (TS) to other plant 
documents.
    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 change moves the requirements from TS to other plant 
documents controlled under 10 CFR 50.59 without affecting their 
technical content. Since this change does not alter the technical 
content of any requirements, the operation of the facility in 
accordance with the proposed change cannot involve a significant 
increase in the probability or consequences of an accident 
previously evaluated, create the possibility of a new or different 
kind of accident from any previous evaluated, or 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: Van Wylen Library, Hope 
College, Holland, Michigan 49423.
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
    NRC Project Director: Mark Reinhart

Duke Power Company, Docket Nos. 50-269, 270 and 50-287, Oconee 
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina

    Date of amendment request: June 6, 1996
    Description of amendment request: The proposed change would remove 
the Engineered Safeguard (ES) signals that presently open the outlet 
valves on the Low Pressure Service Water (LPSW) System coolers, LPSW-4 
and LPSW-5, on high reactor coolant system pressure or high reactor 
building pressure. The valves will continue to be operable from the 
control room when needed. The proposed change to Technical 
Specification (TS) 4.5.1.1.2.a.(2) would require that the refueling 
outage test signal be applied to the LPSW pumps, but no longer to LPSW-
4 and LPSW-5, and that the operability of the valves be verified by 
cycling them from the control room. A note would be added to reflect 
that the refueling outage test of LPSW-4 and LPSW-5 response to the ES 
signal will continue to be verified until the signal is removed from 
the ES system for each unit during the specified refueling outages. In 
addition, TS 4.5.1.1.2.b would be clarified to differentiate between 
test acceptance criteria for automatic actuation of the appropriate 
LPSW pumps and valves in response to the ES signal, and completion of 
travel of LPSW-4 and LPSW-5 in response to manual operation of the 
valves. A proposed change to the Bases would also reflect these 
changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Pursuant to 10CFR50.91, Duke Power Company (Duke) has made the 
determination that this amendment involves a No Significant Hazards 
Consideration by applying the standards established by NRC 
regulations in 10CFR50.92. The following discusses the basis for our 
analysis:
    Will operation of the facility in accordance with the proposed 
amendment:
    A. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. Eliminating the automatic signal that opens Low Pressure 
Service Water (LPSW) System valves, LPSW-4 and LPSW-5, upon an 
Engineered Safeguards (ES) actuation does not increase the 
probability of any accident previously evaluated. The proposed 
change would involve a delay in providing cooling water to the Low 
Pressure Injection (LPI) System coolers after a design basis 
accident. Cooling water flow to the LPI coolers is isolated during 
normal power operation. During normal cold shutdown conditions, 
cooling water flow to the LPI coolers is normally open without 
relying on the ES actuation signal. This cooling water flow is 
needed to mitigate certain accidents, but a delay in providing this 
cooling water flow after a design basis accident does not 
significantly increase the probability of any accident previously 
evaluated.
    Eliminating the ES actuation signal for LPSW-4 and LPSW-5 will 
not increase the consequences of an accident previously evaluated. 
After a loss of coolant accident (LOCA), operators will operate the 
appropriate valves from the control room in sufficient time to 
provide adequate cooling water flow to maintain containment 
temperature and pressure within acceptable limits. Duke has also 
evaluated the delay of LPSW cooling flow's impact on core cooling 
and concluded that there are no adverse impacts on the capability to 
maintain core cooling. Since the containment temperature and 
pressure limits after a LOCA will not be exceeded, this change will 
not increase any potential off-site dose consequences after a LOCA. 
Due to the time available for operator action (approximately one 
hour), there is no significant increase in operator burden during 
this accident scenario.
    B. Create the possibility of a new or different kind of accident 
from the accidents previously evaluated?
    No. As stated above, due to the time available for operator 
action (approximately 1 hour), there is no significant increase in 
operator burden during this accident scenario. Eliminating the ES 
signal that automatically opens valves LPSW-4 and LPSW-5 results in 
significantly lower flow demand on the LPSW pumps. If all LPSW pumps 
are successfully started, this could result in a stronger pump 
causing deadhead conditions on a weaker pump since the pumps feed 
into the same piping system. To prevent any potential adverse 
effects on the LPSW pumps due to inadequate flow during the initial 
stages of a LOCA, minimum flow piping will be installed for the LPSW 
pumps to provide adequate flowpaths for pump minimum flow. Testing 
will be performed to validate that the LPSW pumps can operate at the 
chosen design value for pump minimum flow. In addition, Duke 
conducted an evaluation, based on manufacturer input, of the thermal 
effects on the LPI coolers due to delaying LPSW cooling flow. This 
evaluation concluded that the 30 minute delay of LPSW cooling flow 
has no adverse thermal effects on the LPI coolers. Therefore, 
because there is no significant increase in operator burden and 
because there will be no adverse effects on the LPSW pumps, LPI 
coolers, and associated piping caused by the delayed LPSW cooling 
flow, the proposed change will not create the possibility of a new 
or different kind of accident from the accidents previously 
evaluated.
    C. Involve a significant reduction in a margin of safety?
    No. There are no safety limits or limiting safety system 
settings associated with the LPSW System in the Oconee Nuclear 
Station Technical Specifications. The proposed change will not 
affect any existing safety limits or limiting safety system 
settings. The proposed change will not affect any existing Limiting 
Conditions for Operation in the Technical Specifications. The 
proposed change involves an alternative method of initiating cooling 
water flow to the LPI coolers after a LOCA. This alternative method 
will achieve the required results since there will be no significant 
change in the containment temperature and pressure after a LOCA.
    Duke has concluded based on the above that there are no 
significant hazards considerations involved in this amendment 
request.
    The NRC has reviewed the licensee's analysis and, based on this 
review, it

