[Federal Register Volume 60, Number 11 (Wednesday, January 18, 1995)]
[Notices]
[Pages 3669-3685]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-1026]



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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 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 December 12, 1994, through January 5, 1995. 
The last biweekly notice was published on January 4, 1995 (60 FR 494).

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 [[Page 3670]] 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 Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, 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 February 17, 1995, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves 
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, 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, 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 
[[Page 3671]] 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.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of amendment request: November 22, 1994.
    Description of amendment request: The proposed amendment would 
revise the allowable leak rate for the main steam isolation valves 
(MSIVs) from the current 11.5 standard cubic feet per hour (scfh) for 
each valve, to a maximum combined main steam line leak rate of 46 scfh.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The operation of Pilgrim Station in accordance with the 
proposed Amendment will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendment does not involve a change to structures, 
components, or systems which would affect the probability of an 
accident previously evaluated in the Pilgrim Updated Final Safety 
Analysis Report (UFSAR). The proposed amendment results in no change 
in radiological consequences of the design basis LOCA [loss-of-
coolant accident] as currently analyzed for Pilgrim Station. These 
analyses were calculated using the combined total leakage factor of 
46 scfh for determining acceptance to the regulatory limits for the 
offsite, control room, and Technical Support Center (TSC) doses as 
contained in 10CFR100 and 10CFR50, Appendix A, GDC 19. The proposed 
change does not compromise existing radiological equipment 
qualification, since the combined total leakage rate of 46 scfh has 
been factored into our existing equipment qualification analyses for 
10 CFR 50.49.
    2. The operation of Pilgrim Station in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    There is no modification to the MSIVs or other plant system or 
structure associated with this amendment which could impact their 
capability to perform their design function. The total MSIV leakage 
rate of 46 scfh is included in the current radiological analyses for 
the assessment of dose exposure following an accident. This proposal 
changes the allowable leakage rate from a per valve to a total 
combined line leakage acceptance criteria but does not change the 
cumulative allowable value. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously analyzed.
    3. The operation of Pilgrim Station in accordance with the 
proposed amendment will not involve a significant reduction in a 
margin of safety.
    The allowable leak rate limit specified for the MSIVs is used to 
quantify the maximum amount of bypass leakage assumed in the LOCA 
radiological analysis. Results of the analysis are evaluated against 
the dose guidelines contained in GDC [General Design Criteria] 19 
and 10CFR100. The margin of safety in this context is considered to 
be the difference between the calculated dose exposures and the 
guidelines provided by the GDC 19 and 10CFR100. Therefore, since the 
maximum allowable leakage for each valve was assumed and used as the 
total allowable leakage for the purpose of calculating potential 
dose, the margin of safety is not affected because the dose levels 
remain the same.

    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: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.
    Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
    NRC Project Director: Walter R. Butler.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of amendment request: November 22, 1994.
    Description of amendment request: The proposed amendment would 
revise the mode conditions under which the Scram Discharge Instrument 
Volume-Scram Trip Bypass in Table 3.2.C.1 is required to be operable 
and changes the associated functional test frequency from quarterly to 
once per operating cycle in Table 4.2.C.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The operation of Pilgrim Station in accordance with the 
proposed amendment will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change to Table 3.2.C.1, and the associated change 
to Table 4.2.C, removes incorrect reactor modes listed for the Scram 
Discharge Instrument Volume (SDIV)--Scram Trip Bypass function. The 
Pilgrim control rod block logic for the SDIV Bypass is not operable 
nor is it required by design when in the Run and Startup modes. The 
control logic and the FSAR [Final Safety Analysis Report] (section 
7.2.3.10) specifies SDIV--Scram Trip Bypass operability only in the 
Refuel and Shutdown modes.
    This change will not result in any physical modification or 
operation of the control rod block system. The change conforms the 
technical specifications to the actual design of the SDIV Scram Trip 
Bypass as described in the FSAR. Changing the functional 
surveillance frequency from quarterly to once per operating cycle 
also conforms the technical specifications to the applicable mode 
for the function.
    The change is classified as an administrative change because it 
corrects an administrative requirement that does not reflect the 
logic design. It improves safety by removing the need to install 
jumpers during reactor operations to perform unnecessary and 
potentially risky functional surveillances.
    Therefore, because this is an administrative change, operation 
of Pilgrim will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The operation of Pilgrim Station in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    This proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because it is administrative and requires no physical alteration of 
the plant configuration, changes to setpoints, or operating 
parameters.
    3. The operation of Pilgrim in accordance with the proposed 
amendment will not involve a significant reduction in a margin of 
safety.
    The proposed change serves to enhance the margin of safety by 
eliminating the potential for error caused by installing jumpers to 
the control logic during reactor operation. Changing the functional 
surveillance frequency from quarterly to once per operating cycle 
also enhances the margin of safety by allowing test performance off-
line, the mode for which the SDIV scram trip bypass control rod 
blocks are designed to be operable.

    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: Plymouth Public Library, 11 
[[Page 3672]] North Street, Plymouth, Massachusetts 02360.
    Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
    NRC Project Director: Walter R. Butler.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of amendment request: November 22, 1994
    Description of amendment request: The proposed amendment would 
revise the suppression chamber water level operating range, increasing 
it 2 inches, and revise the water level recorder range.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The operation of Pilgrim Station in accordance with the 
proposed amendment will not involve a significant increase in the 
probability or consequences of an accident previously identified.
    The probability of an accident is not increased by this proposed 
change because there is no relation between the Suppression Chamber 
water level operating range and the probability of an accident.
    The consequences of an accident identified are not increased. 
The Suppression Chamber is an accident mitigating device. Increasing 
the water level operating range has been analyzed and does not 
significantly increase the structural loads and the calculated 
stress levels remain within Mark 1 Acceptance Criteria.
    We have reviewed the FSAR [Final Safety Analysis Report] 
Containment Analyses and concluded that the safety margin is not 
affected. An increase in water level enhances the Suppression Pool's 
ability to mitigate an accident by providing more water for use by 
emergency cooling systems. The higher water level increases the sink 
capabilities resulting in lower torus water temperatures from steam 
blowdowns. There is a minor reduction in the free air volume of the 
torus which has a negligible effect on containment post accident 
pressures. Therefore, there is no significant increase in the 
probability or consequences of an accident previously identified.
    The change in water level recorder range does not involve an 
increase in the probability or consequence of an accident because 
the new recording range accounts for instrument loop uncertainties 
and is thus more conservative than the previous range.
    2. The operation of Pilgrim Station in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously analyzed.
    An increase in the Suppression Chamber water level operating 
range does not create a new or different kind of accident from any 
accident previously analyzed because the Suppression Chamber is an 
accident mitigating device. The Suppression Chamber serves as the 
heat sink for any postulated transient or accident condition when 
the primary heat sink (main condenser) is unavailable and as a 
source of water for the Core Standby Cooling Systems. The structural 
affects of the increase in water volume have been analyzed and do 
not significantly effect the Mark 1 containment loads.
    Revising the water level recording range is more conservative 
than that previously used and does not create the possibility of a 
new or different kind of accident.
    3. The operation of Pilgrim Station in accordance with the 
proposed amendment will not involve a significant reduction in the 
margin of safety.
    Operation with an increased Torus water level does not affect 
the structure and attached piping of the Pilgrim Suppression Chamber 
and does not significantly affect the calculated stress levels; 
therefore, there is no significant reduction in the margin of 
safety.
    The change in the water level recording range is due to 
replacing the transmitter with a smaller span. The change from 0 to 
32 inches to -7 to +7 inches enhances resolution and accuracy of the 
water level instrument loop.

