[Federal Register Volume 60, Number 21 (Wednesday, February 1, 1995)]
[Notices]
[Pages 6296-6324]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-2350]



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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 January 5, 1995, through January 20, 1995. 
The last biweekly notice was published on January 18, 1995 (60 FR 
3669).

NOTICE OF CONSIDERATION OF ISSUANCE OF AMENDMENTS TO FACILITY OPERATING 
LICENSES, PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION, 
AND OPPORTUNITY FOR A HEARING

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the 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. [[Page 6297]] 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 March 3, 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 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 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 
increase the current Emergency Diesel Generator (EDG) allowed out-of-
service time in Specification 3.5.F from 72 hours to 7 days, deletes 
the daily testing of the operable diesel generator in Specification 
4.5.F.1, when it is determined that the other diesel generator is 
inoperable, and revises specification 3.9.B.1 and 2 for EDG 
operability.
    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:

     [[Page 6298]] (1) The proposed amendment does not involve a 
significant increase in the probability of consequences of an 
accident previously evaluated.
    Operation of PNPS [Pilgrim Nuclear Power Station] in accordance 
with the proposed license amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Implementation of the proposed change is expected to result in 
an increase in the probability of core damage, from 5.85E-5/year 
(this is the PNPS IPE [individual plant examination] core damage 
frequency) to 5.88E-5/year. This increase is less than one percent 
and is considered to be insignificant relative to the underlying 
uncertainties involved with probabilistic risk assessments.
    Deleting the testing requirement for an EDG when the other EDG 
is in repair does not increase the probability or consequences of an 
accident previously evaluated because the reliability program and 
Technical Specification required surveillances continue to provide 
the added assurance sought by the testing. The elimination of this 
testing might improve the overall reliability of the EDGs.
    (2) The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Operation of PNPS in accordance with the proposed license 
amendment will not create the possibility of a new or different kind 
of accident from any accident previously evaluated. No change is 
being made in the manner in which the EDG's provide plant 
protection. No new modes of plant operation are involved. Extending 
the EDG OOS [out of service] and, deleting the testing requirement 
for one EDG when the other EDG is in repair does not necessitate 
physical alteration of the plant or changes in plant operational 
limits.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    Operation of PNPS in accordance with the proposed license 
amendment will not involve a significant reduction in a margin of 
safety. [***], incorporation of the proposed change involves an 
insignificant reduction in the margin of safety.
    As previously stated, implementation of the proposed changes is 
expected to result in an insignificant increase in: (1) power 
unavailability to the emergency buses (given that a loss of offsite 
power has occurred), and (2) core damage frequency. EDG reliability 
improvement is expected due to increased quality and thoroughness of 
EDG maintenance. Implementation of the proposed changes does not 
increase the consequences of a previously analyzed accident nor 
significantly reduce a margin of safety. Functioning of the EDGs and 
the manner in which limiting condition of operability are 
established are unaffected.

    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.

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

    Date of amendment request: December 27, 1994.
    Description of amendment request: The requested Technical 
Specifications (TS) change relocates the turbine rotor inspection 
requirement, TS 4.1-3, Item 13, to the Updated Final Safety Analysis 
Report (UFSAR), Section 10.2. This TS requires a turbine inspection, 
including visual, magnaflux, and dye petrant inspections on a frequency 
of every five years with a maximum time between tests of six years.
    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 requested change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The requested change relocates the turbine inspection 
requirement from the TS to the UFSAR. Turbine inspections will 
continue to be controlled and performed such that the low turbine 
missile generation probability will be maintained. The consequences 
of missile generation are unchanged since this change does not 
involve the addition or modification of plant equipment, nor does it 
alter the design or operation of plant systems. Therefore, there 
would be no increase in the probability or consequences of an 
accident previously evaluated.
    2. The requested change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The requested change relocates the turbine inspection 
requirement from the TS to the UFSAR. Turbine inspections will 
continue to be controlled and performed such that the low turbine 
missile generation probability will be maintained. This change does 
not involve the addition or modification of plant equipment, nor 
does it alter the design or operation of plant systems. Therefore, 
the proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The requested change does not involve a significant reduction 
in the margin of safety. The requested change relocates the turbine 
inspection requirement from the TS to the UFSAR. Turbine inspections 
will continue to be controlled and performed such that the low 
turbine missile generation probability will be maintained. 
Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.
    NRC Project Director: William H. Bateman.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
North Carolina

    Date of amendment request: December 19, 1994.
    Description of amendment request: The proposed one-time schedular 
extension would allow the third test of the first 10-year service 
period to be performed during refueling outage no. 7, at approximately 
a 54 month interval instead of the current maximum Technical 
Specification interval of 50 months, and coincident with the 10-year 
service period to be performed during refueling outage no. 7 and the 
10-year inservice inspection,
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This [extension] request applies to the ILRT [integrated leak 
rate testing] and does not affect the local leak rate testing of 
containment penetrations and isolation valves where the majority of 
the leakage occurs. The allowable containment leakage used in the 
accident analysis for offsite doses, La, is 0.1 [weight percent 
per day] and for conservatism the leakage is limited to 75 percent 
La at startup to account for the possible degradation of 
containment leakage barriers between two ILRT tests. Based on the 
``as left'' leakage data for the past two ILRTs, the additional time 
period added to the testing interval would not adversely impact the 
containment leakage barriers to the extent [[Page 6299]] that 
degradation would cause leakage to exceed that assumed in the 
accident analysis.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The change to the Surveillance Requirement is a one time 
[extension] to extend the surveillance interval from the maximum of 
50 months to approximately 54 months for performance of the third 
ILRT in the first service period. There are no design changes being 
made that would create a new type of accident or malfunction and the 
method and manner of plant operation remain unchanged. Extension of 
the surveillance interval for performing the ILRT does not adversely 
impact the surveillances ability to show that containment integrity 
is maintained.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    There are no changes being made to the safety limits or safety 
system settings that would adversely impact plant safety. The change 
is a one time [extension] to extend the time interval for performing 
an ILRT approximately four months beyond the current maximum 
interval. In addition to the indication of continued containment 
integrity provided by the Local Leak Rate Testing program, the 
surveillance test data from the first and second ILRTs illustrates 
that there is sufficient leakage margin to remain well below the 
allowable leakage rate of La. The as-left leakage rate for the 
last ILRT was 0.0614 [weight percent per day], which is well below 
the 0.075 [weight percent per day] allowed by the T.S., and 
therefore provides margin for degradation that is greater than the 
minimum provided by the Technical Specifications. Therefore, this 
change does not significantly reduce the margin of safety for 
Technical Specification 3.6.1.2.
    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: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: December 29, 1994.
    Description of amendment request: The proposed amendment would 
affect the method of controlling the pH of the post-LOCA containment 
sump solution by allowing the replacement of the existing operator 
actuated Iodine Removal System with a passive system of baskets of 
Trisodium Phosphate (TSP) in the lower regions of the containment. The 
current Iodine Removal System provides sodium hydroxide (NaOH) for 
injection into the containment spray to maintain pH of the sump 
solution.
    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 following evaluation supports the finding that operation of the 
facility in accordance with the proposed change from NaOH to TSP 
requirements would not:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The substitution of TSP baskets for the NaOH addition equipment 
would not cause any changes to the capability, settings, or 
operation of the plant systems (other than the Iodine Removal System 
itself) and would not, therefore, have any effect on the probability 
of occurrence of an accident.
    The substitution of TSP baskets for the NaOH addition equipment 
has the effect of providing more immediate control of post-LOCA sump 
pH, thereby increasing the assurance that iodine will remain in 
solution throughout a postulated event. The consequences of 
accidents evaluated in the FSAR [Final Safety Analysis Report] will 
not be increased by this increased assurance.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The TSP baskets are passive components which have no interaction 
with plant equipment unless flooding occurs in the containment. They 
are designed and located such that they will not interact with any 
plant safety equipment during a seismic event. The NaOH equipment, 
which will be replaced by the TSP baskets, has no function or effect 
on other equipment except during accident conditions. Therefore, the 
substitution of TSP baskets for NaOH addition equipment cannot 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The substitution of TSP baskets for the NaOH addition equipment 
would assure that the sump pH at the initiation of RAS 
[recirculation actuation signal] is between 7.0 and 8.0 as assumed 
in the MHA [maximum hypothetical accident] analysis. Therefore, this 
change would not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: December 7, 1994.
    Description of amendment request: The amendments revise the 
Technical Specification action statement to allow the Control Room Air 
Intake to remain open when radiation monitors (EMF-43A and EMF-43B) are 
inoperable. Immediate action to return the monitors to service would be 
required.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment would not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated in the FSAR [Final Safety Analysis Report].
    The amendment change will ensure correct Control Room 
Ventilation system alignment in order to mitigate the consequence of 
a Design Basis LOCA as described in FSAR Section 15.6.5.3, 
Environmental Consequences of a Loss of-Coolant Accidents, Control 
Room Operator Dose.
    The amendment change will permit the intake to remain open and 
will specify that action to repair the affected monitor shall be 
taken immediately. The change itself is not considered to be an 
initiator of any previously evaluated accident. Maintaining the VC 
intake open with an inoperable monitor will not result in any 
accidents that have not been previously evaluated. The 
implementation of immediate actions to repair the inoperable monitor 
does not in itself represent any accidents that have not been 
previously evaluated. Therefore, the proposed Technical 
Specification change does not increase the occurrence probability of 
previously evaluated accidents.
    The change to permit maintenance of open intakes will not 
increase the consequences of any previously evaluated accidents. The 
proposed amendment change is consistent with the original Safety 
Analysis concerning the Dose to the Operators.
    The analysis determined that the Doses to the Operators were 
within acceptable ranges given the assumptions that the intakes 
would [[Page 6300]] remain open and the contaminated air was 
processed through a Safety Related filter train prior to 
introduction into the Control Room. The proposed change remains 
consistent with this analysis and does not change the assumptions or 
methodology utilized to assess the Doses to the Operators for a 
hypothesized DBA; therefore, the proposed amendment change will not 
increase the consequences of any previously evaluated accident.
    2. The proposed amendment would not create the possibility of a 
new or different kind of accident not previously evaluated.
    The proposed change will not modify, delete, or add any systems 
or components; therefore, no new failure modes or accidents 
scenarios will be created.
    No test or experiments will be revised; therefore, no new 
initiating events or unanalyzed condition will be created. 
Administrative changes to surveillance procedures will be minor and 
will not create a safety concern.
    3. No significant reduction in a margin of safety will occur.
    The proposed amendment change requiring immediate action to 
initiate repairs to an inoperable monitor does not impact existing 
Safety Margins. Since requirements for immediate corrective action 
does not currently exist within the Specification, the changes will 
enhance the availability of the subject monitors.
    The proposed amendment does not change/impact any assumption or 
methods utilized to assess the doses to the operators for a 
hypothetical worst case DBA. Accordingly, the proposed amendment 
does not reduce any safety margins.