[[Page 37299]]

appears that the three standards of 10 CFR 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: June 20, 1996
    Description of amendment request: The amendment would redefine the 
secondary containment boundary to allow the enclosure building to be 
inoperable during the upcoming refueling outage 8 (RFO 8) scheduled to 
begin in October 1996. The amendment would add a condition to the 
license that the enclosure building may be inoperable during core 
alterations and movement of non-recently irradiated fuel (i.e., fuel 
that has not occupied part of a critical reactor core for 12 days) 
during RFO 8 and the standby gas treatment (SGT) system may be unable 
to automatically start or achieve and maintain the required vacuum, 
provided the following conditions exist:
    a. All dampers communicating between the auxiliary building and the 
enclosure building are closed.
    b. The access door between the auxiliary building and the enclosure 
building is closed, except when the access opening is being used for 
entry and exit.
    c. The SGT system is blocked from automatic initiation.
    d. SGT system is available for manual initiation or the actions for 
Limiting Condition for Operation 3.6.4.3 in the Technical 
Specifications for GGNS are complied with.
    The non-recently irradiated fuel is spent fuel that has decayed at 
least 12 days after the reactor was shut down for refueling.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not significantly increase the 
probability or consequences of an accident previously evaluated.
    The equipment affected by the proposed change is not considered 
an initiator to any previously analyzed accident, therefore, 
inoperability of the equipment does not increase the probability of 
any previously evaluated accident.
    As described in Updated Final Safety Analysis Report [for GGNS,] 
Chapter 15, the accidents postulated to occur during core 
alterations in addition to fuel handling accidents are [the 
following]: inadvertent criticality due to a control rod removal 
error or continuous control rod withdrawal error during refueling 
and the inadvertent loading of a fuel assembly in an improper 
location. These events are not postulated to result in fuel cladding 
integrity damage. The only accident postulated to occur during core 
alterations that results in a significant radioactive release is the 
fuel handling accident. The proposed requirements in conjunction 
with existing administrative controls on light loads, bounds the 
conditions of the current design basis fuel handling accident 
analysis which concludes that the radiological consequences are 
within the acceptance criteria of NUREG 0800, Section 15.7.4 and 
General Design Criteria [GDC] 19 [of Appendix A to 10 CFR Part 50]. 
Therefore, the proposed changes do not significantly increase 
consequences of any previously evaluated accident.
    Based on the above, the proposed changes do not significantly 
increase the probability or consequences of any accident previously 
evaluated.
    2. The proposed changes would not create the possibility of a 
new or different kind of accident from any previous analyzed.
    The leaktightness of the enclosure building does not affect the 
function of any plant system other than the ability of the SGT 
System to ensure the secondary containment is at the specified 
pressure. The proposed change in [the] normal SGT System 
alignment[,] by defeating the automatic start feature of the SGT 
System and the inability to ensure secondary containment is at the 
specified pressure[,] does not affect the operation of any [other] 
plant system or component. The SGT System is not relied upon to 
provide normal or accident cooling to plant systems or components. 
The function of the enclosure building and the SGT System is only to 
mitigate the release of radioactivity to the environment in the 
event of an accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes continue to ensure that the radiological 
consequences are at or below the current GGNS licensing limit. 
Safety margins and analytical conservatisms have been evaluated and 
are well understood. Substantial margins are retained to ensure that 
the analysis adequately bounds all postulated event scenarios. The 
current margin of safety is retained.
    Specifically, the margin of safety for the fuel handling 
accident is the difference between the 10CFR100 [dose consequence 
guidelines of 300 rem thyroid and 25 rem whole- body] and the 
licensing limit defined by NUREG-0800, Section 15.7.4. With respect 
to the control room personnel doses, the margin of safety is the 
difference between the 10CFR100 [guidelines] and the licensing limit 
defined by 10CFR50 [10 CFR Part 50], Appendix A, Criterion 19 (GDC 
19). The proposed applicability continues to ensure that the whole-
body and thyroid doses at the exclusion area and low population zone 
boundaries[,] as well as control room doses[,] are at or below the 
corresponding licensing limit. The margin of safety is unchanged; 
therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    In excess to the margin of safety supplied by the licensing 
limits of NUREG-0800 and GDC 19, the proposed change incorporates an 
additional layer of conservative requirements. The proposed change 
leaves in effect a redefined secondary containment boundary which 
will provide a low leakage boundary (consisting of the primary 
containment and the auxiliary building) by automatically isolating 
in the event of the design basis fuel handling accident and requires 
that the SGT System be available for manual initiation when desired. 
These requirements will ensure that doses will be even lower than 
those calculated.
    Therefore, the proposed changes do not result in a significant 
reduction in a margin of safety.
    Based on the above evaluation, operation in accordance with the 
proposed amendment involves no significant hazards considerations.
    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