    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: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.
    Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
    NRC Project Director: Walter R. Butler.

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

    Date of amendment request: August 30, 1994.
    Description of amendment request: The proposed amendment relocates 
refueling cycle specific parameters from the technical specifications 
to the Core Operating Limits Report as per recommendations promulgated 
by NRC Generic Letter 88-16. Additionally, the amendment adds a 24 hour 
limit on operations when only one reactor coolant pump is operating in 
each loop. Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Criterion 1--Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The relocation of cycle-specific variables from the Technical 
Specifications to the Core Operating Limits Report (COLR) is 
considered to be administrative in nature and has no impact on plant 
operation or safety. The Technical Specifications will continue to 
require operation within the core operational limits for each cycle 
reload as calculated by the NRC approved reload methodologies. The 
values and setpoints placed in the COLR are addressed in the reload 
report for each particular fuel cycle. The reload report presents 
the results of evaluations of accidents addressed in the ANO-1 
Safety Analysis Report. These evaluations demonstrate that changes 
in the fuel cycle design and the corresponding COLR do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The revision of Specification 3.1.1.1.a and addition of the 
footnote to Table 2.3-1 result in additional restrictions on 
operation with one reactor coolant pump in each loop with the 
reactor critical. This more restrictive specification limits 
operation with one reactor coolant pump in each loop to a 24 hour 
period when the reactor is critical. This change incorporates a more 
restrictive control and does not affect any previously analyzed 
event.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously.
    Criterion 2--Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    This relocation of cycle-specific variables from the Technical 
Specifications to the COLR does not create the possibility of a new 
or different kind of accident from any previously analyzed. The 
cycle-specific variables will continue to be calculated using NRC 
approved methodologies. Technical Specifications will continue to 
require operation within the required core operating limits and 
appropriate actions will be taken if the limits are exceeded. 
Because plant operation continues to be limited in accordance with 
the values of cycle-specific parameter limits that are established 
using NRC approved methodologies, the relocations included in this 
submittal are considered to be administrative in nature and have no 
impact on plant safety as a consequence.
    The revision of Specification 3.1.1.1.a and addition of the 
footnote to Table 2.3-1 result in additional restrictions on 
operation with one reactor coolant pump in each loop with the 
reactor critical. This more restrictive specification limits 
operation with one reactor coolant pump in each loop to a 24 hour 
period when the reactor is critical. This proposed change introduces 
no new mode of plant operation.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3--Does Not Involve a Significant Reduction in the 
Margin of Safety. [[Page 3673]] 
    The proposed relocations are considered to be administrative in 
nature and do not involve a significant reduction in the margin of 
safety since they only involve transferring limits from the 
Technical Specifications to the COLR. The values and setpoints 
placed in the COLR are addressed in the reload report for each 
particular fuel cycle. The development of limits for future reloads 
will continue to conform to methodologies described in NRC approved 
documentation. Each future reload involves a 10CFR50.59 safety 
review to assure that operation of the unit within the cycle-
specific limits will not involve a significant reduction in the 
margin of safety.
    The revision of Specification 3.1.1.1.a and addition of the 
footnote to Table 2.3-1 result in additional restrictions on 
operation with one reactor coolant pump in each loop with the 
reactor critical. This more restrictive specification limits 
operation with one reactor coolant pump in each loop to a 24 hour 
period when the reactor is critical. This change does not involve a 
significant reduction in the margin of safety, rather, it 
constitutes an additional limitation not previously included in the 
Technical Specifications.

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

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

    Date of amendment request: November 9, 1994.
    Description of amendment request: The proposed amendment revises 
those specifications associated with various engineered safety feature 
systems following a design basis fuel handling accident. The proposed 
changes affect conditions where irradiated fuel is handled in the 
primary or secondary containment and when fuel is handled over the 
reactor vessel with fuel in the vessel. These changes are based on a 
recent re-analysis of the fuel handling accident for Grand Gulf Nuclear 
Station (GGNS).
    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 proposed definition of RECENTLY IRRADIATED fuel is used to 
establish operational conditions where specific activities represent 
situations where significant radioactive releases can be postulated. 
These operational conditions are consistent with the design basis 
analysis. Because the equipment affected by the revised operational 
conditions is not considered an initiator to any previously analyzed 
accident, inoperability of the equipment cannot increase the 
probability of any previously evaluated accident. The proposed 
applicability 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 19. 
Therefore, the proposed changes do not significantly increase the 
probability or 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 proposed definition is used to establish operational 
conditions where specific activities represent situations where 
significant radioactive releases can be postulated. These 
operational conditions are consistent with the design basis 
analysis. The proposed changes do not introduce any new modes of 
plant operation and do not involve physical modifications to the 
plant. Therefore, the proposed changes do not create the possibility 
of a new or different kind of accident from any previous analyzed.
    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 revised definition is used to establish operational 
conditions where specific activities represent situations where 
significant radioactive releases can be postulated. These 
operational conditions are consistent with the design basis analysis 
and are established such 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 
proposed change only eliminates the excess margin from the analysis. 
The current margin of safety is retained.
    Specifically, the margin of safety for the fuel handling 
accident is the difference between the 10 CFR 100 limits 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 10 CFR 100 limits and the licensing limit 
defined by 10 CFR 50, Appendix A, Criterion 19 (GDC 19). Excess 
margin is the difference between the postulated doses and the 
corresponding licensing limit.
    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.
    Therefore, the proposed changes do not result in a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three 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, Mississippi 39120.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: December 6, 1994.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to allow the use of the Combustion 
Engineering sleeving process for repairing steam generator tubes. (The 
current requirement specifies that degraded steam generator tubes be 
repaired by plugging.)
    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 has reviewed the licensee's analysis against the 
standard of 10 CFR 59.92(c). The staff's review is presented below:
    1. The proposed amendment would not involve a significant increase 
in the [[Page 3674]] probability or consequences of an accident 
previously evaluated.
    With the sleeve dimensions, materials, and connecting joints 
designed to the applicable American Society of Mechanical Engineers 
Boiler and Pressure Vessel Code (ASME Code), the proposed sleeving 
repair becomes an in-kind substitution for the steam generator tube 
being repaired. The design criteria for the sleeves conform to the 
stress limits and safety margins of Code Section III. Safety factors of 
3 (normal operation) and 1.5 (accident conditions) were applied to the 
sleeve design. Mechanical testing using Code stress allowables also has 
been performed in support of the sleeve design. Based on the results of 
vendor test and analysis programs, the sleeves fulfill their intended 
function as leak tight structural members and meet or exceed all design 
criteria.
    Evaluation of the steam generator tubes and proposed sleeves 
indicates no detrimental effects on the sleeve or sleeve-tube assembly 
from reactor coolant system flow, reactor or steam generator coolant 
chemistry, or thermal or pressure conditions (including transients) 
that may be experienced by the Maine Yankee plant. Corrosion testing of 
sleeve-tube assemblies indicates no evidence of sleeve or steam 
generator tube corrosion considered detrimental under anticipated 
service conditions.
    Installation of the proposed sleeves will be controlled via 
Combustion Engineering's proprietary equipment and process. The process 
has been used 24 separate times since 1984 to install approximately 
4100 steam generator sleeves in nuclear facilities worldwide. The Maine 
Yankee steam generator design has been reviewed and found compatible 
with the sleeve installation equipment and process. Installation of the 
proposed sleeves will have no significant effect on either plant 
configuration or operation.
    The licensee therefore concludes that implementation of the 
proposed change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed amendment would not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    As discussed above, the structural integrity, thermal 
characteristics, and material properties of the proposed sleeves are 
compatible with Maine Yankee's steam generators. Therefore, the 
functions of the steam generators will not be significantly affected by 
installation of the proposed sleeves. In addition, the proposed sleeves 
do not interact with any other plant systems. Finally, the continued 
integrity of installed sleeves is periodically verified by the steam 
generator inspections required by plant Technical Specifications.
    The licensee therefore concludes that implementation of the 
proposed change will not create a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed amendment would not involve a significant reduction 
in a margin of safety.
    Repair of degraded steam generator tubes via the use of the 
proposed sleeves has been confirmed to restore the structural integrity 
of faulted tubes under normal operating and postulated accident 
conditions. The design safety factors used for the sleeves are 
consistent with ASME Code safety factors required in the design of 
Maine Yankee's steam generators. The repair limit for the proposed 
sleeves is consistent with that established for Maine Yankee's steam 
generators. The design of the sleeve-to-tube joint has been verified by 
testing to preclude significant leakage during normal and postulated 
accident conditions. Use of the previously identified design safety 
factors design verification testing assures that margin to safety with 
respect to installation of the proposed sleeves is not significantly 
different from the original steam generator tubes.
    The licensee therefore concludes that implementation of the 
proposed change would not involve a 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 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, Maine 04578.
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, 329 Bath Road, Brunswick, Maine 04011.
    NRC Project Director: Walter R. Butler.