    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: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223.
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: December 9, 1994.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) by revising the allowable 
opening tolerances on the Pressurizer Code Safety Valves and the Main 
Steam Line Code Safety Valves from plus or minus 1% to plus or minus 
3%. This request is submitted as a result of an effort to improve valve 
performance and to ensure that the TS limits are consistent with 
expected valve performance capabilities.
    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 any change to the 
physical characteristics of the PSVs [pressurizer safety valves] and 
MSSVs [main steam safety valves] and will have no impact on the PSVs 
and MSSVs as-left setting. This change only allows for a larger 
(plus or minus 3% versus plus or minus 1%) as-found setpoint 
tolerance. Therefore, this change has no impact on the probability 
of occurrence of any accident previously evaluated. The impact of 
this change on the FSAR [final safety analyses report] analyses has 
been evaluated and the results of the impacted events have been 
found to be within the acceptable limits.
    Therefore, revising the PSV and MSSV as-found opening setpoint 
tolerance from plus or minus 1% to plus or minus 3% does not 
increase the probability or consequences of an accident previously 
evaluated.
    2. The proposed changes to the PSVs and MSSVs as-found opening 
setpoint tolerance do not modify equipment or change the manner in 
which the plant will be operated. The safety valves will continue to 
function per their design. Since no hardware modifications or 
changes in operation procedures will be made, the proposed changes 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The impact of the proposed changes on the Waterford 3 FSAR 
analyses have been evaluated. The evaluation demonstrates that the 
results of the impacted events remained within the acceptable 
limits. The system capabilities to mitigate and/or prevent accidents 
will be the same as they were prior to these changes. Therefore, the 
proposed changes do not involve a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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 New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: December 9, 1994.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) by revising a plant 
protection system (PPS) trip setpoint and several allowable values such 
that they will be consistent with the current setpoint/uncertainty 
methodology being implemented at Waterford 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Implementing the proposed change will not affect any design 
basis accident. The revised Trip Setpoint and Allowable Values are 
based upon the same Analytical Limits that form the basis for the 
current Trip Setpoints and Allowable Values. The design basis for 
each Trip Setpoint was verified to be consistent with the 
appropriate accident analyses as part of the process of revising the 
PPS setpoint analysis. The proposed change would implement a new 
Trip Setpoint for the Reactor Coolant (RC) System Low Flow Reactor 
trip and new Allowable Values for RC Low Flow, HI Log Power, HI 
Steam Generator Water Level, HI Containment Pressure, Low 
Pressurizer Pressure, Low Steam Generator Pressure, Low Steam 
Generator Water Level, and Low RWSP [refueling water storage pool] 
Level, based on the results of calculation EC-I92-019. The revised 
Low RC Flow Trip Setpoint is based on the same analytical limit as 
the current setpoint. The revised calculation uses the same design 
inputs with a similarly based methodology to calculate a smaller 
loop uncertainty. This results in a revised RC Low Flow Trip 
Setpoint that retains the original analysis limit. Therefore, the 
proposed change will not involve a significant increase in the 
probability or consequences of any previously analyzed accident.
    Plant operation and the manner in which the plant is operated 
will not be altered as a result of implementing the proposed change 
since no new system or design change is being implemented. The 
proposed Setpoint and Allowable Value changes do not create any new 
system interactions or interfaces. All information used to calculate 
the new Trip Setpoint is consistent with that of the existing 
accident analyses, and no new system interfaces/interactions are 
created. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed setpoint change revised the point at which the RCS 
Low Flow reactor trip initiates a reactor trip. The Trip Setpoint is 
based on the same Analytical Limit used to determine the current 
setpoint. In addition, the same basic setpoint determination 
[[Page 6301]] methodology is employed. That is, the Trip Setpoint is 
the Analytical Limit plus or minus the Total Loop Uncertainty [TLU]. 
The Allowable Value is the Trip Setpoint plus or minus the Periodic 
Test Error [PTE]. The change in the setpoint and allowable values 
are [sic] due to a change in calculated TLU and PTE. The proposed 
Trip Setpoint and Allowable Values are based on the same Analytical 
Limits for the affected parameters and are determined using approved 
methodology. Therefore, the proposed change will not involve a 
significant reduction in 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 New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: December 27, 1994.
    Description of amendment request: The proposed amendments would 
revise the period for conducting leak testing of containment purge 
valves to every refueling outage.
    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 to the Technical Specifications does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated because the [test] results have 
demonstrated that the resilient seat material does not degrade and 
cause containment isolation valves to leak. Therefore the valves 
will perform as assumed in the accident analyses.
    2. The proposed change to the Technical Specifications does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated because it does not require the 
valves to function in any manner other than that which is currently 
required.
    3. The proposed addition to the Technical Specifications does 
not involve a significant reduction in a margin of safety because it 
only affects the frequency of the test and does not change the 
leakage acceptance criteria. Since sufficient data has been 
collected to demonstrate that the resilient seals do not degrade, 
testing at the same frequency as other containment isolation valves 
will not reduce the margin of safety provided by the Technical 
Specifications.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia 30830.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: December 29, 1994.
    Description of amendment request: This request withdraws a similar 
request dated January 22, 1993, as supplemented August 8, 1993, and 
submits a new one in its place. The proposed amendments would revise 
the Technical Specifications (TS) to add the automatic load sequencer 
specification to TS Section 3/4.3, Instrumentation, and associated 
Bases, and TS Section 3/4.8, Electrical Power Systems.
    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 to the Technical Specifications does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated because the action to be taken when 
an automatic load sequencer is inoperable is consistent with that of 
a more stringent condition already specified, namely, the loss of an 
entire train of emergency power during Modes 1-4, and for Modes 5 
and 6 adding specific actions which previously had never been 
addressed in TS.
    2. The proposed change to the Technical Specifications does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated because it does not involve any 
change to the design, operation, or performance of the automatic 
load sequencer. It only serves to clearly identify the appropriate 
conservative response to an inoperable automatic load sequencer 
applicable to the plant mode of operation.
    3. The proposed change to the Technical Specifications does not 
involve a significant reduction in a margin of safety because the 
proposed actions to take when an automatic load sequencer is 
inoperable [are] the same as the action already required by the 
Technical Specifications when no power is available to the entire 
emergency bus during Modes 1-4 and by adding requirements during 
Modes 5 and 6, which had previously never been addressed.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia 30830.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: January 3, 1995.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) with editorial changes to the 
Action Statements of TS Sections 3.8.1.1 and 3.8.1.2 in order to 
reflect the availability of a third offsite ac electrical source. 
Surveillance Requirement 4.8.1.1.1 is being clarified to distinguish 
that the offsite ac circuits which are connected to the onsite Class 1E 
distribution system are required to be verified OPERABLE. The 
amendments also modify the Technical Specifications with the addition 
of a footnote to TS Section 3.8.3.1, to allow the connection of the 
third offsite ac source to the onsite busses.
    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: [[Page 6302]] 
    Based on the considerations regarding the addition of a footnote 
for proper bus alignment during operating conditions, the licensee 
submitted the following analysis in accordance with 10 CFR 50.92.

    1. The proposed change to the Technical Specifications does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated because the probability of an LOSP 
or an SBO is not increased by the allowance of having both redundant 
emergency busses of 4160 volt switchgear connected to one offsite 
source (RAT). The probability of having an LOSP is not increased 
since the TS currently allow for a 72 hour LCO for one offsite power 
source and the time the two redundant 4160 volt safety busses will 
be temporarily aligned to one RAT is well within this time frame. 
During this time the busses are interconnected, each bus is provided 
adequate protection and separation by having separate and redundant 
Class 1E circuit breakers, one per bus. The probability of an SBO is 
not increased since neither bus' EDG will be affected during this 
operation, and since this is a proceduralized manual alignment, the 
interconnection to one RAT will not be initiated if either EDG were 
inoperable. Also, the addition of the new ``swing'' offsite power 
source (SAT), increases availability and flexibility of the VEGP 
response to either an LOSP or SBO.
    2. The proposed change to the Technical Specifications does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated because the only postulated 
adverse consequences of tying both redundant 4160 volt safety busses 
together to one RAT is an LOSP. An LOSP is a design basis event 
which has already been analyzed for VEGP. In response to an LOSP, 
both EDGs remain capable of carrying the required loads to mitigate 
the consequences of any postulated design basis accident during or 
coincident with an LOSP.
    3. The proposed addition to the Technical Specifications does 
not involve a significant reduction in a margin of safety because 
the only accident mitigating equipment and/or power sources which 
will be unavailable during the transfer of offsite power sources is 
the offsite power source being removed from service, allowed by 
existing TS LCO 3.8.1.1(a). The 13.8 kV loads associated with the 
RAT being removed from service and all of the 4160 volt non-Class 1E 
loads fed from either RAT will be unavailable during this temporary 
alignment. All of these loads are nonsafety related and therefore 
are enveloped by the existing LOSP analysis.

    Based on the considerations regarding clarification of SAT Use and 
Expanded Bases, the licensee submitted the following analysis in 
accordance with 10 CFR 50.92.

    1. The proposed change to the TS does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated because only clarifications to existing TS 
action statements and an additional expanded bases are being made. 
No changes to the existing TS requirements for A.C. sources are 
being made. The safety function of the offsite power source is 
unchanged by the addition of the SAT and the probability of an LOSP 
or SBO is not increased. In actuality, the addition of the SAT 
increases the availability and flexibility of VEGP responses to 
either an LOSP or SBO.
    2. The proposed change to the TS does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated because the loss of the SAT while being utilized to meet 
TS offsite power source requirements is enveloped by existing LOSP 
analysis.
    3. The proposed change does not involve a significant reduction 
in a margin of safety because although the SAT has no 13.8 kV 
secondary winding, nor the same capacity as a RAT for accepting 4.16 
kV non Class 1E loads, these loads are nonsafety related and 
therefore enveloped by existing analysis. If a unit trip were to 
occur while one 4.16 kV safety bus is being powered from the SAT, 
the effect is a loss of the 13.8 kV and non Class 1E 4.16 kV loads 
associated with the out of service RAT. This scenario is enveloped 
by existing LOSP analysis.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia 30830.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308
    NRC Project Director: Herbert N. Berkow.

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 amendment requests: August 12, 1992 and supplemented April 
12, 1993.
    Description of amendment requests: The proposed amendments would 
change the minimum channels operable for the pressurizer safety valve 
position indicator acoustic monitor to two out of three total from one 
per valve. The amendments also delete footnotes which are no longer 
applicable.
    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:

    We [the licensee] have evaluated the proposed T/Ss exemption and 
have determined that it should not require a significant hazards 
consideration based on the criteria established in 10CFR50.92(c). 
Operation of the Cook Nuclear Plant in accordance with the proposed 
amendment will not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Although the proposed exemption results in the operator having 
one less source of information on plant status, it does not create a 
significant increase in the probability or consequences of an 
accident previously evaluated. The acoustic monitors do not perform 
a function vital to safe shutdown or to the isolation of the 
reactor, or the reactor coolant system pressure boundary, nor is 
there a mechanism involving an operable or inoperable pressurizer 
safety valve acoustic monitor which would initiate an accident. 
These monitors were added to meet the requirements of NUREG-0578 and 
NUREG-0737. During normal operations, other instrumentation exists 
that provides the operator with indication of safety valve 
actuation. The acoustic monitors are not necessary to and are not 
used in the emergency operating procedures. In addition, the 
acoustic monitors being inoperable will not result in an 
uncontrolled release of radiation to the environment and will not 
initiate an accident. Finally, although the operator may have one 
less channel operable, the operator receives no less information 
than if all three channels are operable because one valve opening 
causes all operable channels to actuate. Therefore, we conclude that 
the proposed T/Ss changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any previously analyzed.
    As previously stated, the purpose of the acoustic monitor is to 
provide the operator with information regarding safety valve 
position that may assist in the mitigation of the consequences of an 
accident. Specifically, it provides information that a safety valve 
has lifted. However, the operator has other mechanisms for obtaining 
equivalent information. In addition, the signals generated by an 
acoustic monitor do not initiate any other equipment actuation, nor 
will the inoperability of an acoustic monitor initiate any accident. 
Consequently, the proposed T/Ss changes do not create the 
possibility of a new or different kind of accident from any 
previously analyzed.
    (3) Involve a significant reduction in a margin of safety.
    The proposed T/Ss changes result in the operator potentially 
having one less source of information on plant status. However, we 
believe the margin of safety is not reduced for several reasons. 
First, the operator is provided with other viable flow detection 
devices to determine pressurizer safety valve position, i.e., the 
temperature sensor on the discharge line associated with the 
inoperable acoustic monitor, and pressurizer relief tank level (NLA-
351), temperature (NTA-351) [[Page 6303]] and pressure (NPA-351) 
indications. Also, the acoustic monitors are not used by the 
operators in an emergency situation, as the operator relies on other 
indications of loss of reactor coolant inventory per the emergency 
operating procedures. In addition, previous experience with the 
pressurizer safety valve position indicator acoustic monitoring 
system has shown that, when any one of the pressurizer safety valves 
opens, all three safety valve position indicator acoustic monitors 
are actuated. Because of this, the operator receives no less 
information regardless if only two or three channels are operable.
    Based on the above, we believe that having an acoustic monitor 
inoperable does not warrant reactor and plant shutdown. As the T/Ss 
are currently stated, should one pressurizer safety valve position 
indicator acoustic monitor become inoperable, it must be restored to 
operable status within thirty days or the unit must be in hot 
shutdown within the subsequent twelve hours. Thermal cycling from 
unwarranted plant shutdowns increases the likelihood of reactor 
vessel embrittlement and unnecessarily challenges the safety 
systems. Because a signal from the pressurizer safety valve position 
indicator acoustic monitors is not necessary nor used to ensure the 
safe shutdown of the unit even if a pressurizer safety valve is 
opened or stuck open during an emergency situation, we believe that 
a plant shutdown due to an inoperable acoustic monitor would be 
unwarranted.
    We believe that the unit can be operated safely and that we 
would still meet the intent of NUREG-0538 and NUREG-0737 with only 
two out of three pressurizer safety valve position indicator 
acoustic monitors operable.