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

    Date of amendment request: June 17, 1996
    Description of amendment request: The proposed amendments would 
revise Technical Specification Section 5.3.1 to allow use of fuel 
assemblies containing fuel rods clad with ZIRLO<SUP>TM.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 37300]]

licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:
    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because:
    The methodologies used in the accident analyses remain 
unchanged. The proposed change does not change or alter the design 
assumptions for the systems or components used to mitigate the 
consequences of an accident. Use of ZIRLO<SUP>TM fuel cladding does 
not adversely affect fuel performance or impact nuclear design 
methodology. Therefore, accident analysis results are not 
significantly impacted.
    The operating limits will not be changed and the analysis 
methods to demonstrate operation within the limits will remain in 
accordance with NRC-approved methodologies. Other than the changes 
to the fuel assemblies cladding, there are no physical changes to 
the plant associated with this Technical Specification change. A 
safety analysis will continue to be performed for each specific 
reload cycle to demonstrate compliance with all fuel safety design 
bases.
    The 10 CFR 50.46 criteria are applied to the ZIRLO<SUP>TM clad 
fuel rods. The use of these fuel assemblies will not result in a 
change to the reload design and safety analysis limits. Since the 
original design criteria are met, the ZIRLO<SUP>TM clad fuel rods 
will not be an initiator for any new accident. The clad material is 
similar in chemical composition and has similar physical and 
mechanical properties as Zircaloy-4. Thus, the cladding integrity is 
maintained and the structural integrity of the fuel assembly is not 
affected. ZIRLO<SUP>TM cladding improves corrosion performance and 
dimensional stability. Since the dose predictions in the safety 
analyses are not sensitive to the fuel rod cladding material used, 
the radiological consequences of accidents previously evaluated in 
the safety analysis remain valid.
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because:
    The possibility for a new or different kind of accident from any 
accident previously evaluated is not created since the fuel 
assemblies containing ZIRLO<SUP>TM clad fuel rods will satisfy the 
same design bases as that currently used for Zircaloy-4 clad fuel 
assemblies. All design and performance criteria will continue to be 
met and no new single failure mechanisms have been defined. In 
addition, the use of ZIRLO<SUP>TM fuel assemblies does not involve 
any alterations to plant equipment or procedures which would 
introduce any new or unique operational mode or accident precursor. 
Therefore, the possibility for a new or different kind of accident 
from any accident previously evaluated is not created.
    The proposed change does not involve a significant reduction in 
a margin of safety because:
    The margin of safety is not significantly reduced since the 
ZIRLO<SUP>TM clad fuel assemblies will not change the reload design 
and safety analysis limits. Their use will take into consideration 
the normal core operating conditions allowed for in the Technical 
Specifications. Each specific cycle's reload core will continue to 
be specifically evaluated using NRC approved reload design methods 
and approved fuel rod design models. This will include consideration 
of the core physics analysis peaking factor and core average linear 
heat rate effects.
    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:  BurkeCounty Public Library, 
412 Fourth Street, Waynesboro, Georgia 30830
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: June 17, 1996
    Description of amendment request: The proposed amendments would 
clarify the requirement of Technical Specification Surveillance 
Requirement 4.8.1.1.2.j(2) that requires a pressure test of those 
portions of the diesel fuel-oil system that are designed to Section 
III, Subsection ND of the American Society of Mechanical Engineers 
(ASME) Code. The system pressure test would be performed at a pressure 
of 110% of the design pressure, at least once per 10 years and only on 
those sections of piping that are isolable.
    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 clarification of T/S [Technical Specification] 
4.8.1.1.2.j(2) does not involve a significant hazards consideration 
because operation of [the Vogtle Electric Generating Plant] with 
this change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated. The configuration 
of the diesel fuel-oil system as currently installed and operated is 
such that a pressure test of 110% of design pressure would be 
impractical to perform. The system contains tanks designed for 
atmospheric pressure and isolation of them and their vent lines from 
the specified pressure test is not practical. The ASME Code, Section 
XI, provides alternate test methods to use when storage tanks are 
involved in a system pressure test. By clarifying this T/S 
requirement, the requirements set forth in ASME Section XI can be 
utilized as guidance for testing requirements to ensure the 
integrity of the diesel fuel-oil system to perform its intended 
safety function.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated. There are no design changes 
being made that would create a new type of accident or malfunction 
and the method and manner of plant operation remain unchanged. Using 
ASME Section XI as guidance for pressure testing the isolable 
sections of piping provides assurance that the fuel oil supply 
system will perform its intended function.
    3. Involve a significant reduction in a margin of safety. There 
are no changes being made to the safety limits or safety system 
settings that would adversely impact plant safety. Utilizing ASME 
Section XI as guidance for determining those sections of piping that 
should be pressure-tested and atmospheric-tested will ensure proper 
operation of the diesel generator fuel oil supply system.
    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: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia 30830
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: April 10, 1996
    Description of amendment request: The proposed changes bring the 
surveillance requirements to conformance with Amendment No. 196 issued 
September 19, 1995. Additionally, this request changes frequency 
notation for a group of 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