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

    Date of amendment request: December 23, 1994.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) 2.1.2, ``Fuel Cladding 
Integrity,'' 3.6.2/4.6.2, ``Protective Instrumentation,'' and 
associated Bases to extend the calibration frequency of the reactor 
recirculation flow transmitters from once per quarter to once per 
operating cycle and for the square rooters and summers from once per 
quarter to once per year. The proposed amendment would revise the flow 
biased average power range monitor (APRM) scram and rod block, 
recirculation flow comparator, and flow unit upscale setpoints and the 
associated Bases of TSs 2.1.2, 2.2.2, and 3.6.2/4.6.2.
    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:

    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes extend the calibration interval for the 
recirculation flow square rooters, summers and transmitters and 
revise the setpoints for the recirculation flow upscale and 
comparator rod block trips. The associated analytical limits for 
APRM flow biased scram and rod block increase by 2% and 8% 
respectively. Setpoints are for plant protective functions (i.e., 
scram and rod block) which respond to an accident or transient. The 
scram and rod block function responds to mitigate the consequences 
of an accident or transient. Therefore, a change to the setpoints 
cannot increase the probability of these accidents or transients. 
Likewise, changes to surveillance intervals for the protective 
functions which respond to an accident or transient cannot increase 
the probability. In fact, the proposed increase in the surveillance 
intervals reduce the probability of an inadvertent scram by reducing 
the duration that the plant is in the one-half scram condition.
    The new surveillance intervals, setpoints and allowable setpoint 
deviations are calculated using the approved GE [General Electric 
Company] setpoint methodology documented in NEDC-31336. The 
methodology in NEDC-31336 provides assurance that safety system 
actuation (i.e., reactor scram or control rod withdrawal block) will 
occur prior to the associated system parameters (neutron flux and 
recirculation flow) exceeding their analytical limits. Based upon 
re-evaluation of NMP1 [Nine Mile Point Nuclear Station Unit No. 1] 
accidents and transients, it has been shown that the fuel thermal 
limits are not significantly impacted. Therefore, the consequences 
of an accident or transient has not significantly increased.
    Thus, plant response to previously analyzed accidents remains 
within previously determined limits. Therefore, the operation of 
Nine Mile Point Unit 1, in accordance with the proposed amendment, 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
[[Page 3675]]

    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes to extend the calibration frequency do not 
represent a physical change to the plant as described in the NMP1 
Final Safety Analysis Report (Updated). However, this change results 
in increasing the analytical limits for the APRM flow based scram 
and rod block by 2% and 8% respectively. The proposed changes do not 
alter the plant configuration and the initial conditions used for 
the design basis accident analysis are still valid. Thus, no 
potential initiating events are created which would cause any new or 
different kinds of accidents. As such, the plant initial conditions 
utilized for the design basis accident analysis are still valid. 
Therefore, operation of Nine Mile Point Unit 1 in accordance with 
the proposed change will not create the possibility of a new or 
different kind of accident from any previously assessed.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The analytical limits for the APRM flow biased scram and rod 
block increase by 2% and 8% respectively. The trip units in the APRM 
and recirculation flow instrumentation systems will continue to be 
calibrated every three months. In addition, the entire APRM and 
recirculation flow instrumentation systems will still be subject to 
Instrument Channel Tests every three months. These tests, together 
with the calibration of the flow square rooters and summers once per 
year and the flow transmitters once per operating cycle, will assure 
that system reliability and availability are maintained at their 
current levels. Reanalysis of the design basis transients was 
performed utilizing these new values. The results showed that the 
increase had an insignificant effect on the consequences of these 
events. Therefore, the proposed amendment will not involve a 
significant reduction in the margin of safety.