    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 requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: November 18, 1994.
    Description of amendment request: The proposed amendment would 
change the title of certain Plant Operation Review Committee (PORC) 
members to reflect recent Maine Yankee organizational changes; update 
training requirements to comply with 10 CFR 50.120, Training and 
qualification of nuclear power plant personnel; and reporting frequency 
requirements for the Radioactive Effluent Release and Estimated Dose 
and Meteorological Summary Reports.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). A summary of the licensee's 
analysis is presented below:
    1. The proposed amendment would not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The changes proposed by this amendment request are administrative 
in nature. Because the proposed changes do not involve any physical 
alterations to plant equipment, operating setpoints, parameters or 
conditions, the plant's response to previously evaluated accidents is 
not affected.
    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.
    The administrative nature of the proposed changes does not affect 
the design, operation, maintenance or testing of the plant. Thus, no 
new modes of failure are created.
    The licensee therefore concludes that implementation of the 
proposed change will not create the possibility of 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.
    The proposed change reflects an organizational change that does not 
modify the qualification requirements or competence of the members of 
the PORC. Thus, the capability of PORC to meet its responsibilities in 
accordance with the plant Technical Specifications is unchanged.
    Deleting the current training requirement for Shift Technical 
Advisors eliminates duplicative training requirements and represents 
conformance to 10 CFR 50.120, Training and qualification of nuclear 
power plant personnel.
    Elevating the responsibility for training the plant staff from the 
Manager, Operations Department, to the Vice President of Operations, 
does not represent a reduction in a margin of safety.
    The proposed change to the Radioactive Effluent Release and 
Estimated Dose and Meteorological Summary Reports is related to the 
submittal schedule for statistical data and is administrative in 
nature. The change in submittal frequency provides consistency between 
the various required reports and also is administrative in nature.
    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.

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

    Date of amendment request: December 16, 1994.
    Description of amendment request: The proposed change to the 
Technical Specifications would require the wind direction and wind 
speed sensors at the 142 foot elevation to identify the data to 
determine action required to preclude flood damage to the Service Water 
Pumps. Also, the proposed change would correct a typographical error in 
the location of the sensors at the 374 foot elevation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    * * * The proposed changes do not involve a significant hazards 
consideration because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    NNECO [Northeast Nuclear Energy Company] is proposing to revise 
LCOs [Limiting Conditions for Operation] 3.7.5.1.b.3 and 3.7.5.1.b.4 
and Table 3.3-8 of the Millstone Unit No. 2 Technical 
[[Page 6304]] Specifications by changing the elevation that the 
average wind speed and average wind direction are measured and by 
correcting a typographical error, respectively. The proposed changes 
have no effect on any of the accidents analyzed in Chapter 14 of the 
Millstone Unit No. 2 FSAR [Final Safety Analysis Report]. Site 
flooding is considered in Section 2.5.4.2.1 of the FSAR. Utilizing 
the wind speed indicator at the 142-foot elevation, in lieu of the 
indicator on the 374-foot elevation will not significantly change 
the ability of personnel to predict the potential for a major storm 
with flooding.
    The proposed changes do not alter the intent of the 
surveillances, do not involve any physical changes to the plant, do 
not alter the way any structure, system, or component functions, and 
do not modify the manner in which the plant is operated.
    Based on the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    NNECO is proposing to revise LCOs 3.7.5.1.b.3 and 3.7.5.1.b.4 
and Table 3.3-8 of the Millstone Unit No. 2 Technical Specifications 
by changing the elevation that the average wind speed and average 
wind direction are measured and by correcting a typographical error, 
respectively. The proposed changes do not alter the intent of the 
surveillances, do not involve any physical changes to the plant, do 
not alter the way any structure, system, or component functions, and 
do not modify the manner in which the plant is operated.
    While the proposed changes to LCOs 3.7.5.1.b.3 and 3.7.5.1.b.4 
do change the measurement location stipulated by the technical 
specifications, this change is insignificant. Utilizing the wind 
speed indicator at the 142-foot elevation, in lieu of the indicator 
on the 374-foot elevation will not significantly change the ability 
of personnel to predict the potential for a major storm with 
flooding.
    Based on the above, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    NNECO is proposing to revise LCOs 3.7.5.1.b.3 and 3.7.5.1.b.4 
and Table 3.3-8 of the Millstone Unit No. 2 Technical Specifications 
by changing the elevation that the average wind speed and average 
wind direction are measured and by correcting a typographical error, 
respectively. The proposed changes will have no impact on the 
physical protective boundaries (fuel matrix/cladding, reactor 
coolant system pressure boundary, and containment). The proposed 
changes do not alter the intent of the surveillances, do not involve 
any physical changes to the plant, do not alter the way any 
structure, system, or component functions, and do not modify the 
manner in which the plant is operated.
    While the proposed changes to LCOs 3.7.5.1.b.3 and 3.7.5.1.b.4 
do change the manner in which potential flooding is predicted, this 
change is insignificant. Utilizing the wind speed and direction 
indicators at the 142-foot elevation, in lieu of the indicators at 
the 374-foot elevation will not significantly change the ability of 
personnel to predict the potential for a major storm with flooding.
    Based on the above, the proposed changes do not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local public document room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee.

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

    Date of amendment request: December 21, 1994.
    Description of amendment request: Proposed revision to License 
Condition and Technical Specifications to relocate the Fire Protection 
Requirements from the Technical Specifications to another controlled 
document, the technical requirements manual (TRM).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    * * * The proposed changes do not involve a significant hazards 
consideration because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes relocates the provisions of the Fire 
Protection Program that are contained in the Technical 
Specifications and places them in the TRM. No current requirements 
are being added or deleted aside from removal of the special reports 
section. Review of the Fire Protection Program and its revisions 
will be the responsibility of the PORC [Plant Operations Review 
Committee] and SORC [Station Operations Review Committee], just as 
it has always been the responsibility of these groups to review 
changes to the fire protection Limiting Condition for Operation and 
Surveillance Requirements when they were part of the Technical 
Specifications. In addition, no design basis accidents are affected 
by this change, nor are safety systems adversely affected by the 
changes. Therefore, there is no impact on the probability of 
occurrence or the consequences of any design basis accidents.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The proposed changes relocate the provisions of the Fire 
Protection Program that are contained in the Technical 
Specifications and places them in the TRM. No current requirements 
are being added or deleted aside from removal of the special report 
section. There are no new failure modes associated with the proposed 
changes. Since the plant will continue to operate as designed, the 
proposed changes will not modify the plant response to the point 
where it can be considered a new accident.
    3. Involve a significant reduction in a margin of safety.
    No change is being proposed for the Fire Protection Program 
requirements themselves. The relevant Technical Specifications are 
being relocated, and the requirements contained therein are being 
incorporated into the TRM. Plant procedures will continue to provide 
the specific instructions necessary for the implementation of the 
requirements, just as when the requirements resided in the Technical 
Specifications. Fire Protection Program changes will be governed by 
the provisions of 10 CFR 50.59 and the current fire protection 
license condition. As such, the changes do not directly affect any 
protective boundaries nor does it impact the safety limits for the 
boundary. Thus, there are no adverse impacts on the protective 
boundaries, safety limits, or 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: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee.

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

    Date of amendment request: December 2, 1994.
    Description of amendment request: The proposed amendment modifies 
the [[Page 6305]] surveillance requirements for the power range neutron 
flux instrumentation to permit entering reactor operating modes 1 and 2 
to perform necessary test for power range detectors.
    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 (SHC), which is presented below:

    * * * The proposed changes do not involve an SHC because the 
changes would not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously analyzed.
    NNECO is proposing to modify Table 4.3-1 by adding Note 5 to 
Functional Units 2b, 3, and 4. This note provides an exception from 
the provisions of Technical Specification 4.0.4. Entry into Mode 2 
or Mode 1, as appropriate, would allow for appropriate test 
conditions to complete the channel calibration of power range 
neutron detectors (i.e., Functional Units 2b, 3, and 4 of Table 4.3-
1). This will improve plant safety by performing tests at proper 
conditions. The acceptance criteria, such as response times, test 
frequency, or test methods, are not revised. Therefore, the power 
range neutron detectors will perform their intended function when 
called upon. Additionally, the proposed changes are consistent with 
the new, improved STS for the Westinghouse plants (NUREG-1431).
    Based on the above, the proposed changes to Functional Units 2b, 
3, and 4 of Table 4.3-1 of the Millstone Unit No. 3 Technical 
Specifications do not involve a significant increase in the 
probability or consequences of an accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed changes to Functional Units 2b, 3, and 4 of Table 
4.3-1 do not make any physical or operational changes to existing 
plant structures, systems, or components. The proposed changes do 
not introduce any new failure mode. They simply allow tests to be 
performed at appropriate conditions (e.g., Mode 2 or Mode 1 rather 
than Mode 4 or Mode 3).
    Additionally, the proposed changes do not modify the acceptance 
criteria for the tests. The purpose of the tests is to ensure that 
the power range neutron detectors can perform their intended 
function.
    Thus, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously analyzed.
    3. Involve a significant reduction in the margin of safety.
    The proposed changes to Functional Units 2b, 3, and 4 of Table 
4.3-1 do not have any adverse impact on the design basis accident 
analyses. The applicable acceptance criteria for the power range 
neutron detectors will not be modified by the proposed changes. The 
proposed changes will permit the tests to be conducted under the 
proper conditions, so that the ability of the power range neutron 
detectors to perform their intended safety function can be 
confirmed.
    Based on the above, there is no 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: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee.