[[Page 37301]]

consideration (SHC), which is presented below:
    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence of the consequences of an accident 
previously evaluated.
    The proposed amendment extends the interval between successive 
refueling interval surveillances to once every 24 months for those 
surveillances evaluated herein, and to make administrative changes 
serving to conform the Technical Specifications to Amendment No. 
196. Except for the administrative changes, the proposed 
surveillance interval changes do not involve any change to the 
actual surveillance requirements, nor does it involve any
    change to the limits and restrictions on plant operations. The 
reliability of systems and components relied upon to prevent or 
mitigate the consequences of accidents previously evaluated is not 
degraded by the proposed change to the surveillance interval. 
Assurance of system and equipment availability is maintained. This 
change does not involve any change to system or equipment 
configuration. Therefore, this change does not increase the 
probability of occurrence or the consequences of an accident 
previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendment extends the interval between successive 
refueling interval surveillances to once every 24 months for those 
surveillances evaluated herein, and to make administrative changes 
serving to conform the Technical Specifications to Amendment No. 
196. Except for the administrative changes the proposed surveillance 
interval changes do not involve any change to the limits and 
restrictions in plant operation. This change does not involve any 
change to system or equipment configuration. Therefore, this change 
is unrelated to the possibility of creating a new or different kind 
of accident from any previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed amendment extends the interval between successive 
refueling interval surveillances to once every 24 months for the 
surveillances evaluated herein, and to make administrative changes 
serving to conform the Technical Specifications to Amendment No. 
196. Except for the administrative changes the proposed surveillance 
interval changes do not involve any change to the actual 
surveillance requirements, nor does it involve any change to the 
limits and restrictions on plant operation. The reliability of 
systems and components is not degraded by the proposed change to the 
surveillance interval. Assurance of system and equipment 
availability is maintained. Therefore, it is concluded that 
operation of the facility in accordance with the proposed amendment 
does not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: May 2, 1996
    Description of amendment request: The proposed change would remove 
Technical Specification Figure 5.1, which is used in maintaining 
K<INF>eff values, and substitute in its place a defined requirement for 
maximum K-infinity for any fuel placed in the Millstone Unit 1 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:
    Pursuant to 10CFR50.92, NNECO [Northeast Nuclear Energy Company] 
has reviewed the proposed change and concludes that the change does 
not involve a significant hazards consideration (SHC) since the 
proposed change satisfies the criteria in 10CFR50.92(c). That is, 
the proposed change does not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    There are no spent fuel pool accident conditions discussed in 
Chapter 15 of the FSAR [Final Safety Analysis Report]. FSAR section 
15.8 discusses a fuel handling accident which drops a fuel assembly 
into the core during refueling. Changing the maximum allowed fuel 
reactivity or allowing gaps in the Boraflex
    panels will have no effect on the probability or consequences of 
a fuel assembly drop onto the core.
    Therefore, based on the above, the proposed change to the 
Technical Specifications does not involve a significant increase in 
the probability or consequences of any previously analyzed accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The reduction in the allowable fuel reactivity in the SFP [spent 
fuel pool] is conservative and does not create the possibility of a 
new or different type of accident. Allowing boraflex gaps does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The margin to safety, for this proposed technical specification 
change, is to maintain the SFP K<INF>eff to be less than or equal to 
0.90. As described in the HOLTEC analysis, gaps in the Boraflex of 
up to 5 inches can exist in every boraflex panel of every rack with 
Boraflex in the SFP, with K<INF>eff still less than 0.90. This is 
true even if all of the gaps are uniformly lined up at the same 
elevation. These calculations conservatively assumed 4% Boraflex 
width shrinkage as well as the axial Boraflex gaps. Older fuel 
designs were also considered to ensure that they had not become 
limiting with the reduced allowable K-infinity limit of 1.24. With 
no boraflex gaps, the maximum K<INF>eff is less than .844. With 5 
inch Boraflex gaps in every panel at the same elevation, the maximum 
K<INF>eff is 0.896, which is less than 0.90. NNECO has implemented a 
1 year decay time requirement to minimize gamma irradiation damage 
to the Boraflex, and will continue to measure via ``blackness 
testing'' the actual gap size to ensure the margin of safety in 
maintained.
    Therefore, this change has no impact on the margin to 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: 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
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: March 29, 1996
    Description of amendment request: The proposed amendment would add 
limits associated with Departure from Nucleate Boiling (DNB) to the 
Indian Point 3 (IP3) Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 37302]]