    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Michael J. Case

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

    Date of amendment request: December 13, 1994.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Table 3.6.1.2-1, ``Allowable Leak Rates 
Through Valves in Potential Bypass Leakage Paths,'' to increase the 
maximum allowable leakage rate of each of the eight main steamline 
isolation valves from 6.0 scfh to 24.0 scfh.
    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 operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes to Technical Specification Table 3.6.1.2-1 
would allow a maximum leakage of 24.0 scfh for each of the eight 
MSIVs [main steamline isolation valves]. The current Technical 
Specifications allow a maximum leakage for an MSIV of 6.0 scfh.
    Closure of one or more of the MSIVs at rated power is a pressure 
transient for the reactor coolant pressure boundary. This pressure 
transient is evaluated in Section 15.2.4 of the USAR [Updated Safety 
Analysis Report]. Closure of MSIV(s), as analyzed in the USAR, could 
occur due to manual or automatic actions. A change to the leakage 
limit for the MSIVs does not affect either the manual or automatic 
actions that would close the MSIVs. Therefore, the proposed change 
to the table cannot affect the probability of the closure of one or 
more MSIVs at rated power.
    The radiological evaluation of the DBA-LOCA [Design Basis 
Accident--Loss-of-Coolant Accident] incorporates a maximum leakage 
of 24.0 scfh for each of the four main steam lines. In addition, the 
revised radiological evaluation includes the impact of the proposed 
license amendment currently under review by the Staff which would 
increase the rated operation of NMP2 from 3323 to 3467 megawatts 
thermal (see NMPC letter dated July 22, 1993 to the NRC). The 
revised radiological evaluation also includes the impact of License 
Amendment No. 56 (see NMPC letter dated July 1, 1994 to the NRC and 
License Amendment No. 56, dated August 30, 1994).
    The new doses from the revised radiological analysis for a DBA-
LOCA, as shown in Table 1 [of December 13, 1994, amendment request], 
continue to remain below 10 CFR [Part] 100 guideline values and GDC 
[General Design Criterion] 19 limits. The impact of the increased 
MSIV leakage on vital area access and equipment qualification is 
minimal and acceptable. Therefore, operation with the proposed 
change to the Technical Specifications will not significantly 
increase the consequences of an accident previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The safety function of the MSIVs is to isolate the main steam 
lines in a timely manner to preclude the uncontrolled leakage of 
radioactive steam. This is accomplished by providing the MSIVs with 
the capability of rapidly closing automatically in response to 
various plant conditions. The increase in the leakage limit for the 
MSIVs from 6.0 scfh to 24.0 scfh will not inhibit the MSIVs' 
isolation function. Therefore, operation with the proposed increase 
in the MSIV leakage will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The revised radiological analysis follows the very conservative 
fuel failure and instantaneous release assumptions of RG [Regulatory 
Guide] 1.3, with the exception of regulatory position C.1.f as 
permitted by SRP [Standard Review Plan] Section 6.5.5, ``Pressure 
Suppression Pool as a Fission Product Cleanup.'' The Staff approved 
the use of SRP Section 6.5.5. as part of the licensing basis of NMP2 
in License Amendment No. 56.
    The revised radiological analysis incorporates the maximum 
allowable leakage limit of 24.0 scfh for each of the four main steam 
lines. The revised radiological analysis also includes the impacts 
of the proposed power uprate of NMP2 and License Amendment No. 56. 
The new doses from the revised radiological analysis remain below 
the Staff acceptance criteria of 10 CFR [Part] 100 guideline values 
and GDC 19 (see Table 1 [of December 13, 1994, amendment request]). 
Therefore, operation with the proposed changes to the Technical 
Specifications will not significantly reduce a margin of safety.
    Accordingly, as determined by the analysis above, this proposed 
amendment involves no significant hazards consideration.


    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Michael J. Case. [[Page 3676]] 

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

    Date of application for amendments: August 3, 1994.
    Description of amendment request: The proposed changes would delete 
a footnote in the Technical Specifications (TS) regarding snubber 
functional testing frequency and make permanent the current one-time 
snubber functional test frequency of 24 months.
    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, because the probability of a seismic or other dynamic 
event is independent of the surveillance period for snubber tests. 
The change does not introduce any failure mechanisms to the 
previously considered events. The consequences of an accident 
previously evaluated in the SAR [Safety Analysis Report] is not 
increased by the proposed revision to [t]he snubber TS. No physical 
changes are being made to the plant. The snubbers' role in 
mitigating the consequences of an accident is to provide restraint 
during seismic or other dynamic events while permitting the slow 
movement of piping and components during heatup and cooldown. The 
proposed TS change will not affect the snubbers ability to continue 
to perform this role for the following reasons: (1) Changing the 
inspection cycle to 24 months will not reduce the ability of the 
functional testing to confirm the operability of the snubber 
population. The original interval of 18 months was selected to 
accommodate the need to test snubbers that were inaccessible during 
normal operation. Since snubbers do not require preventative 
maintenance during the operating cycle, the additional time added by 
a 24 month operating cycle has minimal impact, if any, on snubber 
operability. (2) The requirement to monitor service life remains 
part of TS. The review of snubber service life records is a 
documentation review of the snubbers service life. If a snubber's 
service life would expire prior to the next scheduled review then 
the snubber is reconditioned, replaced or reevaluated to extend its 
service life. (3) Snubber functional testing has shown no failure 
mechanism which would be aggravated by an extension of the test 
interval to 24 months. A historical search of completed snubber 
functional STs was completed. The historical search indicated that 
even though the snubbers did not always meet the initial screening 
functional test criteria of the ST, the piping system was operable 
based on an engineering evaluation and there was no evidence of a 
time dependent failure mechanism. To ensure the snubber remains 
operational during the next operating cycle, snubbers not meeting 
the screening ST acceptance criteria are either replaced or 
reconditioned.
    (2) The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated because 
the proposed change does not involve operational procedure or 
physical changes to the plant. Since snubbers will continue to meet 
their design basis of protecting the piping and equipment during 
dynamic events, the possibility of a different type of accident will 
not be created.
    (3) The proposed change does not involve a significant reduction 
in a margin of safety. There may be a slight increase, if any, in 
the possibility of undetected snubber failures because of the 
increase in the interval of functional testing for snubbers; 
however, the historical data of previous snubber functional 
surveillance testing and the supporting engineering evaluations 
indicate that on those occasions where snubbers did not meet initial 
surveillance testing requirements, the piping systems were all 
operable. Therefore, the probability of occurrence of a malfunction 
of equipment is minimal and equipment important to safety (ITS) that 
use snubbers will continue to meet design requirements and the 
margin of safety will be unaffected.

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

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

    Date of amendment request: September 29, 1994.
    Description of amendment request: This amendment requests revision 
of Table 4.3.6-1 ``Control Rod Block Instrumentation Surveillance 
Requirements.'' The channel calibration frequencies for the Source 
Range Monitor (SRM) and the Intermediate Range Monitor (IRM) would be 
changed as follows: the up-scale and the down-scale trip functions on 
each instrument would be changed from Note ``SA'', once-per-184 days to 
note ``R'', once-per-refuel interval.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes involve no hardware changes, no changes to 
the operation of any systems or components, and no changes to 
existing structures. The revision of channel calibration frequencies 
for the SRM and IRM trip function portion of the control rod block 
instrumentation represent changes that do not affect plant safety 
and do not alter existing accident analyses.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed changes are procedural in nature concerning the 
calibration frequency of instrumentation that have historically 
shown little set point drift. The channel calibration methodology 
for the SRM and IRM control rod block trip functions remain 
unchanged. The proposed changes while slightly increasing the 
possibility of an undetected instrument error will not create a new 
or unevaluated accident or operating condition.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed changes are in accordance with recommendations 
provided by the NRC regarding the improvement of Technical 
Specifications. These changes will result in the perpetuation of 
current safety margins while reducing regulatory burden and 
decreasing equipment degradation.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070.
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: December 16, 1994. [[Page 3677]] 
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 
3.4 by removing the Limiting Conditions for Operation (LCOs) for the 
Turbine Overspeed Protection System (TOPS). Tables TS 4.1-1 and TS 4.1-
3 would also be revised to remove the surveillance requirements for the 
TOPS instrumentation and turbine valves. The TOPS and related 
requirements would be relocated to the Updated Safety Analysis Report 
(USAR).
    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:

    Significant Hazards Determination for Proposed Changes to TS 
3.4.c and Table TS 4.1-1 and Associated Bases Changes
    In accordance with 10 CFR Part 50, Section 50.91 and using the 
standards provided in Section 50.92, the proposed change has been 
reviewed to determine that no significant hazards exist as a result 
of this change. The analysis showed:
    (1) The proposed amendment will not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    The purpose of the Turbine Overspeed Protection System (TOPS) is 
to prevent an overspeed event, which is a precursor to a potential 
turbine-generated missile. Neither Transient Analyses nor Design 
Basis Accidents (DBAs) evaluated in the accident analyses contained 
in Chapter 14 of the Kewaunee Nuclear Power Plant (KNPP) Updated 
Safety Analysis Report (USAR) assume operation of the TOPS. The 
calculations and probabilities associated with USAR section 14.2.7, 
``Turbine Missile Damage to the Spent Fuel Pool,'' are not affected 
by this amendment. This amendment does not implement physical 
changes to the plant and does not change the KNPP's existing 
requirements. As a result, this change will not increase the 
probability of a previously evaluated accident.
    The purpose of the TOPS is preventative and it serves no 
function to mitigate the consequences of any accident previously 
evaluated. Therefore, removing the requirements associated with the 
TOPS from the TSs will not affect the consequences of an accident 
previously evaluated.
    (2) The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    This amendment does not involve any changes in the operational 
characteristics of the surveillance tests and will impose no new 
requirements. This change will simply relocate the same testing 
requirements from the KNPP Technical Specifications to the KNPP 
USAR. Since this change is administrative in nature, it will not 
create a new or different kind of accident from any accident 
previously evaluated.
    (3) The proposed amendment will not involve a significant 
reduction in the margin of safety.
    KNPP's USAR section 14.2.7, ``Turbine Missile Damage to the 
Spent Fuel Pool,'' will not be affected by this amendment. 
Relocating the TOPS and related requirements is a change that is 
administrative in nature and does not alter the intent of any 
requirements. Therefore it can be concluded that this change will 
not involve a significant reduction in the margin of safety.
    Significant Hazards Determination for Proposed Change to Table 
TS 4.1-3 and associated Basis Change
    In accordance with 10 CFR Part 50, Section 50.91 and using the 
standards provided in Section 50.92, the proposed change has been 
reviewed to determine that no significant hazards exist as a result 
of this change. The analysis showed:
    (1) The proposed amendment will not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    This amendment does not involve any changes in the operation or 
frequency of the turbine valve tests. This amendment will simply 
relocate the turbine valve testing requirements from the Kewaunee 
Nuclear Power Plant's (KNPP's) Technical Specifications (TSs) to the 
Updated Safety Analysis Report (USAR). This change is administrative 
in nature and therefore will not involve a significant increase in 
the probability or consequence of an accident previously evaluated.
    (2) The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    This amendment is administrative in nature and will not change 
any requirements. This change will simply relocate the requirements 
from the KNPP TSs to the USAR. The purpose of the turbine stop and 
governor valves is to control steam flow to the turbine. This 
amendment will not adversely affect the steam flow control 
capability of the turbine valves. Therefore, this change will not 
create the possibility of a new or different type of accident from 
any accident previously evaluated.
    (3) The proposed amendment will not involve a significant 
reduction in the margin of safety.
    This amendment will simply relocate the existing turbine valve 
testing requirements and will not result in any changes to the 
requirements. The KNPP will continue to follow the recommendations 
of WCAP 11525, ``Probabilistic Evaluation of Reduction in Turbine 
Valve Test Frequency.'' As a result, KNPP will continue to maintain 
acceptably low probabilities of turbine valve failure. Since the 
same requirements still exist and turbine valve testing will 
continue to be consistent with the recommendations of WCAP 11525, 
this amendment will not involve a significant decrease 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 Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497.
    NRC Project Director: Leif J. Norrholm.

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

    Date of amendment request: December 12, 1994.
    Description of amendment request: This amendment request proposes 
revising Technical Specifications 4.7.1.2.1.b.1 and 4.7.1.2.1.b.2 to 
clarify the surveillance requirements for verifying the correct 
required position for the valves in the auxiliary feedwater system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not affect the ability of the auxiliary 
feedwater system to perform its intended safety function. The 
changes are administrative in nature since they merely clarify the 
demonstration of operability required in the surveillance 
requirements.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    There are no new failure modes or mechanisms associated with the 
proposed changes. The changes are administrative changes to remove 
confusion when performing surveillance requirements to demonstrate 
operability.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    These proposed changes do not effect [sic] any technical 
specification margin of safety. The changes only provide 
clarification for performance of surveillance requirements.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration. 
[[Page 3678]] 
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Project Director: Theodore R. Quay

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.

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

    Date of amendment request: November 25, 1994.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 5.3.1.E to allow 2645 fuel assemblies to 
be stored in the fuel pool. This is an increase of 45 fuel assemblies 
from the current limit of 2600. The 45 additional storage locations 
currently exist in the racks in the fuel pool. They were included in 
the re-racking project allowed by License Amendment No. 76 but were not 
incorporated in the Technical Specifications since, at the time, it was 
believed they would not be needed.
    Date of publication of individual notice in Federal Register: 
December 20, 1994 (59 FR 65542).
    Expiration date of individual notice: January 19, 1995.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: December 22, 1994.
    Description of amendment request: The proposed amendment is a Line 
Item Technical Specifications Improvement and would revise the Cooper 
Nuclear Station Technical Specifications, definition 1.0.J, concerning 
entering an operational condition consistent with the wording proposed 
in NRC Generic Letter 87-09, ``Sections 3.0 and 4.0 of the Standard 
Technical Specifications on the Applicability of Limiting Conditions 
for Operation and Surveillance Requirements,'' dated June 4, 1987.
    Date of individual notice in the Federal Register: January 3, 1995 
(60 FR 153).
    Expiration date of individual notice: February 2, 1995.
    Local Public Document Room location: Auburn Public Library, 118 
15th Street, Auburn, Nebraska 68305.

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: August 11, 1994.
    Brief description of amendment: The amendment deletes the 
requirement to perform a 5-year interval hydrostatic test on the 
auxiliary coolant system critical headers from TS Section 4.1.3, Table 
4.1-3, Item 11.
    Date of issuance: December 28, 1994.
    Effective date: December 28, 1994.
    Amendment No.: 155.
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 23, 1994 (59 
FR 60379).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 28, 1994. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College, 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: July 28, 1994.
    Brief description of amendment: The amendment allows an increased 
limit for fuel enrichment. The changes allow for the storage of fuel 
with an enrichment not to exceed 4.95 + 0.05 w/o U-235 in the new and 
spent fuel storage racks.
    Date of issuance: January 5, 1995.
    Effective date: January 5, 1995.
    Amendment No. 156.
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 
45018).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 5, 1995.
    No significant hazards consideration comments received: No. 
[[Page 3679]] 
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College, Hartsville, South Carolina 29550.