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 amendment requests: August 30, 1994.
    Description of amendment requests: The proposed amendments would 
revise the Technical Specifications (TS) for Prairie Island Nuclear 
Generating Plant as recommended by Generic Letter (GL) 93-05, ``Line-
Item Technical Specification Improvements to Reduce Surveillance 
Requirements for Testing During Power Operation.'' The proposed 
amendments would also revise testing and calibration requirements 
associated with the containment hydrogen recombiners. The proposed TS 
changes are as follows:
    (1) TS Table 4.1-1C, ``Miscellaneous Instrumentation Surveillance 
Requirements.'' Delete Item 14, ``Accumulator Level and Pressure'' and 
corresponding frequency interval designations.
    (2) TS Table 4.1-2A, ``Minimum Frequencies For Equipment Tests,'' 
Item 2. Revise the frequency for partial movement of all control rod 
assemblies from every 2 weeks to once per quarter.
    (3) TS 4.3, ``Primary Coolant System Pressure Isolation Values.'' 
Under Specification heading, extend the amount of time the plant can be 
shut down before pressure isolation valve testing will be required from 
72 hours to 7 days.
    (4) TS SR 4.4.I, 4.4.I.a, 4.4.I.b, 4.4.I.b.1, 4.4.I.b.2, and 
4.4.I.b.3, ``Electrical Hydrogen Recombiners.'' Revise the containment 
hydrogen recombiner testing surveillance frequency from every 6 months 
to every refueling interval. Delete the specific requirement to perform 
CHANNEL CALIBRATION of recombiner instruments and control circuits. 
Delete the requirement to sequentially perform the resistance to ground 
test following the functional test.
    (5) TS SR 4.5.A.2.b, ``Containment Spray System.'' Revise the 
containment spray system nozzle testing surveillance frequency from 
once every 5 years to once every 10 years.
    (6) TS SR 4.8.A.1, 4.8.A.2, and Footnote, ``Auxiliary Feedwater 
System.'' Revise the testing frequency for the auxiliary feedwater 
pumps from intervals of 1 month to semi-quarterly on a staggered test 
basis.
    (7) BASES 4.8, ``Steam And Power Conversion Systems.'' Revise the 
Bases to include testing frequency for the auxiliary feedwater pumps 
from intervals of 1 month to semi-quarterly on a staggered test basis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment[s] will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Except for hydrogen recombiner changes to conform to Standard 
Technical Specifications, the requested changes were extensively 
reviewed by the NRC during the preparation of NUREG-1366 and Generic 
Letter 93-05. For the sake of clarity each proposed change is 
discussed separately in the order appearing in the Prairie Island 
Technical Specifications.
    A. This Technical Specification amendment removes the 
accumulator water level and pressure channel surveillance from the 
Technical Specifications and places them into a licensee controlled 
test procedure. These changes are consistent with industry 
recognition that accumulator instrumentation operability is not 
directly related to the capability of the accumulators to perform 
their safety function.
    Relocating the instrumentation surveillance requirements is an 
administrative change which will not affect equipment testing, 
availability, or operation. Therefore, it will not have an effect on 
the probability or consequences of an accident.
    B. This Technical Specification amendment changes control rod 
movement from every two weeks to once every quarter. Control rod 
movement testing is performed to determine if the control rods are 
immovable. Control rods may be electrically stuck due to a problem 
in the control rod drive circuitry or mechanically stuck. Electrical 
problems with the control rod drive system, in general, do not 
prevent insertion of a control rod into [[Page 6306]] the core when 
the reactor trip breakers are opened.
    NUREG-1366 determined that control rod movement testing is not 
effective in determining immovable control rods. Most of the 
mechanically immovable control rods are discovered during plant 
startup during initial pulling of the rods or during rod drop 
testing. Extending the surveillance interval will not affect this 
failure discovery method.
    The accident analyses assume that the single highest worth rod 
is struck while fully withdrawn and will not insert. One immovable 
control rod will still bound this accident analysis. For these 
reasons, the extension of the surveillance frequency from once every 
two weeks to once every quarter will not involve a significant 
increase in the probability or consequences of a previously 
evaluated accident.
    C. This Technical Specification amendment will require Reactor 
Coolant Systems Pressure Isolation Valves (PIV) to be surveillance 
tested after seven days at cold shutdown instead of after three days 
at cold shutdown.
    The PIVs are important in preventing over pressurization and 
rupture of the Emergency Core Cooling System low pressure piping 
which could result in a LOCA [loss-of-coolant accident] that 
bypasses containment. Allowable leakage from any PIV is sufficiently 
low to ensure early detection of possible in-series check valve 
failure. This change will not change the refueling outage 
surveillance, nor will it change the required testing to be 
performed after maintenance, repair, or replacement. The proposed 
level of surveillance is appropriate for these valves.
    These valves have had very good operating performance and should 
continue to have the same performance record with continuation of 
the same maintenance and testing program. Furthermore, these valves 
are backed by motor or air-operated valves which have performed 
reliably.
    For these reasons, the extension of the amount of time from 
three days to seven days before pressure isolation valve testing is 
required will not result in a significant increase in the 
probability or consequences of a previously evaluated accident.
    D. This Technical Specification amendment will revise the 
containment hydrogen recombiner testing surveillance from every six 
months to every refueling interval.
    The two independent containment hydrogen recombiners provide 
post-accident hydrogen control of the containment atmosphere. The 
recombiners are designed to be passive until an accident occurs.
    Industry experience and in particular, Prairie Island experience 
has demonstrated that this equipment is highly reliable. Since the 
recombiners are not required until after an accident, there would 
likely be time to effect accessible repairs if the equipment were 
not operable.
    Relocation of the recombiner calibration is an administrative 
change which will not affect recombiner operability. Deletion of 
specific testing sequence will not affect the performance of 
recombiner testing.
    Equipment redundancy, reliability and time for repairs ensures 
post-accident control. For these reasons, these changes will not 
result in a significant increase in the probability or consequences 
of a previously evaluated accident.
    E. This Technical Specification amendment will revise the 
containment spray system nozzle testing surveillance from once every 
five years to once every ten years.
    Two independent containment spray systems provide post-accident 
cooling of the containment atmosphere and provide a mechanism for 
removing iodine from the containment atmosphere. This surveillance 
test verifies by air flow test that the spray nozzles are 
unobstructed. The extension of the surveillance frequency does not 
affect administrative controls that preclude entry of foreign 
material into the nozzles.
    At Prairie Island the piping headers and nozzles are fabricated 
from austenitic stainless steel. There have been no reported in-
service problems noted with spray nozzle testing from plants with 
stainless steel headers and nozzles and there is no indication that 
the lines would corrode and become obstructed.
    For these reasons, this change will not result in a significant 
increase in the probability or consequences of a previously 
evaluated accident.
    F. This Technical Specification amendment will revise the 
frequency for testing the Auxiliary Feedwater Pumps (AFWP) from 
monthly to semi-quarterly on a STAGGERED TEST BASIS.
    Two 100% redundant, diverse pumps provide an emergency source of 
feedwater to the steam generators. The Prairie Island AFWPs have 
performed reliably. However, frequent testing of the pumps and 
associated equipment wears out the equipment resulting in equipment 
unavailability. AFWP availability will be increased by semi-
quarterly surveillance testing on a STAGGERED TEST BASIS.
    For these reasons, this change will not result in a significant 
increase in the probability or consequences of previously evaluated 
accident.
    Therefore, the probability or consequences of an accident 
previously evaluated are not affected by any of the proposed 
amendments.
    2. The proposed amendment[s] will not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    The extension of facility surveillance intervals as discussed 
previously will not result in changes in plant configuration or 
operation. The changes in recombiner calibration and testing will 
not result in changes in plant configuration or operation. 
Therefore, the possibility of a new or different kind of accident 
from any accident previously evaluated would not be created.
    3. The proposed amendment[s] will not involve a significant 
reduction in the margin of safety.
    The amendments proposed in this License Amendment Request do not 
reduce the ability of any system or component to perform its safety 
related function. The basis of NUREG-1366, Generic Letter 93-05, and 
the analysis performed in support of this License Amendment Request 
is that the reduction in surveillance testing can improve safety by 
reducing challenges to plant systems, personnel exposure, and 
equipment wear or degradation. The proposed changes to surveillance 
frequencies do not change the method of performing any surveillance. 
The operation of systems and equipment remains unchanged. Therefore, 
a significant reduction in the margin of safety would not be 
involved.

    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 requests involve no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: John N. Hannon.

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 amendment requests: December 5, 1994.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification 3.8 to allow containment airlock doors 
to remain open during core alterations provided certain conditions are 
met. This request is similar to the amendment for Calvert Cliffs 
Nuclear Power Plant which the NRC approved on August 30, 1994. In 
addition, these amendments would allow containment penetrations to 
remain open during core alterations provided certain conditions are 
met.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed containment refueling integrity amendments do not 
affect the probability of a fuel handling accident, they only deal 
with the containment systems.
    The containment is provided for the purpose of mitigating the 
consequences of postulated accidents. For the fuel handling accident 
in containment, the licensing basis analyses, including the NRC 
safety [[Page 6307]] evaluation report transmitted February 2, 1982, 
assumed that containment was completely abrogated and all 
radioactive materials released from the containment refueling pool 
are assumed to be released to the outside atmosphere. The requested 
amendments to Technical Specification 3.8.A.1.a modify the use of 
containment to mitigate the consequences of a fuel handling accident 
in containment, however, since instantaneous offsite release of all 
fuel handling accident materials released to containment has already 
been considered, the probability and consequences of a loss of 
containment accident are not increased.
    Therefore, the probability or consequences of an accident 
previously evaluated are not affected by any of the proposed 
amendments.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The requested amendments to Technical Specification 3.8.A.1.a 
modify the use of containment to mitigate the consequences of a fuel 
handling accident in containment. There are no new failure modes or 
mechanisms associated with the proposed changes, nor do the proposed 
changes involve any modification of plant equipment or changes in 
plant operational limits. Previous analyses, including the NRC fuel 
handling accident safety evaluation for Prairie Island, have already 
assumed the containment is abrogated. The proposed license 
amendments may affect the release path for fission products released 
during a fuel handling accident in containment, but no new or 
different kind of accident will result.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated would not be 
created.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety
    The margin of safety as defined by the licensing bases fuel 
handling accident analyses is not reduced. The previous analyses are 
very conservative, assuming all radioactive material released from 
[containment] by the fuel handling accident is immediately released 
to the outside atmosphere, and bound any changes introduced by these 
requested amendments.
    Technical Specification 3.8.A.1.a exists to minimize the 
consequences of a fuel handling accident in containment. However, 
with the current Technical Specification 3.8.A.1.a, there will still 
be releases due to the necessity to open the containment airlocks to 
evacuate personnel. With implementation of this amendment, the 
ability of the closed airlocks to contain the accident releases may 
improve.
    Some radioactive material could be released through containment 
penetrations that are open at the time of the accident. Since it is 
not likely that containment will be pressurized by a fuel handling 
accident, the releases are expected to be minimal. This amendment 
will maintain containment post-fuel handling accident offsite 
releases well within the limits of 10CFR100 and the current license 
basis releases.
    Therefore, a significant reduction in the margin of safety would 
not be involved.

    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 requests involve no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: John N. Hannon.

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 amendment requests: January 9, 1995.
    Description of amendment requests: The proposed amendments would 
revise Prairie Island Nuclear Generating Plant Technical Specification 
(TS) 4.12, ``Steam Generator Tube Surveillance,'' to incorporate 
revised acceptance criteria for steam generator tubes with degradation 
in the tubesheet roll expansion region. These criteria for steam 
generator tube acceptance were developed by Westinghouse Electric 
Corporation and are known as F* (``F-Star'') and L* (``L-Star''). These 
criteria would be utilized to avoid unnecessary plugging and sleeving 
of steam generator tubes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment[s] will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The supporting technical and safety evaluations of the subject 
criterion demonstrate that the presence of the tubesheet will 
enhance the tube integrity in the region of the hardroll by 
precluding tube deformation beyond its initial expanded outside 
diameter. The resistance to both tube rupture and tube collapse is 
strengthened by the presence of the tubesheet in that region. The 
results of hardrolling of the tube into the tubesheet is an 
interference fit between the tube and the tubesheet. Tube rupture 
cannot occur because the contact between the tube and tubesheet does 
not permit sufficient movement of tube material. The radial preload 
developed by the rolling process will secure a postulated separated 
tube end within the tubesheet during all plant conditions. In a 
similar manner, the tubesheet does not permit sufficient movement of 
tube material to permit buckling collapse of the tube during 
postulated LOCA loadings.
    The F* length of roll expansion is sufficient to preclude tube 
pullout from tube degradation located below the F* distance, 
regardless of the extent of the tube degradation. The existing 
Technical Specification leakage rate requirements and accident 
analysis assumptions remain unchanged in the unlikely event that 
significant leakage from this region does occur. As noted above, 
tube rupture and pullout is not expected for tubes using the F* 
criterion. Any leakage out of the tube from within the tubesheet at 
any elevation in the tubesheet is fully bounded by the existing 
steam generator tube rupture analysis included in the Prairie Island 
Plant USAR [Updated Safety Analysis Report]. For plants with partial 
depth roll expansion like Prairie Island, a postulated tube 
separation within the tube near the top of the roll expansion (with 
subsequent limited tube axial displacement) would not be expected to 
result in coolant release rates equal to those assumed in the USAR 
for a steam generator tube rupture event due to the limited gap 
between the tube and tubesheet. The proposed plugging criterion does 
not adversely impact any other previously evaluated design basis 
accident.
    Leakage testing of roll expanded tubes indicates that for roll 
lengths approximately equal to the F* distance, any postulated 
faulted condition primary to secondary leakage from F* tubes would 
be insignificant.
    2. The proposed amendment[s] will not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    Implementation of the proposed F* criterion does not introduce 
any significant changes to the plant design basis. Use of the 
criterion does not provide a mechanism to initiate an accident 
outside of the region of the expanded portion of the tube. Any 
hypothetical accident as a result of any tube degradation in the 
expanded portion of the tube would be bounded by the existing tube 
rupture accident analysis. Tube bundle structural integrity will be 
maintained. Tube bundle leaktightness will be maintained such that 
any postulated accident leakage from F* tubes will be negligible 
with regards to offsite doses.
    3. The proposed amendment[s] will not involve a significant 
reduction in the margin of safety.
    The use of the F* criterion has been demonstrated to maintain 
the integrity of the tube bundle commensurate with the requirements 
of Reg Guide 1.121 [``Bases for Plugging Degraded PWR Steam 
Generator Tubes''] (intended for indications in the free span of 
tubes) and the primary to secondary pressure boundary under normal 
and postulated accident conditions. Acceptable tube degradation for 
the F* criterion is any degradation indication in the tubesheet 
region, more than the F* distance below the bottom of the transition 
between the roll [[Page 6308]] expansion and the unexpanded tube. 
The safety factors used in the verification of the strength of the 
degraded tube are consistent with the safety factors in the ASME 
Boiler and Pressure Vessel Code used in steam generator design. The 
F* distance has been verified by testing to be greater than the 
length of roll expansion required to preclude both tube pullout and 
significant leakage during normal and postulated accident 
conditions. Resistance to tube pullout is based upon the primary to 
secondary pressure differential as it acts on the surface area of 
the tube, which includes the tube wall cross-section, in addition to 
the inner diameter based area of the tube. The leak testing 
acceptance criteria are based on the primary to secondary leakage 
limit in the Technical Specifications and the leakage assumptions 
used in the USAR accident analysis.
    Implementation of the tubesheet plugging criterion will decrease 
the number of tubes which must be taken out of service with tube 
plugs or repaired with sleeves. Both plugs and sleeves reduce the 
RCS (reactor coolant system) flow margin; thus, implementation of 
the F* criterion will maintain the margin of flow that would 
otherwise be reduced in the event of increased plugging or sleeving.
    Based on the above, it is concluded that the proposed change 
does not result in a significant reduction in margin with respect to 
plant safety as defined in the USAR or the Technical Specification 
Bases.

    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 requests involve no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: John N. Hannon.