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:
    Consistent with the criteria of 10 CFR 50.92, the enclosed 
application is judged to involve no significant hazards based on the 
following information:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously analyzed?
    Response:
    The proposed amendment makes no changes to the way in which the 
plant is operated and has no effect on accident initiators 
associated with analyzed transients. The probability of previously 
analyzed accidents is not increased. The proposed amendment 
clarifies the relationship between measurable parameters (RCS 
[reactor coolant system] temperature, pressure, and flow rate) and 
the resulting heat transfer regime in the reactor core, as 
characterized by the Departure from Nucleate Boiling (DNB) ratio. 
This clarification ensures that safety analysis initial conditions 
regarding heat transfer remain valid, so that the consequences of 
previously analyzed accidents are not increased. The changes ensure 
that RCS pressure, temperature, and flow are within analytical 
bounds. This ensures that the plant is operated in a manner that 
will not increase the probabilities of previously analyzed accidents 
nor the consequences of previously analyzed accidents.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response:
    The proposed amendment does not involve any modifications to 
plant systems, structures, or components. The proposed change 
clarifies existing limits on RCS parameters and makes no changes to 
plant setpoints or operating limits. The amendment does not involve 
any physical mechanism which could contribute to a new or different 
kind of accident. The changes ensure that RCS pressure, temperature, 
and flow are within analytical bounds. This ensures that the plant 
is operated in a manner that will not create the possibility of a 
new [or] different kind of accident from any previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response:
    The proposed amendment clarifies existing limits on the 
measurable parameters (RCS temperature, pressure, and flow rate) so 
that the resulting DNB value is consistent with initial condition 
assumptions used in existing safety analyses. Maintaining these 
limits during normal plant operation ensures that the existing 
margins of safety remain valid. The proposed amendment does not 
involve a reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle, 
New York, New York 10019.
    NRC Project Director: Jocelyn A. Mitchell, Acting

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

    Date of amendment request: June 18, 1996
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 5.2.2, ``Design Pressure and 
Temperature,'' by adding design parameters for Main Steam Line Break 
(MSLB). The MSLB analysis results in a higher containment air 
temperature than the current value in TS 5.2.2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The accidents considered for this change are the Loss of Coolant 
Accident (LOCA) and the Main Steam Line Break (MSLB). The proposed 
change ensures the design limiting containment pressure and 
temperature data specified in the TS is consistent with the [Updated 
Final Safety Analysis Report] UFSAR. Since no physical changes to 
the containment are being made there will be no change in the 
probability of either accident occurring.
    Detailed structural analysis presented in Supplement 1 of 
Licensee Event Report (LER) 272/95-016 shows that the Design Basis 
LOCA combination of pressure and temperature result in more severe 
loading for the containment concrete structure and, therefore, 
bounds the temperature and pressure scenario associated with a MSLB 
accident. The pressure retaining capability of the liner is governed 
by the loads generated in the MSLB. Since containment leakage is 
maintained within the limits assumed in the Accident Analysis for 
either scenario there is no change in the consequences of either 
accident.
    Therefore the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The changes proposed affect the post-accident condition of the 
containment, and have no impact on the pre-accident condition. Since 
there is no physical change proposed the containment and all systems 
in the containment will continue to perform as designed. With no 
physical changes being proposed and no change to the pre-accident 
condition of the containment it can be concluded that there will be 
no change in the probability of a new or different accident being 
created.
    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.
    Although calculations indicate that some yielding of the liner 
plate could occur during a MSLB, loading is transferred to the 
containment concrete structure and leakage from the containment is 
maintained within the limits assumed in the Accident Analysis. Since 
containment leakage is maintained within the limits assumed in the 
Accident Analysis the proposed change does not involve a significant 
change the margin of safety provided by the containment for the 
MSLB.
    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: Salem Free Public library, 112 
West Broadway, Salem, NJ 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