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: October 4, 1994.
    Brief description of amendments: The amendments revised the 
Technical Specification to eliminate a compliance conflict when 
swapping the Centrifugal Changing (NV) pumps in Modes 4, 5, and 6. In 
eliminating the conflict, this amendment permits flexibility in the 
operation of the NV pumps during unit startup without a safety concern.
    Date of issuance: November 17, 1994.
    Effective date: To be implemented within 30 days from the date of 
issuance.
    Amendment Nos.: 152 and 134.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 13, 1994 (59 FR 
52003).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 17, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223

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

    Date of application of amendments: November 11, 1993, as 
supplemented February 23, April 12, and July 29, 1994.
    Brief description of amendments: The amendments reflect the 
consolidation of the Quality Verification Department with the Nuclear 
Generation Department that realigned the Nuclear Safety Review Board to 
report to the Senior Nuclear Officer, change an organizational unit 
term from ``group'' to ``division,'' modify titles of positions 
designated to approve modifications, clarify the responsibilities of 
the Safety Assurance Manager, and delete the requirement to perform an 
annual fire protection audit.
    Date of Issuance: January 4, 1995.
    Effective date: To be implemented within 30 days from the date of 
issuance.
    Amendment Nos.: 208, 208, and 205.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 5, 1994 (59 FR 
619). The February 23, April 12 and July 29, 1994, letters provided 
clarifying information that did not change the scope of the November 
11, 1993, application or the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 4, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691.

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power 
Station, Unit 2, Shippingport, Pennsylvania

    Date of application for amendment: February 16, 1994.
    Brief description of amendment: This amendment deletes the Appendix 
B Section 4.2.2 requirement to perform infrared aerial photography 
every other year.
    Date of issuance: January 5, 1995.
    Effective date: January 5, 1995.
    Amendment No: 65.
    Facility Operating License No. NPF-73. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 6, 1994 (59 FR 
34663). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 5, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

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

    Date of application for amendment: November 29, 1994, as 
supplemented by letters dated December 20 and 21, 1994.
    Brief description of amendment: The amendment deleted the 
requirement to perform the full complement of steam generator 
surveillances as outlined in the technical specifications (TSs) when 
the steam generators are subjected to special inspections that are in 
addition to the periodic inspections required by the TSs. This 
amendment is applicable only to the special steam generator inspection 
scheduled for January 1995.
    Date of issuance: January 5, 1995.
    Effective date: January 5, 1995.
    Amendment No.: 158.
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 5, 1994 (59 FR 
62416). The additional information contained in the supplemental 
letters dated December 20 and 21, 1994, was clarifying in nature and 
thus, within the scope of the initial notice and did not affect the 
staff's proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 5, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

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

    Date of application for amendment: June 17, 1994, as supplemented 
by letter dated August 17, 1994.
    Brief description of amendment: The amendment removed License 
Condition 2.C.(25)(b) and Attachment 2 to Facility Operating License 
No. NPF-29, ``Transamerica Delaval Inc. (TDI) Diesel Generator 
Maintenance and Surveillance Requirements (NUREG-1216, August 1985).''
    Date of issuance: January 4, 1995.
    Effective date: January 4, 1995.
    Amendment No: 114.
    Facility Operating License No. NPF-29. Amendment revises the 
License.
    Date of initial notice in Federal Register: September 14, 1994 (59 
FR 47167).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 4, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, Mississippi 39120.

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

    Date of application for amendments: September 23, 1993 and 
clarified July 25, 1994. The July 25, 1994 submission did not change 
the amendment described in the initial Federal Register notice. 
[[Page 3680]] 
    Brief description of amendments: This amendment makes changes to 
Technical Specification 6.2.3, Independent Safety Engineering group. 
The change maintains the requirement to perform independent technical 
reviews while providing increased flexibility to accomplish this 
function.
    Date of Issuance: December 22, 1994.
    Effective Date: December 22, 1994.
    Amendment Nos.: 69.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 27, 1993 (58 FR 
57851). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 22, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

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: October 20, 1994.
    Brief description of amendments: These amendments relocate the 
diesel fuel oil testing program requirements to Technical 
Specifications (TS) Section 6 and to the Bases section of the TS. Also 
added were actions statements to address diesel fuel oil which does not 
meet the program limits.
    Date of issuance: December 28, 1994.
    Effective date: December 28, 1994.
    Amendment Nos. 169 and 163.
    Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55870).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 28, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

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: October 20, 1994.
    Brief description of amendments: The amendments remove the schedule 
for withdrawal of reactor vessel material specimens from the Technical 
Specifications as discussed in Generic Letter 91-01.
    Date of issuance: December 28, 1994.
    Effective date: December 28, 1994.
    Amendment Nos. 170 and 164.
    Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 23, 1994 (59 
FR 60381).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 28, 1994.
    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. 50-424 and 
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of application for amendments: September 13, 1994, as 
supplemented by letter dated December 6, 1994.
    Brief description of amendments: The amendments replace Containment 
Systems Technical Specification (TS) 3.6.2.2 for the Spray Additive 
System with a new Emergency Core Cooling System (ECCS) TS 3.5.5 for the 
ECCS Recirculation Fluid pH Control System.
    Date of issuance: January 5, 1995.
    Effective date: Phase I to be implemented following Unit 2 Cycle 4 
refueling outage; Phase II to be implemented following Unit 1 Cycle 6 
refueling outage.
    Amendment Nos.: 77 and 56 Phase 1; 78 and 57 Phase II.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 26, 1994 (59 FR 
53840). The December 6, 1994, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 5, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830.

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

    Date of application for amendment: October 19, 1991, as 
supplemented March 9, April 27, and December 15, 1994.
    Brief description of amendment: The amendment establishes 
additional requirements for the availability of Local Power Range 
Monitors (LPRMs) associated with the Average Power Range Monitoring 
(APRM) system. These additional requirements further restrict the 
allowable number of out of service LPRM/APRM detectors in order to 
ensure a sufficient response to regional thermal hydraulic oscillations 
in the reactor core to prevent violation of the Minimum Critical Power 
Ratio (MCPR) safety limit. The amendment also identifies a lower bound 
MCPR operating limit for each cycle as identified in the Core Operating 
Limits Report. This limit shall be greater than or equal to 1.47.
    Date of Issuance: December 29, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 176.
    Facility Operating License No. DPR-16. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 13, 1991 (56 
FR 57697). The March 9, April 27, and December 15, 1994, letters 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated December 29, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

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: July 18, 1994.
    Brief description of amendments: The amendments revised TS Table 
4.3-1, Reactor Trip System Instrumentation Surveillance Requirements; 
TS 3.3.4, Turbine Governor Valves; and TS [[Page 3681]] 3.7.1.2, 
Turbine Driven Auxiliary Feedwater Pump, to remove one-time amendments 
that are no longer necessary. In addition, six minor editorial changes 
were made.
    Date of issuance: December 27, 1994.
    Effective date: December 27, 1994, to be implemented within 31 days 
of issuance.
    Amendment Nos.: Unit 1--Amendment No. 67; Unit 2--Amendment No. 56.
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 
45024).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 27, 1994.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Wharton County Junior College, 
J.M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.