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 amendment requests: January 13, 1995.
    Description of amendment requests: The proposed amendments would 
revise Prairie Island Nuclear Generating Plant Technical Specification 
4.4.D.1 to change the interval for the performance of the Residual Heat 
Removal (RHR) System leakage test from once every 12 months to perform 
the test during each refueling shutdown.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment[s] will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes to the RHR system leakage test interval 
only involve the leak-tightness of the RHR system for postaccident 
operation. As such, the proposed changes will have no impact on the 
probability of an accident previously evaluated.
    The extension of the RHR system leakage test interval could 
increase the possibility of undetected RHR system leakage outside 
the containment during post accident conditions. However, the 
possible consequences of leakage from the RHR system outside 
containment are minor relative to those of the design basis 
accident. Therefore, because leakage from the RHR system has a minor 
effect on offsite dose, and since previous testing on a 12 month 
interval has not found significant RHR system leakage, the extension 
of the test interval to refueling is not expected to significantly 
impact the offsite dose consequences of an accident. In addition, it 
is probable that RHR system leakage would be identified during the 
normal quarterly functional testing and inspection of the RHR 
system.
    Therefore, for the reasons discussed above, the proposed changes 
will not significantly affect the probability or consequences of an 
accident previously evaluated.
    2. The proposed amendment[s] will not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    There are no new failure modes or mechanisms associated with the 
proposed changes. The proposed changes do not involve any 
modification of the plant equipment or any changes in operational 
limits. The proposed changes only modify the interval for the 
performance of the RHR system leakage test. The performance of the 
RHR system leakage test on a refueling basis instead of every 12 
months cannot create a new or different kind of accident.
    Therefore, for the reasons discussed above, the proposed changes 
do not create the possibility of a new or different kind of accident 
from any previously evaluated, and the accident analyses presented 
in the Updated Safety Analysis Report [USAR] will remain bounding.
    3. The proposed amendment[s] will not involve a significant 
reduction in the margin of safety.
    The performance of the RHR system leakage test at power is more 
complex than performing the test during refueling shutdown. It is 
preferable, from an RHR system reliability and plant safety 
standpoint, to perform the test during refueling shutdown when the 
RHR system is already operating and when no changes to the RHR 
system configuration are required. Any possible increase in the risk 
to the public health and safety incurred by extending the RHR leak 
test interval from 12 months to refueling shutdown will be off-set 
by the reduction in risk obtained by not performing the RHR system 
leakage test during power operation.
    The extension of the test interval would mean that possible RHR 
leakage could exist undetected for a longer period than allowed by 
the current Technical Specifications. However, the possible 
consequences of leakage from the RHR system outside containment are 
minor relative to those of the design basis accident. In addition, 
it is probable that RHR system leakage would be identified during 
the normal quarterly functional testing and inspection of the RHR 
system.
    Based on the above, it is concluded that the proposed change 
does not result in a significant reduction in margin with respect to 
plant safety as defined in the USAR or the Technical Specification 
Bases.

    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 requests involve no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: December 23, 1994.
    Description of amendment request: The proposed amendment to the 
Technical Specifications revises the surveillance requirement to 
perform a visual inspection of containment areas affected by 
containment entry when containment integrity is established. It is 
consistent with Item 7.5 of Generic Letter 93-05, ``Line-Item Technical 
Specifications Improvements to Reduce Surveillance Requirements for 
Testing During Power Operation.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does not involve a significant increase in the probability or 
consequences of an accident previously evaluated. [[Page 6309]] 
    The proposed change does not alter the assumptions, design 
parameters or results of Updated Final Safety Analysis Report 
(UFSAR) accidents analyzed. The proposed change does not involve a 
hardware change, a change to the operation of any systems or 
components, or a change to any existing structures. The proposed 
change leads to a reduction in radiation exposure to plant personnel 
and the elimination of an unnecessary burden on plant staff. The 
revised visual inspection practice will not increase the probability 
or consequences of an accident previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed change does not modify equipment, affect system 
design bases or operability. This change does not alter parameters 
utilized in the analyzed accident scenarios. The proposed change in 
surveillance frequency is consistent with the guidance provided in 
GL 93-05. The performance of a visual inspection of containment 
areas affected by multiple containment entries on a daily bases 
[basis] and at the completion of the final entry when containment 
integrity is established will not create the possibility of a new or 
different kind of accident from those previously evaluated.
    3. Does not involve a significant reduction in a margin of 
safety.
    The proposed change only involves a decrease in surveillance 
frequency when multiple entries are made in a single day and does 
not alter the performance of the surveillance itself. System 
equipment and operation remains unchanged. Operability and 
reliability is still maintained by the required inspection. The 
adaptation of the proposed surveillance frequency does not involve a 
significant reduction in the margins 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: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: December 16, 1994 (TS 94-06).
    Description of amendment request: The proposed change would revise 
the auxiliary feedwater system technical specifications and associated 
Bases by incorporating the Westinghouse Standard Technical 
Specification limits and format, extending the limiting condition for 
operation to Mode 4, relaxing the achievement of hot shutdown from 6 
hours to 12 hours, relaxing the verification of valve position 
surveillance frequency from 7 days to 31 days, and verifying the 
position of automatic valves every 31 days in lieu of valve 
manipulation.
    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:

    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed TS change replaces SQN's auxiliary feedwater (AFW) 
system specification and the associated bases section with improved 
requirements that are modeled after the Westinghouse Standard 
(NUREG-1431) Technical Specification (STS). The proposed change is 
consistent with the STS for ensuring that three trains of AFW remain 
operable in Modes 1, 2, and 3. In addition, the proposed change 
provides a TS improvement by extending the limiting condition for 
operation (LCO) applicability to Mode 4. This LCO requirement for 
Mode 4 ensures that at least one motor-driven AFW pump remains 
operable when steam generators are being used for decay heat 
removal. The proposed 72 hour allowed outage time (for one 
inoperable train of AFW) is consistent with the STS and remains 
unchanged from SQN's current allowed outage time. One proposed 
change to relax shutdown requirements from 6 hours to 12 hours for 
achieving hot shutdown is considered to be acceptable. This 
relaxation is based on shutdown times contained in the STS and the 
operating experience to reach thus condition from full power in an 
orderly manner without challenging plant systems. The proposed 
surveillance requirements (SRs) provide test frequencies that are 
consistent with the STS and are based on operating experience and 
the design reliability of the equipment. The proposed relaxation in 
surveillance frequency from 7 days to 31 days for verifying valve 
position in the AFW flow path is considered acceptable based on 
existing procedural controls for valve configuration. The proposed 
change to include a STS SR for verifying automatic valves in the 
flow path are in their correct position every 31 days (in lieu of 
valve manipulation) is considered acceptable based on existing 
surveillance that verify proper actuation of SQN's automatic AFW 
valves.
    The proposed changes provide TS improvements for SQN's AFW 
system that ensure the system operates within the bounds of SQN's 
AFW accident analysis as contained in the Final Safety Analysis 
Report (FSAR). This change does not involve a physical modification 
to SQN's AFW system. Accordingly, the proposed changes do not 
involve an increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed TS change incorporates requirements that bound the 
limiting design-basis accidents (DBAs) evaluated in SQN's FSAR. The 
TS bases have been revised to reflect the limiting DBAs and provide 
clarification with regard to the assumptions used in SQN's AFW 
accident analysis. No new event initiator has been created, not 
[sic] has any hardware been changed. This change does not involve a 
physical change to SQN's AFW system or any other system. Therefore, 
the proposed change will not create the possibility of a new or 
different kind of accident from any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    TVA's proposed change replaces SQN's AFW system TS requirements 
with TS requirements adopted from the Westinghouse STS. Because the 
overall similarity in the requirements between SQN's current AFW 
specification and the STS version, the TS requirements remain 
essentially unchanged. The proposed 72-hour allowed outage time (for 
one inoperable train of AFW) is consistent with the STS and remains 
unchanged from SQN's current allowed outage time. One proposed 
change to relax shutdown requirements from 6 hours to 12 hours for 
achieving hot shutdown is considered to be acceptable. This 
relaxation is based on shutdown times contained in the STS and the 
operating experience to reach this condition from full power in an 
orderly manner without challenging plant systems. The proposed SRs 
provide test frequencies that are consistent with the STS and are 
based on operating experience and the design reliability of the 
equipment. The proposed relaxation in surveillance frequency from 7 
days to 31 days for verifying valve position in the AFW flow path is 
considered acceptable based on existing procedural controls for 
valve configuration. The proposed relaxation in surveillance 
frequency from 7 days to 31 days for verifying valve position in the 
AFW flow path is considered acceptable based on existing procedural 
controls for valve configuration. The proposed change to include a 
STS SR for verifying automatic valves in the flow path are in their 
correct position every 31 days (in lieu of valve manipulation) is 
considered acceptable based on other existing surveillances that 
verify proper actuation of SQN's automatic AFW valves.
    The proposed changes provide TS improvements for SQN's AFW 
System that ensure the system operates within the bounds of SQN's 
AFW accident analysis as contained in the FSAR. This change does not 
[[Page 6310]] involve a physical modification to SQN AFW system. 
Accordingly, the margin of safety has not been reduced.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: December 16, 1994.
    Description of amendment request: The proposed license amendment 
would revise Technical Specification 6.3, ``Unit Staff 
Qualifications.'' Currently, the Technical Specifications require that 
the Operations Manager obtain a senior reactor operator (SRO) license 
by August 1995. A change is proposed to relieve the requirement for the 
Operations Manager to hold a Perry Nuclear Power Plant (PNPP) SRO 
license if an Operations section middle manager holds a PNPP SRO 
license.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change affects an administrative control, which was 
based on the guidance of ANSI N18.1-1971, ``Selection and Training 
of Nuclear Power Plant Personnel.'' ANSI N18.1-1971 recommended that 
the Operations Manager hold a senior reactor operator (SRO) license. 
The current guidance in Section 4.2.2 of ANSI/ANS-3.1-1993, 
``American National Standard for Selection, Qualification, and 
Testing of Personnel for Nuclear Power Plants'' recommends, as one 
alternative, that the Operations Manager have plant operational 
knowledge consistent with the requirements of the Operations 
Manager's position, providing an Operations middle manager holds an 
SRO license. This individual (currently designated as the Operations 
Superintendent) would be required to meet the criteria for, and 
would have responsibilities as recommended in, ANSI/ANS-3.1-1993 for 
the Operations Middle Manager position. The proposed change is 
consistent with the recommendations of ANSI/ANS-3.1-1993.
    The proposed change does not alter the design of any system, 
structure or component, nor does it change the way plant systems are 
operated. It does not reduce the knowledge, qualifications, or 
skills of licensed operators, and does not affect the way the 
Operations Section is managed by the Operations Manager. The 
Operations Manager will continue to maintain the effective 
performance of section personnel and ensure the plant is operated 
safely and in accordance with the requirements of the operating 
license. Additionally, the control room operators will continue to 
be supervised by the licensed senior operators such as the Unit 
Supervisors and the Shift Supervisors. For those areas of knowledge 
that require an SRO license, the Operations Superintendent will 
provide the appropriate technical guidance to the control room 
staff.
    In summary, the proposed change does not affect the ability of 
the Operations Manager to provide the plant oversight required of 
the position. Thus, it does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change to Technical Specification 6.3.1 does not 
affect the design or function of any plant system, structure, or 
component, nor does it change the way plant systems are operated. It 
does not affect the performance of NRC licensed operators. Operation 
of the plant in conformance with the Technical Specifications and 
other license requirements will continue to be supervised by 
personnel who hold an NRC SRO license. The proposed change to 
Technical Specifications 6.3.1 ensures that either the Operations 
Manager or Operations Superintendent will be a knowledgeable and 
qualified individual by requiring one of the individuals to hold an 
SRO license for PNPP. Based on the above, the proposed change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. The proposed change does not result in a significant 
reduction in the margin of safety.
    The proposed change involves an administrative control which is 
not related to the margin of safety as defined in the Technical 
Specifications. The proposed change provides an alternative which 
ensures that the level of knowledge and experience required of an 
individual who fills the Operations Manager position is acceptable. 
The proposed change does not affect the conservative manner in which 
the plant is operated. The control room operators will continue to 
be supervised by personnel who hold an SRO license. Thus, the 
proposed change does not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Leif J. Norrholm.