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

    Date of amendment request: June 7, 1996 (TSC 95-19)
    Description of amendment request: The proposed change would revise 
Section 6 of the plant Technical Specifications to be more closely 
aligned with the Revised Standard Technical Specifications for 
Westinghouse-designed nuclear plants (NUREG-1431). Additionally, the 
proposed changes would be consistent with the guidance provided in 
Administrative Letter 95-06, ``Relocation of Technical Specification 
Administrative Controls Related to Quality Assurance.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 37303]]

licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:
    TVA [Tennessee Valley Authority] has concluded that operation of 
SQN [Sequoyah Nuclear Plant] Units 1 and 2 in accordance with the 
proposed changes to the TS [Technical Specification] does not 
involve a significant hazards consideration. TVA's conclusion is 
based on its evaluation, in accordance with 10 CFR 50.91(a)(1), of 
the three standards set forth in 10 CFR 50.92(c).
    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed TS change is administrative. TVA has evaluated the 
proposed TS changes and has determined that the proposed changes are 
administrative in nature. 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. They do not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    The proposed TS change is administrative. TVA has evaluated the 
proposed TS changes and has determined that the proposed changes are 
administrative in nature. 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.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed TS change is administrative. TVA has evaluated the 
proposed TS changes and has determined that the proposed changes are 
administrative in nature. 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 has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, 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.

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

    Date of amendment request: June 20, 1996
    Description of amendment request: The proposed amendment would 
increase the allowed time for an inoperable service water cooling tower 
loop electrical supply to be the same as the allowed outage time for an 
operable service water cooling tower loop.
    Date of publication of individual notice in Federal Register: June 
26, 1996 (61 FR 33142)
    Expiration date of individual notice: July 26, 1996
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, New Hampshire

Northeast Utilities Service Company, Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: June 3, 1996
    Description of amendment request: The proposed amendments would 
provide a one-time change to Technical Specification 3.9.1, ``Refueling 
Operations, Boron Concentration.'' This change would remove the 
requirement that the boron concentration in all filled portions of the 
Reactor Coolant System be ``uniform'' and would only be applicable 
during Millstone 2 Cycle 13 mid-cycle core offload.
    Date of publication of individual notice in Federal Register: June 
12, 1996 (61 FR 29771)
    Expiration date of individual notice: July 12, 1996
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut

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

    Date of amendment request: May 23, 1996
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) for the Overtemperature delta 
T time constants in TS Table 2.2-1 and the Steam Line Pressure Negative 
Rate High Steam Line Isolation time constant in TS Table 3.3-4. Date of 
publication of individual notice in Federal Register: June 17, 1996 (61 
FR 30639)
    Expiration date of individual notice: July 17, 1997
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut

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

    Date of amendment request: June 10, 1996
    Brief description of amendment request: The amendment proposes 
changes to Technical Specification 3/4.7.6, ``Control Room Emergency 
Air Conditioning System,'' to reflect a control room design in which 
the common Salem Unit 1 and 2 control room envelope is supplied by 2 
one hundred percent capable Control Room Emergency Air Conditioning 
System trains. Date of publication of individual notice in Federal 
Register: June 24, 1996 (61 FR 32468)
    Expiration date of individual notice: July 24, 1996
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079

[[Page 37304]]

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 amendment request: June 24, 1996
    Description of amendment request: The proposed amendments would 
revise Technical Specification Table 4.3.1 to delete the requirement 
for surveillance of the manual safety injection to the reactor trip 
circuitry until the next unit shutdown, following which, this testing 
will be performed prior to Mode 2 entry. This change is applicable only 
to Unit 1, Cycle 14 and Unit 2, Cycle 11. Date of publication of 
individual notice in Federal Register: July 3, 1996 (61 FR 34880)
    Expiration date of individual notice: August 2, 1996
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, P. O. Box 1369, Dothan, Alabama

Notice Of Issuance Of Amendments To Facility Operating Licenses

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

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

    Date of application for amendments: September 16, 1994, as 
supplemented January 31, 1996.
    Brief description of amendments: The amendments revise the 
technical specifications to eliminate periodic response time testing 
requirements for selected pressure and differential pressure sensors in 
the reactor trip system and engineered safety features actuation 
instrumentation channels.
    Date of issuance: June 26, 1996
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 84, 84, 76 and 76
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 13, 1996 (61 FR 
10393). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 26, 1996. No significant 
hazards consideration comments received: No
    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.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois 
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, 
Units 1 and 2, Rock Island County, Illinois