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

    Date of application for amendment: June 30, 1994, as supplemented 
November 10, 1994.
    Brief description of amendment: The proposed amendment would add 
Operability Requirements, Limiting Conditions for Operations (LCO) and 
Surveillance Requirements for the Control Building Chillers.
    Date of issuance: December 29, 1994.
    Effective date: Date of issuance, to be implemented within 120 
days.
    Amendment No.: 205.
    Facility Operating License No. DPR-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39592). The additional information contained in the supplemental letter 
dated November 10, 1994, was clarifying in nature and did not change 
the NRC staff's initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 29, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, Iowa 52401.

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: November 15, 1993.
    Brief description of amendments: The amendments make various 
administrative and editorial changes to the Technical Specifications.
    Date of issuance: December 30, 1994.
    Effective date: December 30, 1994.
    Amendment Nos.: 186 and 172.
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67849).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 30, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

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: July 19, 1994.
    Brief description of amendments: The amendments remove the specific 
requirements for Types A, B, and C containment leakage rate tests from 
the Technical Specifications and replace these requirements with a 
requirement to perform Types A, B, and C testing in accordance with 
Appendix J to 10 CFR Part 50.
    Date of issuance: January 5, 1995.
    Effective date: January 5, 1995.
    Amendment Nos.: 187/173.
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 28, 1994 (59 
FR 49430).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 5, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

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: November 15, 1993, and 
supplemented October 7, 1994.
    Brief description of amendments: The amendments replace the current 
Technical Specification testing requirements for the Event V reactor 
coolant system pressure isolation valves with the requirements from 
ASME Boiler and Pressure Vessel Code, Section XI.
    Date of issuance: January 5, 1995.
    Effective date: January 5, 1995.
    Amendment Nos.: 188/174.
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 5, 1994 (59 FR 
623).
    At the request of the NRC, the licensee submitted the October 7, 
1994, supplement to clarify the new requirements. This supplement did 
not change the NRC's initial proposed no significant hazards 
considerations finding; therefore, renoticing was not warranted.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 5, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

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

    Date of application for amendment: August 26, 1994.
    Brief description of amendment: The amendment revises Technical 
Specification 4.3.3.c.(1) to permit a one-time extension of the second 
10-year service period for the primary containment integrated leakage 
rate (Type A) test. The one-time extension permits delaying the third 
Type A test of the second 10-year service period from the 1995 
refueling outage until the 1997 refueling outage. This delay will 
result in an interval of approximately 46 months between the second and 
third Type A tests of the second 10-year service period.
    Date of issuance: December 29, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 151. [[Page 3682]] 
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 28, 1994 (59 
FR 49431). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 29, 1994.
    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: April 22, 1994.
    Brief description of amendment: The amendment changes Table 3.3-9 
of the Technical Specifications by modifying the indicated measurement 
range for the neutron flux monitor on the remote shutdown panel. The 
amendment also includes some corrections of typographical errors in the 
Technical Specifications.
    Date of issuance: December 20, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 183.
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27059).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 20, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Thames Valley State Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360.

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: July 22, 1994.
    Brief description of amendment: The amendment (1) changes the title 
of Figure 3.1-5 to be consistent with the applicable Limiting Condition 
For Operation (LCO), (2) relocates the Chemical and Volume Control 
System (CVCS) valve position requirements to the Reactivity Control 
Systems--Shutdown Margin specifications, and (3) consolidates action 
statements to be expressed in the LCOs rather than in Surveillance 
Requirements. The amendment also clarifies the requirements for 
calculating the heat flux hot channel factor F<INF>Q(z) when using the 
base load option.
    Date of issuance: December 29, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 99.
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 
45029).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 29, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.

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 2, 1994, as supplemented 
August 25, 1994.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) to remove expired one-time extensions of 
surveillance, removes an obsolete definition of charging pump 
operability, and incorporates 11 line item improvements in accordance 
with the guidance provided in Generic Letter 93-05. Several editorial 
changes have been made to renumber TS pages and delete the blank pages 
from the TS.
    Date of issuance: January 3, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 100.
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 14, 1994, (59 
FR 47170).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 3, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County, 
Minnesota

    Date of application for amendments: October 3, 1994, as 
supplemented November 30, 1994.
    Brief description of amendments: The amendments revise Prairie 
Island Nuclear Generating Plant Technical Specification 4.6, ``Periodic 
Testing of Emergency Power Systems.'' Specifically, the amendments 
modify the emergency diesel generator (EDG) 24-hour load test 
requirements to provide an indicated load range of 103-110 percent of 
the continuous rating. These amendments also rephrase various EDG test 
requirements to provide clarity and delete the requirements to verify 
that the auto-connected loads do not exceed 3000 kilowatts (Unit 2 5100 
kilowatts).
    Date of issuance: January 5, 1995.
    Effective date: January 5, 1995, with full implementation within 30 
days.
    Amendment Nos.: 113 and 106.
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55877).
    The November 30, 1994, request provided additional clarification 
that was within the scope of the initial notice and did not affect the 
staff's proposed no significant hazards consideration findings.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 5, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: October 7, 1994.
    Brief description of amendment: The amendment (1) deletes the 
surveillance requirements contained in Technical Specification (TS) 
3.6(3)a for the raw water backup valves to the containment cooling 
coils, (2) deletes the surveillance requirements in TS 3.2, Table 3-5, 
item 6, for raw water valves, and (3) revises the basis of TS 2.4 to 
reflect these changes. [[Page 3683]] 
    Date of issuance: December 29, 1994.
    Effective date: December 29, 1994.
    Amendment No.: 166.
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55879).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 29, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.

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: April 5, 1994.
    Brief description of amendments: These amendments delete the 
frequency requirements for a number of audits listed under Technical 
Specification 6.5.2.8 and also remove the audit requirements for the 
Emergency Plan and the Security Plan since these requirements have been 
added to the respective plan documents. The TS changes included in the 
April 5, 1994, application were approved with the exception of those 
related to the fire protection and loss prevention programs. These 
proposed changes are still under evaluation by the staff and will be 
addressed in a future safety evaluation.
    Date of issuance: December 22, 1994.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment Nos.: 137 and 107.
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27061).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 22, 1994.
    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. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: September 16, 1994, as 
supplemented November 29, 1994.
    Brief description of amendment: The amendment revised Technical 
Specifications Section 6.0 (Administrative Controls) to reflect, in 
part, licensee management changes in the corporate organization. 
Specifically, the title of Executive Vice President--Nuclear Generation 
was changed to Executive Vice President and Chief Nuclear Officer and a 
new position, Vice President Regulatory Affairs and Special Projects, 
which reports to the Executive Vice President and Chief Nuclear 
Officer, was established. In addition, the list of Safety Review 
Committee (SRC) members, which was previously by job title, was deleted 
and replaced with a description of SRC membership requirements, 
including individual qualifications and the minimum number of SRC 
members was reduced from 8 to 6.
    Date of issuance: December 22, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 220.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 30, 1994 (59 
FR 50021).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 22, 1994.
    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.