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

    Date of amendment request: December 21, 1994.
    Description of amendment request: The proposed license amendment 
would revise Technical Specification 3/4.3.7.7, ``Traversing In-Core 
Probe System,'' and its Bases to allow the use of substitute data 
generated from the process computer, normalized with available 
operating measurements, to replace data from inoperable local power 
range monitor (LPRM) strings for up to 10 LPRM strings.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The TIP [traversing in-core probe] system is not used to prevent 
or mitigate the consequences of any previously analyzed accident or 
transient. No assumptions are made in any accident analysis relative 
to the operation of the TIP system. No other safety related system 
is affected by this change.
    The use of substitute values from calculations performed by the 
on-line computer core monitoring system does not affect the 
consequences of plant transients previously evaluated in the USAR 
[Updated Safety Analysis Report] because the total core TIP reading 
(nodal power) uncertainty remains less than 8.7%. Thus, the MCPR 
[minimum critical power ratio] safety limit is not affected.
    2. The proposed change does not create the possibility of a new 
or different kind of [[Page 6311]] accident from any accident 
previously evaluated.
    The proposed change does not involve the installation of any new 
equipment or the modification of any equipment designed to prevent 
or mitigate the consequences of accidents or transients. Therefore, 
the change has no effect on any accident initiator, and no new or 
different type of accidents are postulated to occur.
    3. The proposed change does not result in a significant 
reduction in the margin of safety.
    The total core TIP reading uncertainties will remain within the 
assumptions of the licensing basis; thus, the margin of safety to 
the MCPR safety limits is not reduced. The ability of the computer 
to accurately represent nodal powers in the reactor core is not 
compromised. The ability of the computer to accurately predict the 
LHGR [linear heat generation rate], APLHGR [average planar linear 
heat generation rate], MCPR, and its ability to provide for LPRM 
calibration, are not compromised. Therefore, the margin of safety is 
not significantly reduced.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Leif J. Norrholm.

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

    Date of amendment request: December 6, 1994.
    Brief description of amendments: The proposed amendment would 
revise Technical Specifications to allow appropriate remedial action 
for high particulate levels in the diesel generator fuel oil inventory 
and other out-of-limit properties in new diesel generator fuel oil that 
has been added to the existing diesel generator fuel oil storage 
inventory.
    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 allow 7 days to correct particulate 
contamination in the stored fuel oil for the diesel generators and 
30 days to confirm or restore the adequacy of the stored fuel oil if 
certain properties of new fuel that has been added to the fuel oil 
storage inventory have been discovered to exceed the specified 
values. These changes do not affect plant operations and the only 
equipment affected are the diesel generators. The ability of the 
diesel generators to provide electrical power when needed is 
directly dependent upon, in part, having fuel oil of adequate 
quality. The only accident which is potentially initiated by a 
diesel generator failure is the station blackout event. The 
mitigation of many accidents is dependent upon the availability of 
at least one train of electrical power from an emergency diesel 
generator (EDG). With the proposed changes, the fuel oil should 
continue to have sufficient quality to assure the operability of the 
diesel generators until the particulate and other properties are 
returned to within limits. This is due in part to the existing fuel 
oil quality requirements that are more stringent than the vendor 
requires for the EDG to operate and the system of filters installed 
to insure good quality fuel actually reaches the EDG. Even though 
the margin provided in the quality of the fuel oil may be affected 
(see the response to question 3 below), adequate fuel oil quality is 
being maintained to assure the operability of the diesel generators 
and therefore, these changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no hardware changes and no changes in system 
operations involved. These changes only affect the quality of the 
stored fuel oil for the diesel generators. The availability of a 
diesel generator has been addressed by the CPSES [Comanche Peak 
Steam Electric Station] design and in particular by the analysis of 
the station blackout event. These changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The margin of safety of interest for these changes is the 
quality of the stored fuel oil for the diesel generators as compared 
to minimum quality which will support the diesel generators ability 
to supply electrical power when needed. Particulate contamination 
increases slowly over a period of time due to the chemical breakdown 
of the fuel oil (or its additives or the surfaces on the tanks 
themselves) or due to the introduction of foreign material during 
refueling activities. When considered with the fact that the 
existing limitation of 10 mg/L was developed for engines which 
require much cleaner fuel oil (aircraft engines) and that the CPSES 
diesel engines have in line duplex fuel oil filters which can be 
switched while the engine is operating, the 7 days which are being 
provided to restore the particulate levels do not involve a 
significant reduction in the margin of safety. The levels of 
particulate are expected to not exceed the specified value by a 
significant amount and the specified value is already quite 
conservative. Seven days is a reasonable time period in which to 
restore the parameter but is short enough to ensure that the 
contamination values do not exceed the vendors recommended fuel oil 
tolerances required for the EDGs to run. In a similar manner, the 
properties of the new fuel oil that has been added to the fuel oil 
storage inventory are not expected to deviate significantly from the 
allowed values. The testing for gravity, viscosity, flash point, 
clarity, water and sediment prior to adding the new fuel oil 
provides adequate assurance that the stored fuel oil will be of 
sufficient quality to support diesel generator operation. The 
quality of the stored fuel oil is further protected from problems 
being introduced by new fuel oil that has been added to the fuel oil 
storage inventory by the fact that the new fuel oil is generally 
diluted by a factor of four or more when it is added to the storage 
tanks by the fuel oil that is already in the tanks. Allowing 30 days 
to confirm or restore the properties of the stored fuel oil when a 
sample of new fuel that has been added to the fuel oil storage 
inventory has properties which exceed their specified values does 
not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.
    Attorney for licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: December 7, 1994.
    Brief description of amendments: The proposed amendment to the 
technical specifications (TSs) would: (1) revise the Comanche Peak 
Steam Electric Station (CPSES), Technical Specification Limiting 
Condition for Operation (LCO) for the main steam isolation valves 
(MSIVs) to increase the allowed outage time (AOT) in Mode 1; (2) 
relocate the MSIVs full closure time requirement to a program 
administratively controlled by the TS; and (3) revise the associated 
Bases to [[Page 6312]] adopt the expanded Bases format adding 
information specific to CPSES.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes are to (1) revise the CPSES Technical 
Specification Limiting Condition for Operation (LCO) for the MSIVs 
to increase the Allowed Outage Time (AOT) from 4 hours to 8 hours in 
Mode 1; (2) modify the Mode 2 and 3 Action Statement to better 
reflect the safety significance of these valves by requiring that 
the valves be closed within 8 hours and verified at least every 7 
days; (3) relocate the MSIVs full closure time requirement to a 
program administratively controlled by the TS; and (4) revise the 
associated Bases to adopt the expanded Bases format adding 
information specific to CPSES.
    The revision of the CPSES Technical Specification Limiting 
Condition For Operation (LCO) for the MSIVs to increase the Allowed 
Outage Time (AOT) from 4 hours to 8 hours in Mode 1 only affects the 
time that a condition can exist and as such does not affect any of 
the conditions that could initiate an accident; therefore the 
probability of an accident is not affected. Likewise, no new 
conditions are created that would affect the analyses of any 
accident; therefore the consequences of the accidents postulated for 
CPSES are not affected.
    Modifying the Mode 2 and 3 Action Statement to better reflect 
the safety significance of these valves by requiring that the valves 
be closed within 8 hours and verified at least every 7 days provides 
clarity and adds a new verification requirement. Again no new plant 
conditions are established, time limits and verification 
requirements are merely being established; therefore, no accident 
initiators are affected and there is no impact on the probability of 
any accident. Likewise no conditions are being altered which affect 
the analyses of any accidents which are postulated at CPSES and thus 
the consequences of those accidents are unaffected.
    Relocating the MSIVs full closure time requirement to a program 
administratively controlled by the TS is an administrative change 
only. It has no impact on actual plant operation and thus there is 
no impact on the probability of any accident or on the consequences 
of any accident.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated? 
None of the changes in this request affect plant design or create 
new operating configurations. The only things affected are the times 
that certain conditions are allowed, how soon actions need be 
performed, how often to verify conditions and the administrative 
location of certain requirements. These items do not create the 
possibility of a new type or different kind of accident.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The Technical Specifications LCOs ensure that the assumptions of 
the safety analyses are preserved. There are no substantive changes 
to the LCO; therefore, the safety analyses are unaffected and there 
is no affect on the margin of safety.
    Revising the CPSES Technical Specification Limiting Condition 
For Operation (LCO) for the MSIVs to increase the Allowed Outage 
Time (AOT) from 4 hours to 8 hours in Mode 1 allows the unit to 
operate with an inoperable MSIV for a longer period of time. 
Although the unavailability of equipment required to mitigate or 
assess the consequence of an accident is increased, a more 
reasonable completion time is provided to diagnose the problem, 
mobilize the corrective action, obtain administrative clearances, 
complete the maintenance, restore the valve to an operable 
condition, and perform post-maintenance verification, where 
appropriate. The additional time would reduce the probability of 
unnecessary plant transients and plant shutdowns, thus improving 
plant safety and increasing plant availability, while a qualitative 
assessment has concluded that the impact on Core Damage Frequency is 
negligible. TU Electric has concluded based on the discussion above 
that there is no significant impact on the overall margin of safety 
due to this change.
    Modifying the Mode 2 and 3 Action Statement to better reflect 
the safety significance of these valves by requiring that the valves 
be closed within 8 hours and verified at least every 7 days is 
primarily a clarification and a new verification requirement. 
Specifying that an inoperable valve be closed within 8 hours makes 
the requirement specific where no time limit was provided before. 
The 8 hours specified is the same as is allowed in Mode 1 which was 
qualitatively assessed as noted above and thus is a logical 
limitation. The new requirement to verify the valves closed on a 
periodic basis will increase assurance that the valves remain closed 
and will thus enhance the margin of safety. Overall, TU Electric 
concludes that these Mode 2 and 3 changes do not significantly 
affect the margin of safety.
    Relocating the MSIVs full closure time requirement to a program 
administratively controlled by the TS is an administrative change 
only. There is no impact on the margin of safety.
    Revising the associated Bases to adopt the expanded Bases format 
adding information specific to CPSES enhances the useability of the 
Technical Specification. Overall, this is considered an improvement 
which will benefit both the operators and support personnel. There 
is no significant impact on the margin of safety and if there is an 
impact, it improves the margin by providing easy access to support 
information.