    Date of application for amendments: November 14, 1995, as 
supplemented by letters dated February 23, March 1, March 13, March 25, 
March 26, May 10, June 10, June 14, two letters dated June 25 and a 
letter dated June 26, 1996.
    Brief description of amendments: The proposed amendments closed out 
additional open items identified in the NRC staff's review of the 
upgrade of the Dresden and Quad Cities Technical Specifications (TS) to 
the Standard Technical Specifications (STS) contained in NUREG-0123. 
The Technical Specification Upgrade Program (TSUP) is not a complete 
adaptation of the STS. The TS upgrade focuses on (1) integrating 
additional information such as equipment operability requirements 
during shutdown conditions, (2) clarifying requirements such as 
limiting conditions for operation and action statements utilizing STS 
terminology, (3) deleting superseded requirements and modifications to 
the TS based on the licensee's responses to Generic Letter (GL), and 
(4) relocating specific items to more appropriate TS locations.
    Date of issuance: June 28, 1996
    Effective date: June 28, 1996
    Amendment Nos.: 150, 145, 171, and 167
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. 
The amendments revised the Technical Specifications and operating 
licenses.
    Date of initial notice in Federal Register: November 29, 1995 (60 
FR 61272) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 28, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021.

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

    Date of application for amendment: November 22, 1995 (NRC-95-0124)
    Brief description of amendment: The amendment revises the Technical 
Specifications to remove accelerated testing frequencies and special 
reporting requirements for Fermi 2 emergency diesel generators (EDGs) 
in accordance with guidance contained in Generic Letter 94-01, dated 
May 31, 1994. NRC will issue a separate safety evaluation on extending 
the allowed outage time for the EDGs at a later date.
    Date of issuance: June 20, 1996
    Effective date: June 20, 1996, with full implementation within 60 
days
    Amendment No.: 107
    Facility Operating License No. NPF-43. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7550) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated

[[Page 37305]]

June 20, 1996. No significant hazards consideration comments received: 
No
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161

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

    Date of application for amendments: December 12, 1995, as 
supplemented by letter dated June 10, 1996
    Description of amendment request: The amendments revise the 
absolute values in the Axial Flux Difference (AFD) Equations to reflect 
the proper AFD limit reduction in the current Technical Specifications.
    Date of issuance: July 2, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 167 and 149
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 24, 1996 (61 FR 
18166) The June 10, 1996, letter provided clarifying information that 
did not change the scope of the December 12, 1995, 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 July 2, 1996. No significant hazards 
consideration comments received: No
    Local Public Document Room location:  Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223

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

    Date of amendment request: August 11, 1995, as supplemented by 
letter dated February 12, 1996
    Brief description of amendment: The amendment reduced the minimum 
reactor coolant cold leg temperature to 541  deg.F from 544  deg.F in 
Technical Specification Section 3.2.6, ``Reactor Coolant Cold Leg 
Temperature.''
    Date of issuance: June 24, 1996
    Effective date: June 24, 1996
    Amendment No.: 120
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
25706) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 24, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

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

    Date of application for amendments: March 20, 1996, as supplemented 
by letter date April 23, 1996.
    Brief description of amendments: These amendments relocate the 
requirements for surveillance testing of the water level and pressure 
channel instrumentation for the reactor coolant system accumulators. 
These amendments also modify the existing action statements of TS 3.5.1 
for accumulators to reflect the requirements of NUREG-1431 by requiring 
a 72- hour period to restore boron concentration if it is not within 
the limits, and a 1-hour period to restore any other condition 
rendering the accumulators inoperable.
    Date of issuance: June 24, 1996
    Effective date: June 24, 1996
    Amendment Nos. 185 and 179Facility Operating Licenses Nos. DPR-31 
and DPR-41: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25707) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 24, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

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

    Date of application for amendments: February 21, 1996, as 
supplemented by letters dated May 1 and June 4, 1996.
    Brief description of amendments: The amendments revise the 
Technical Specifications to change the Drywell Air Temperature Limiting 
Condition for Operation (LCO) from less than or equal to 135 deg.F to 
less than or equal to 150 deg.F. The proposed change would provide a 
margin for the primary containment Drywell Air Temperature LCO when 
prolonged summer and high river temperatures are experienced. Also, a 
strictly editorial correction to a Final Safety Analysis Report 
reference would be made.
    Date of issuance: 201 and 142
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 201 and 142
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 24, 1996 (61 FR 
18167) The May 1 and June 4, 1996, letters provided clarifying 
information that did not change the scope of the February 21, 1996, 
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 June 27, 1996. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513

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

    Date of application for amendments: May 19, 1995, and supplemented 
October 20, 1995, and April 8, 1996 (AEP:NRC:1213A)
    Brief description of amendments: The amendments modify the neutron 
flux high setpoints for one or more main steam safety valves inoperable 
in response to Westinghouse Nuclear Safety Advisory Letter 94-001. The 
associated action statements are also revised and an exemption to TS 
4.0.4 is added to support the operability surveillance.
    Date of issuance: June 28, 1996
    Effective date: June 28, 1996, with full implementation within 45 
days.
    Amendment Nos.: Unit 1 - 210, Unit 2 - 195
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65681) The April 8, 1996, submittal provided information clarifying 
the location of the TS 4.0.4 exemption statement. This information was 
within the scope of the original application and did not alter the 
staff's no significant hazards considerations determination. Therefore 
renoticing was not warranted. The Commission's related evaluation of 
the amendments is contained in a Safety Evaluation dated June 28, 1996. 
No significant hazards consideration comments received: No.