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: August 4, 1994, as supplemented 
November 10, 1994.
    Brief description of amendment: The amendment revises requirements 
in the Technical Specifications (TSs) related to primary containment 
atmosphere monitoring and drywell to torus differential pressure. 
Specifically, TS 3.7.A.6. has been revised to adopt primary containment 
inerting/deinerting requirements that are consistent with NUREG-1433, 
``Standard Technical Specifications--General Electric Plants, BWR/4.'' 
TSs 4.7.A.6.a. and 4.7.A.7.a. have been revised to provide frequencies 
for the verification of primary containment oxygen concentration and 
pressure differential between the drywell and torus. TSs 3.7.A.7.a.(1), 
3.7.A.7.a.(3), and 3.7.A.8. have been revised to provide requirements 
for establishing and maintaining differential pressure between the 
drywell and torus that are consistent with NUREG-1433. TS 3.7.A.9. has 
been deleted and related requirements have been incorporated into Notes 
for Table 3.2-8. Several administrative changes to Tables 3.2-8 and 
4.2-8 have also been made to improve the overall quality of the TSs.
    Date of issuance: December 28, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 221.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 
45032).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 28, 1994.
    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.

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: September 16, 1994, as 
supplemented November 29, 1994.
    Brief description of amendment: The amendment revised Technical 
Specifications Section 6.0 (Administrative Controls) to reflect, in 
part, licensee management changes in the corporate organization. 
Specifically, the title of Executive Vice President--Nuclear Generation 
was changed to Executive Vice President and Chief Nuclear Officer and a 
new position, Vice President Regulatory Affairs and Special Projects, 
which reports to the Executive Vice President and Chief Nuclear 
Officer, was established. In addition, the list of Safety Review 
Committee (SRC) members, which was previously by job title, was deleted 
and replaced with a description of SRC membership requirements, 
including [[Page 3684]] individual qualifications and the minimum 
number of SRC members was reduced from 8 to 6.
    Date of issuance: December 22, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 156.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 30, 1994 (54 
FR 50021).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 22, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendments: September 29, 1994.
    Brief description of amendments: The amendments change the 
Technical Specification surveillance interval for performing an air or 
smoke flow test through each containment spray header from 5 to 10 
years.
    Date of issuance: December 27, 1994.
    Effective date: December 27, 1994.
    Amendment Nos. 163, 144.
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 23, 1994 (59 
FR 60385).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 27, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.

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: April 18, 1994, as supplemented 
October 26, 1994.
    Brief description of amendment: The amendment revises Sections 2.C 
and 2.D of the San Onofre Nuclear Generating Station, Unit 1 (SONGS 1) 
Operating License. Section 2.C will be revised to modify or delete 
several licensing conditions which either no longer apply or require 
revision to apply to SONGS 1 in its permanently shutdown and defueled 
condition. Section 2.D will be revised to exempt Fire Protection 
reporting from the reporting requirements of Section 2.D.
    Date of issuance: December 22, 1994.
    Effective date: January 21, 1995.
    Amendment No.: 156.
    Facility Operating License No. DPR-13: The amendment revised the 
license conditions.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27066).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 22, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P.O. Box 19557, 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: October 20, 1994.
    Brief Description of amendments: The amendments delete the 
requirements for the control room chlorine detection system from the TS 
and the associated Bases Sections. This request is based on the fact 
that all stored gaseous chlorine has been removed from the plant site 
except for containers having an inventory of 150 pounds or less.
    Date of issuance: December 28, 1994.
    Effective date: December 28, 1994.
    Amendment Nos.: 111 and 102.
    Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: November 23, 1994 (59 
FR 60386).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 28, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

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

    Date of application for amendments: September 9, 1994 (TS 94-04).
    Brief description of amendments: The amendments revise the 
technical specifications related to the cold leg injection 
accumulators.
    Date of issuance: December 27, 1994.
    Effective date: December 27, 1994.
    Amendment Nos.: 192 and 184.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: October 12, 1994 (59 FR 
51629).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 27, 1994.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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

    Date of application for amendment: November 2, 1994 (TS 94-17).
    Brief description of amendment: The amendment adds Operating 
License Condition 2.C.(25) to provide a limited extension of the 
surveillance test intervals for certain specified instrumentation on 
Unit 1 to coincide with the Cycle 7 refueling outage. The surveillance 
intervals that are affected are specified in the attached safety 
evaluation and are for tests that would be extended to October 1, 1995, 
and would result in extension of the specified 18-, 36- and 54-month 
surveillances to 29.5, 48 and 71.5 months, respectively.
    Date of issuance: January 3, 1995.
    Effective date: January 3, 1995.
    Amendment No.: 193.
    Facility Operating License Nos. DPR-77: Amendment revises the 
operating license.
    Date of initial notice in Federal Register: November 23, 1994 (59 
FR 60387).
    The Commission's related evaluation of the change to the operating 
license is contained in a Safety Evaluation dated January 3, 1995.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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

    Date of application for amendments: October 11, 1994. 
[[Page 3685]] 
    Brief description of amendments: The amendments revise the 
surveillance frequencies of the hydrogen analyzer channel functional 
test and channel calibration.
    Date of issuance: December 23, 1994.
    Effective date: December 23, 1994.
    Amendment Nos. 195 and 195.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55893).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 23, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.

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

    Date of amendment request: July 22, 1994.
    Brief description of amendment: This amendment revises Section 6 of 
the Technical Specifications to reflect title changes in the Wolf Creek 
Nuclear Operating Corporation organization.
    Date of issuance: December 29, 1994.
    Effective date: December 29, 1994.
    Amendment No.: 81.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 26, 1994 (59 FR 
53845).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 29, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.

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

    Date of amendment request: February 23, 1994.
    Brief description of amendment: This amendment revised Technical 
Specifications 3.8.1.1, ``AC Sources Operating,'' and 3.8.1.2, ``AC 
Sources Shutdown,'' to increase the minimum volume of fuel oil required 
for the emergency diesel generator fuel oil day tanks. Several other 
revisions are included that make editorial corrections and incorporate 
requirements that were inadvertently omitted from previous amendment 
requests that have been approved.
    Date of issuance: December 29, 1994.
    Effective date: December 29, 1994.
    Amendment No.: 82.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17609).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 29, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.

    Dated at Rockville, Maryland, this 10th day of January 1995.

    For the Nuclear Regulatory Commission.
Jack W. Roe,
Director, Division of Reactor Projects--III/IV, Office of Nuclear 
Reactor Regulation
[FR Doc. 95-1026 Filed 1-17-95; 8:45 am]
BILLING CODE 7590-01-P