    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 Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.
    Attorney for licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: December 19, 1994.
    Brief description of amendments: The proposed changes to the 
Technical Specification Action Statements of Tables 3.3-1 and 3.3-2 
would allow testing of the reactor protective system (RPS) and the 
engineered safety features actuation system (ESFAS) with the channel 
under test in bypass.
    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) Do the proposed changes involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed changes will revise those Action Statements which 
limit the use of bypass while testing for Reactor Protection System 
(RPS) and Engineered Safety Feature Actuation System (ESFAS) 
functions. The Actions Statements concern testing with a channel 
inoperable and will be revised to allow testing with either the 
inoperable channel or the channel being tested (but not both) placed 
in bypass.
    Testing in a bypass condition when all channels are operable 
will not introduce new operating configurations. The number [of] 
available channels with one channel in bypass for testing will 
remain the same as the minimum number of channels and is the same as 
the number of channels available when testing in trip. The number of 
channels to trip will be unchanged when testing in bypass while the 
number of channels to trip is reduced to one when testing in trip. 
Although there may be a sight [slight] increase in possibility that 
the failure of a channel could prevent the actuation of a function 
(because testing in bypass could result in two-out-of-two logic 
while testing in trip would have resulted in one-out-of-two logic), 
testing in bypass will reduce the vulnerability to inadvertent 
actuation of a function while maintaining the normal channels to 
trip and the minimum channels [[Page 6313]] operable requirements 
per the current technical specifications. Overall TU Electric 
concludes (and WCAP-10271 with its associate SER from the NRC 
supports) that testing in bypass when all channel [s] are operable 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Testing in bypass with one channel inoperable will not introduce 
new configurations. The current Actions Statements for ESFAS already 
allow testing in bypass if one channel is inoperable. Under the 
current Technical Specifications for an RPS function, an inoperable 
channel is placed in bypass (via leads and jumpers) while 
surveillance testing another channel (the channel under test is 
placed in trip). Under the proposed changes, either the inoperable 
channel or the channel being tested may be bypassed.
    In either case, the result is one channel in bypass and the 
other in trip, which leaves one-out-of-two operable channels to 
initiate the protective function (if the initial logic was two-out-
of-four) or one-out-of-one operable channels to initiate the 
protective function (if the initial logic was two-out-of-three). 
Thus, testing in bypass with one channel inoperable does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed technical specification changes will also allow 
certain ESFAS functions to be tested with an inoperable channel in 
bypass and the channel being tested in trip. The current technical 
specifications require that the inoperable channel be in trip and 
that the channel being tested be in bypass. Per the same logic 
provided above on testing in bypass with an inoperable channel, this 
change has no impact on the capability of the system to respond to 
plant conditions and does increase the potential for inadvertent 
actuation of a function.
    In summary, the proposed changes to the technical specifications 
and testing in bypass do not increase the probability or 
consequences of an accident previously evaluated.
    (2) Do the proposed changes create the possibility of a new or 
different type of accident from any accident previously evaluated?
    No new operating configurations and no new failure modes are 
being introduced by testing in bypass or by the proposed technical 
specification changes; therefore, no new or different type of 
accident from any accident previously evaluated is being created.
    (3) Do the proposed changes involve a significant reduction in 
the margin of safety?
    Testing in bypass does not affect accident configurations, 
sequences, or response scenarios as modeled in the safety analyses. 
Testing or maintenance in a bypass configuration does not cause any 
design or analysis acceptance criteria to be exceeded, nor does it 
affect the integrity of the fission product barriers. The severity 
of any accident previously evaluated is not increased. Bypass 
testing does not affect the functional integrity of the Reactor 
Protection System (RPS) or the Engineered Safety Features Actuation 
System (ESFAS). Bypass testing and the proposed technical 
specification changes do not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.
    Attorney for licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: December 30, 1994
    Brief description of amendments: The proposed amendments would 
revise the technical specification for fuel storage to authorize use of 
the high density fuel storage racks, to increase the spent fuel storage 
capacity, and to adopt the wording, content, and format of the Improved 
Standard Technical Specifications.
    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. Do the proposed changes involve a significant increase in the 
probability or consequence of an accident previously evaluated?
    This proposed license amendment includes changes which clarify 
the Technical Specifications, identify existing licensing basis 
criteria, revise the wording and format to be consistent with the 
Improved Standard Technical Specifications (NUREG-1431), and provide 
the criteria for acceptable fuel storage in high density racks. The 
clarification and the revised wording and format are purely 
administrative changes and have no impact on the probability or 
consequences of an accident. The criteria for acceptable fuel 
storage in the high density racks are discussed below.
    The high density racks differ from the low density racks in that 
the center to center storage cell spacing is decreased from a 
nominal 16 inches to a nominal 9 inches and the high density racks 
are free standing whereas the low density racks are bolted to the 
pool. The allowed storage pattern in the high density racks results 
in a nominal 12.7 inch center to center spacing (measured 
diagonally) with a two out of four storage pattern (high density (2/
4)). Administrative controls are used to maintain the specified 
storage patterns and to assure storage of a fuel assembly in a 
proper location based on initial U-235 enrichment and burnup. The 
increased storage capacity results in added weight in the pools and 
additional heat loads.
    The only potential impact on the probability of an accident 
concerns the potential insertion of a fuel assembly in an incorrect 
location in the high density racks. TU Electric has used 
administrative controls to move fuel assemblies from location to 
location since the initial receipt of fuel on site. Through receipt 
of fuel for two initial core loads and four refueling outages (each 
of which includes a complete core offload), TU Electric has not 
inserted a fuel assembly into an improper location. This record 
demonstrates the adequacy of the administrative controls in place 
and confirms that the use of such administrative controls will not 
involve a significant increase in the probability of an accident 
previously evaluated.
    The consequences of all of these changes have been assessed and 
the current acceptance criteria in the licensing basis of CPSES will 
continue to be met. The nuclear criticality, thermal-hydraulic, 
mechanical, material and structural designs will accommodate these 
changes. Potentially affected analyses, including a dropped spent 
fuel assembly, a loss of spent fuel pool cooling, a seismic event, 
and a fuel assembly placed in a location other than a prescribed 
location, continue to satisfy the CPSES licensing basis acceptance 
criteria. The analysis methods used by TU Electric are consistent 
with methods used by TU Electric in the past or methods used 
elsewhere in the industry and accepted by the NRC.
    Based on the acceptability of the methodology used and 
compliance with the current CPSES licensing basis, TU Electric 
concludes that the use of the high density racks and the increase in 
storage capacity do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The administrative changes to the Technical Specifications have 
no impact on plant hardware or operations and therefore cannot 
create a new or different kind of an accident.
    The spacing changes between fuel assemblies, the administrative 
controls, the storage limitations, and the increased storage 
capacity do not generate new failure modes that could create a new 
or different kind of an accident. The change from bolted low density 
racks to free standing high density racks will not create the 
possibility of a new or different kind of an accident. Free standing 
racks have been commonly used at nuclear power plants to provide for 
high density storage of spent fuel, and their use [[Page 6314]] does 
not entail any unproven or unusual design or technology. In this 
regard, a number of plants have previously changed from bolted or 
restrained racks to free standing racks, including Millstone 1 
(amendment dated November 27, 1989) and San Onofre 2 and 3 
(amendment dated May 1, 1990), and such changes have not been 
classified as involving a significant hazards consideration. 
Furthermore, CPSES is not located in an area subject to severe 
seismic events. A seismic event at CPSES would result in little 
movement of the free standing racks and would not cause the high 
density racks to collide with each other or the spent fuel pool 
walls. Therefore, use of the free standing high density racks would 
not create the possibility of a new or different kind of an 
accident.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The proposed administrative changes to the Technical 
Specifications have no impact on any acceptance criteria, plant 
operations or the actual failure of any systems, components or 
structure; therefore these administrative changes have no impact on 
the margin of safety.
    The NRC guidance [Nuclear Regulatory Commission, Letter to all 
Power Reactor Licensees, from B. K. Grimes, April 14, 1978, ``OT 
Position for Review and Acceptance of Spent Fuel Storage and 
Handling Applications,'' as amended by the NRC Letter dated January 
18, 1979] has established that an evaluation of margin of safety 
should address the following areas:
    (1) Nuclear criticality considerations.
    (2) Thermal-Hydraulic considerations.
    (3) Mechanical, material and structural consideration.
    The established acceptance criterion for criticality is that the 
neutron multiplication factor in the spent fuel pool storage racks 
shall be less than or equal to 0.95, including uncertainties, under 
all conditions. The keff for the high density racks for CPSES 
is always less than 0.95, including uncertainties at a 95/95 
probability confidence level. Because the existing acceptance 
criterion is shown to be satisfied, the high density racks do not 
involve a significant reduction in the margin of safety with respect 
to criticality considerations.
    The thermal-hydraulic evaluation demonstrates that the 
temperature margin of safety will be maintained. Re-evaluation of 
the spent fuel pool cooling system for the increased heat loads 
shows, with minor modifications, that the spent fuel cooling system 
will maintain the abnormal maximum temperature of the spent fuel 
pool water within the limits of the existing licensing basis (i.e., 
below 212  deg.F). Additionally, it shows that, with minor 
modifications, the normal maximum temperature will be within the 
existing design basis temperatures for the high density racks, 
liner, structure, and cooling system and will not have any 
significant impact on the spent fuel pool demineralizers. Thus, the 
existing licensing basis remains valid, and there is no significant 
reduction in the margin of safety for the thermal-hydraulic design 
or spent fuel cooling.
    The main safety function of the spent fuel pool and the high 
density racks is to maintain the spent fuel assemblies in a safe 
configuration through normal and abnormal operating conditions. The 
design basis floor responses of the Fuel Building were confirmed to 
be adequate and conservative and the floor loading will not exceed 
the capacity of the Fuel Building. The high density rack materials 
used are compatible with the spent fuel pool and the spent fuel 
assemblies. The structural considerations of the high density racks 
maintain margin of safety against tilting and deflection or 
movement, such that the high density racks do not impact each other 
or the pool walls, damage spent fuel assemblies, or cause 
criticality concerns. Thus, the margin of safety with respect to 
mechanical, material and structural considerations are not 
significantly reduced by the use of the high density racks.

    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 Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.
    Attorney for licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, DC 20036.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: December 9, 1994.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 4.3.2.2, 4.7.1.2.1, and the Bases 
for Specification 3/4.7.1.2. The changes would decrease the frequency 
of testing auxiliary feedwater pumps, provide consistent testing 
requirements for the steam turbine-driven auxiliary feedwater pump, and 
clarify performance parameters in the Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed revision does not involve a significant hazards 
consideration because operation of Callaway Plant with this change 
would not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The Callaway Final Safety Analysis Report has been reviewed and 
been found to be unaffected by these proposed changes. The changes 
proposed by this Technical Specification amendment do not affect the 
performance parameters of the Auxiliary Feedwater System (AFWS). The 
changes proposed involve a decrease in the frequency of pump testing 
from once per 31 days to once per 92 days as recommended by NRC 
Generic Letter 93-05 and reflected in NUREG-1431 (T/S 4.7.1.2.1.a). 
This change will decrease the out-of-service time of the AFWS due to 
testing. This change will also decrease the number of component 
manipulations performed on the system and will therefore decrease 
the probability of a restoration error rendering the system 
incapable of performing its intended function.
    The pumps will be required to meet the same acceptance criteria 
and will continue to be monitored as required by ASME Section XI. As 
stated earlier, the overall effect is a slight decrease in the CDF 
for Callaway. These proposed changes will also eliminate an 
inconsistency among Specifications 4.7.1.2.1.b.2 and 4.3.2.2 and 
Specification 4.7.1.2.1.a.2 regarding an exception to Specification 
4.0.4 for entry into Mode 3 for the TDAFP. The methodology and 
acceptance criteria of surveillance testing will not be changed. The 
ability of the AFWS to perform its intended function during accident 
conditions will continue to be demonstrated via surveillance 
testing. The proposed changes to the Technical Specifications do not 
affect any accident initiators for any accident evaluated in the 
Final Safety Analysis Report (FSAR). The Bases changes are 
corrections to errors which have no effect on any accident 
initiators nor equipment failure modes.
    (2) Create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed Technical Specification changes do not modify any 
equipment nor create any potential accident initiators. The proposed 
change herein of potential interest is the exception to 
Specification 4.0.4 for entry into Mode 3 for TDAFP response time 
testing and auto-start testing. This allowance is already recognized 
via Specification 4.7.1.2.1.a.2 and NUREG-1431, Standard Technical 
Specifications-Westinghouse Plants.
    (3) Involve a significant reduction in a margin of safety.
    The Bases for Specification 3/4.7.1.2 are to be clarified to 
correctly state the design flow and pressure parameters for the 
AFWS. No plant design changes are involved in any of the proposed 
changes and the method and manner of plant operation remain the 
same. The specific surveillance test methodology and acceptance 
criteria remain unchanged.
    As discussed above, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated or create the possibility of a new or 
different kind of accident from any previously evaluated. These 
changes do not result in a significant reduction in a margin of 
safety. Therefore, it has been determined that the proposed changes 
do not involve a significant hazards consideration.

     [[Page 6315]] 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: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
    NRC Project Director: Leif J. Norrholm.

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

    Date of amendment request: December 9, 1994, as supplemented on 
December 22, 1994.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Surveillance Requirement 
4.8.1.1.2f.7 to remove the requirement to perform the hot restart test 
within 5 minutes of completing the 24-hour endurance test and place 
that requirement in a separate TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed revision does not involve a significant hazards 
consideration because operation of Callaway Plant with this change 
would not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed revision to the T/S will not adversely impact plant 
safety since the requirement to perform the hot restart test will 
still be implemented via a separate surveillance requirement that 
demonstrates the hot restart functional capability of the diesel 
generators.
    (2) Create the possibility of a new or different kind of 
accident from any previously evaluated.
    There are no design changes being made that would create a new 
type of accident or malfunction and the method and manner of plant 
operation remain unchanged. The performance capability of the 
emergency diesel generators will not be affected. The verification 
of the hot restart capability of the diesel generators will still be 
performed, only the timing of the performance will be changed to 
give plant operators added flexibility and prevent critical path 
complications during outages.
    (3) Involve a significant reduction in a margin of safety.
    There are no changes being made to the safety limits or safety 
system settings that would adversely impact plant safety. The diesel 
generators will still perform their intended safety function 
following a loss of offsite power, to achieve and maintain the plant 
in a safe shutdown condition.
    Based on the above discussions, it has been determined that the 
requested Technical Specification change does not involve a 
significant increase in the probability or consequences of an 
accident or create the possibility of a new or different kind of 
accident or condition over previous evaluations; or involve a 
significant reduction in a margin of safety. Therefore, the 
requested license amendment does not involve a significant hazards 
consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Leif J. Norrholm.