[[Page 37306]]

    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: January 17, 1996
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) and associated Bases by relocating certain 
response time limit tables from the TSs to the Updated Safety Analysis 
Report in accordance with the guidance of NRC Generic Letter 93-08. The 
relocated tables are for instrumentation for the Reactor Protection 
System, Isolation Actuation System, Emergency Core Cooling System, and 
the Recirculation Pump Trip System.
    Date of issuance: June 25, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 73
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20850) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 25, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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: December 18, 1995
    Brief description of amendment: The amendment changes the Reactor 
Coolant Flow - Low Flow in Technical Specification Table 2.2-1, 
``Reactor Instrumentation Protective Trip Setpoint Limits.'' The 
proposed change increases the allowable value from greater than or 
equal to 90.1% to greater than or equal to 90.9% of the reactor coolant 
flow with four pumps operating. As an editorial change for 
clarification, the word ``flow'' is added after ``reactor coolant'' in 
the above sentence.
    Date of issuance: July 2, 1996
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 199
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 14, 1996 (61 
FR 5815) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 2, 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

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

    Date of application for amendment: June 27, 1995, as supplemented 
July 21, 1995
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) to relocate TS requirements for the containment 
purge exhaust and supply valves, and to remove a duplicate testing 
requirement for the safety injection input from engineered safety 
features from the TS.
    Date of issuance: June 27, 1996
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 129
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 6, 1995 (60 FR 
62494) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 27, 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

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

    Date of application for amendment: January 26, 1996
    Brief description of amendment: The amendment deletes three 
residual heat removal (RHR) system relief valves from Technical 
Specification (TS) Table 3.6.3-1, ``Primary Containment Isolation 
Valves.'' These valves are no longer needed to support the steam 
condensing mode of RHR and are being removed from the plant during the 
Unit 1 ninth refueling outage.
    Date of issuance: June 24, 1996
    Effective date: As of date of issuance to be implemented within 60 
days.
    Amendment No.: 157
    Facility Operating License No. NPF-14: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 27, 1996 (61 FR 
13531) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 24, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.

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 29, 1996
    Brief description of amendments: These amendments relocate 
Specification 3/4.9.6, ``Refueling Platform,'' to the Susquehanna Steam 
Electric Station Technical Requirements Manual, a document which is 
controlled under the requirements of 10 CFR 50.59.
    Date of issuance: July 2, 1996
    Effective date: July 2, 1996
    Amendment Nos.: 158 and 129
    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 July 2, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 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: March 12, 1996
    Brief description of amendment: The proposed changes would remove a 
requirement to cross tie safety injection accumulators.
    Date of issuance: July 3, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.

[[Page 37307]]

    Amendment No.: 167
    Facility Operating License No. DPR-64: 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 July 3, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: April 24, 1996
    Brief description of amendment: The amendment proposes to relocate 
Specification 3.11.B/4.11.B ``Crescent Area Ventilation'' and 
associated Bases from the TS to an Authority controlled procedure.
    Date of issuance: June 28, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 231
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25710) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 28, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendments: February 6, 1996
    Brief description of amendments: The amendments change the 
Technical Specifications to lower the 125 Volt Battery Charger 
surveillance amperage from at least 200 amps to at least 170 amps.
    Date of issuance: June 27, 1996
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment Nos. 183 and 164
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7556) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 27, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: June 26, 1995, as supplemented 
by letter dated February 2, 1996.
    Brief description of amendment: The amendment revised the allowed 
outage time for component cooling water motor operated containment 
isolation valves, moved the list of containment isolation valves from 
the technical specifications to the final safety analysis report, and 
allowed containment penetration check valves to be used as isolation 
devices.
    Date of issuance: June 28, 1996
    Effective date: June 28, 1996, to be implemented within 30 days of 
the date of issuance.
    Amendment No.: 113
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 30, 1995 (60 FR 
45187) The February 2, 1996, supplemental letter provided additional 
clarifying information and did not change the staff's original no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated June 28, 1996. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Dated at Rockville, Maryland, this 10th day of July 1996.
    For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear 
Reactor Regulation.
[Doc. 96-18007 Filed 7-16-96; 8:45 am]
BILLING CODE 7590-O1-F