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

    Date of amendment request: December 14, 1994.
    Description of amendment request: The proposed amendment would 
revise instrument identification for low reactor pressure instrument 
trip cards in emergency core cooling system (ECCS) actuation to reflect 
a design change to be installed during the 1995 refueling outage.
    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 to the identification numbers for certain 
reactor pressure instrumentation as included in the Technical 
Specifications for ECCS Actuation Instrumentation is only necessary 
because the specific identification numbers (Tag Nos.) have been 
listed in the [***]. This is considered an administrative type 
change. Acceptable measurement of Low Reactor Pressure is still 
assured. All automatic control or trip functions will continue to be 
provided.
    The proposed change does not result in any function or setpoint 
change. The hardware changes which have resulted in a need to change 
the Technical Specifications have removed instrumentation no longer 
required to be installed in the circuitry for measuring ECCS Low 
Reactor Pressure. The existing logic for Low Reactor Pressure will 
remain the same. The only change applicable to implementation of the 
design modification is the use of different trip cards to provide 
the trip function for ECCS Low Reactor Pressure.
    The requested change to ECCS Actuation Instrumentation Tables 
does not impact any FSAR [Final Safety Analysis Report] safety 
analysis involving the ECCS or Protection Systems. These measurement 
functions are not contributors to the initiation of accidents.
    The change in instrument Tag Nos. on Tables 3.2.1 and 4.2.1 will 
have no affect on any safety limit setting or plant system operation 
and, therefore, does not modify or add any initiating parameters 
that would significantly increase the probability or consequences of 
any previously analyzed accident.
    The administrative change to correct a typographical error on 
Table 4.2.1 will have no affect on plant hardware, plant design, 
safety limit setting or plant system operation and, therefore, does 
not modify or add any initiating parameters that would significantly 
increase the probability or consequences of any previously analyzed 
accident.
    Therefore, it is concluded that there is not a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. The proposal to change instrument Tag Nos. does not result in 
any function changes or changes to Technical Specification 
requirements pertaining to these functions.
    The proposed change does not involve any change in Technical 
Specification trip setpoints, plant operation, redundancy, 
protective function or design basis of the plant. There is no impact 
on any existing safety analysis or safety design limits. Low Reactor 
Pressure instrumentation functions do not initiate nuclear system 
parameter variations which are considered potential initiating 
causes of threats to the fuel and the nuclear system process barrier 
or that would create any new or different kind of accident.
    As discussed above, the proposed administrative change only 
corrects a typographical error concerning equipment identification 
numbers. This change does not affect any equipment and it does not 
involve any potential initiating events that would create any new or 
different kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposal to change the identification numbers for certain 
reactor pressure instrumentation as included in the Technical 
Specifications for ECCS Actuation Instrumentation does not affect 
any existing safety margins. The change by itself is administrative. 
The hardware changes which have resulted in a need to change the 
Technical Specifications have been reviewed per 10 CFR 50.59(a)(2) 
and determined to not constitute an unreviewed safety question.
    The change in Tag Nos. or the change in the instrumentation used 
to measure low [[Page 6316]] reactor pressure does not preclude the 
ability of the Core Spray (CS) or Low Pressure Coolant Injection 
(LPCI) Systems to perform their safety function to mitigate the 
consequences of accidents or of any other safety system to 
accomplish its safety functions. Proper post-accident ECCS 
functioning will still be provided by safety class instruments used 
to measure reactor pressure.
    The change to instrument Tag Nos. as listed in the Technical 
Specifications has no affect on the bases of Protective 
Instrumentation which is to operate to initiate required system 
protective actions. The changes to be implemented which have 
resulted in a need to change the Technical Specifications will 
actually improve the accuracy of reactor pressure measuring loops.
    [***]
    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: Brooks Memorial Library, 224 
Main Street, Brattleboro, Vermont 05301.
    Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray, 
One International Place, Boston, Massachusetts 02110-2624.
    NRC Project Director: Walter R. Butler.

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

    Date of amendment request: December 22, 1994.
    Description of amendment request: The proposed amendment would 
modify Point Beach Nuclear Plant Technical Specification (TS) Section 
15.3.3, ``Emergency Core Cooling System, Auxiliary Cooling Systems, Air 
Recirculation Fan Coolers, and Containment Spray,'' TS Section 15.3.4, 
``Steam and Power Conversion System,'' TS Section 15.3.5, 
``Instrumentation System,'' TS Section 15.3.7, ``Auxiliary Electrical 
Systems,'' TS Section 15.3.14, ``Fire Protection System,'' and TS 
Section 15.4.1, ``Operation Safety Review.'' The modifications would 
delete obsolete TSs, would provide spring 1995 outage-specific TSs as 
part of the ongoing diesel upgrade project, would update several TSs to 
be consistent with the upgrade project design changes, and would change 
one monthly testing requirement. In addition, the bases for Section TS 
15.3.7 would be modified to be consistent with the proposed TS changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    In accordance with the requirements of 10 CFR 50.91(a), 
Wisconsin Electric Power Company (Licensee) has evaluated the 
proposed changes against the standards of 10 CFR 50.92 and has 
determined that the operation of Point Beach Nuclear Plant, Units 1 
and 2, in accordance with the proposed amendments [sic] does not 
present a significant hazards consideration. The analysis of the 
requirements of 10 CFR 50.92 and the basis for this conclusion are 
as follows:
    1. Operation of the facility under the proposed Technical 
Specifications will not create a significant increase in the 
probability or consequences of an accident previously evaluated.
    The probabilities of accidents previously evaluated are based on 
the probability of initiating events for these accidents. Initiating 
events for accidents previously evaluated for Point Beach include: 
control rod withdrawal and drop, CVCS malfunction (Boron Dilution), 
startup of an inactive reactor coolant loop, reduction in feedwater 
enthalpy, excessive load increase, losses of reactor coolant flow, 
loss of external electrical load, loss of normal feedwater, loss of 
all AC power to the auxiliaries, turbine overspeed, fuel handling 
accidents, accidental releases of waste liquid or gas, steam 
generator tube rupture, steam pipe rupture, control rod ejection, 
and primary coolant system ruptures.
    This license amendment request proposes to remove the 
specifications associated with the 4160 volt safeguards bus tie, add 
and modify specifications associated with the degraded and loss of 
voltage protection functions, and remove specifications and 
surveillance exceptions that are obsolete. The modifications being 
performed and the changes proposed by this license amendment request 
have been reviewed and we conclude that these changes do not 
increase the probability of any initiating event for accidents 
previously analyzed for Point Beach Nuclear Plant.
    The consequences of the accidents previously evaluated in the 
PBNP FSAR are determined by the results of analyses that are based 
on initial conditions of the plant, the type of accident, transient 
response of the plant, and the operation and failure of equipment 
and systems. The changes proposed in this license amendment request 
provide appropriate limiting conditions for operation, action 
statements, allowable outage times, surveillances and bases for the 
Point Beach Nuclear Plant Technical Specifications.
    The proposed specification that allows a Train A service water 
pump powered from the alternate shutdown system to be considered 
operable under the provisions of Technical Specification 15.3.0.c is 
appropriate to maintain operability of the service water system for 
the continued safe operation of Unit 2 under the applicable standby 
emergency power limiting condition for operation.
    The modifications that are being performed have been designed 
and will be installed in accordance with the applicable design and 
installation requirements for Point Beach Nuclear Plant.
    Therefore, this proposed license amendment does not affect the 
consequences of any accident previously evaluated in the Point Beach 
Nuclear Plant FSAR because the factors that are used to determine 
the consequences of accidents are not being changed.
    2. Operation of this facility under the proposed Technical 
Specifications change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    New or different kinds of accidents can only be created by new 
or different accident initiators or sequences. New and different 
types of accidents (different from those that were originally 
analyzed for Point Beach) have been evaluated and incorporated into 
the licensing basis for Point Beach Nuclear Plant. Examples of 
different accidents that have been incorporated into the Point Beach 
Licensing basis include anticipated transients without scram and 
station blackout.
    The modifications being performed and the changes proposed by 
this license amendment request have been reviewed and we conclude 
that these changes do not create any new or different accident 
initiators or sequences. Therefore, these modifications and proposed 
Technical Specification changes do not create the possibility of an 
accident of a different type than any previously evaluated in the 
Point Beach FSAR.
    3. Operation of this facility under the proposed Technical 
Specifications change will not create a significant reduction in a 
margin of safety.
    The margins of safety for Point Beach are based on the design 
and operation of the reactor and containment and the safety systems 
that provide their protection. The modifications that are being 
performed have been designed and will be installed in accordance 
with the applicable design and installation requirements for Point 
Beach Nuclear Plant.
    The modification to change the loss of voltage protection 
function from 1-out-of-2 logic on each bus to 2-out-of-3 logic on 
each bus is an improvement over the original design, because with 
the new design an inadvertent trip of a single channel will not 
cause the protection actions. Also, when any single channel is taken 
out-of-service for testing, maintenance, or calibration it can be 
placed in the trip condition to allow actuation of the protection 
function by the trip of either of the remaining operable channels.
    The Technical Specification change to allow an operating pump 
powered from alternate shutdown to be considered operable is 
justified because the pump is able to perform its safety function 
powered from the alternate shutdown power source. The alternate 
shutdown system is powered via offsite power or from the onsite gas 
turbine generator and is being considered a normal power supply for 
the service water pump.
    The alternate shutdown system was installed to provide an 
alternate means of [[Page 6317]] providing power to service water 
pumps, component cooling water pumps, and residual heat removal 
pumps for certain 10 CFR 50 Appendix R fire scenarios in which the 
normal power supplies for this equipment become inoperable. As such, 
the alternate shutdown system is a qualified alternate source of 
power for the service water pump.
    Therefore, the margins of safety for Point Beach are not being 
reduced because the design and operation of the reactor and 
containment are not being changed and the safety systems that 
provide their protection that are being changed are being modified 
in accordance with the applicable design and installation 
requirements for the Point Beach Nuclear Plant.

    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: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Leif J. Norrholm.

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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of application for amendments: October 31, 1994, supplemented 
by letter dated December 28, 1994.
    Brief description of amendment requests: The proposed amendments 
would change the refueling machine overload cutoff limit from less than 
or equal to 1556 pounds to less than or equal to 1600 pounds. The 
change is a consequence of the fuel assembly weight increase which 
resulted from design and fabrication improvements.
    Date of individual notice in Federal Register: January 6, 1995 (60 
FR 2160).
    Expiration date of individual notice: February 6, 1995.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.

Entergy Operations, Inc., Docket No. 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 revises 
technical specifications to address the installation of two battery 
chargers on each vital 125 vdc power train in lieu of the ``swing'' 
battery charger that is currently used.
    Date of individual notice in the Federal Register: January 17, 1995 
(60 FR 3439).
    Expiration date of individual notice: February 16, 1995.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 728011.

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 amendment request: October 21, 1994.
    Brief description of amendment request: The proposed amendment 
would add the Special Test Exception 3/4.10.6, ``Inservice Leak and 
Hydrostatic Testing,'' that allows the performance of pressure testing 
at reactor coolant temperature up to 212  deg.F while remaining in 
OPERATIONAL CONDITION 4. This special test exception would also require 
that certain OPERATIONAL CONDITION 3 Specifications for Secondary 
Containment Isolation, Secondary Containment Integrity and Standby Gas 
Treatment System operability be met. This change would also revise the 
Index, Table 1.2, ``OPERATIONAL CONDITIONS,'' and the Bases to 
incorporate the reference to the proposed special test exception.
    Date of publication of individual notice in Federal Register: 
December 22, 1994 (59 FR 66057).
    Expiration date of individual notice: January 23, 1995.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.

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

    Date of application for amendment: December 8, 1994.
    Brief description of amendment: The proposed amendment would revise 
Section 4.4 of the Indian Point 3 Technical Specifications. 
Specifically, TS 4.4.E.1 would be revised to allow a one-time extension 
to the 30-month interval requirement for leak rate testing of Residual 
Heat Removal (RHR) containment isolation valves AC-732, AC-741, AC-MOV-
743, AC-MOV-744, and AC-MOV-1870. This one-time extension for leak rate 
testing of the RHR valves would be deferred until prior to return to 
power following the current outage, which is defined as prior to 
Tavg exceeding 350  deg.F.
    Date of publication of individual notice in Federal Register: 
December 13, 1994 (59 FR 64224).
    Expiration date of individual notice: January 12, 1995.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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: December 16, 1994.
    Brief description of amendments: This amendment would revise 
Technical Specifications regarding diesel generator surveillance 
requirements.
    Date of publication of individual notice in the Federal Register: 
December 22, 1994 (59 FR 67350).
    Expiration date of individual notice: January 23, 1995.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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: January 3, 1995.
    Brief description of amendments: The amendments add a permissive 
statement to Surveillance Requirement 4.9.7.1 that [[Page 6318]] will 
allow the auxiliary building bridge crane interlocks and physical stops 
to be defeated during implementation of the spent fuel pool storage 
capacity increase modification.
    Date of publication of individual notice in the Federal Register: 
January 9, 1995 (60 FR 2404).
    Expiration date of individual notice: January 24, 1995.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennesee 37402.

Notice of Insurance 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.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of application for amendments: September 23, 1994.
    Brief description of amendments: The amendments revise the Unit 2 
Shutdown AC Power Sources TSs to permit a one-time increase the allowed 
outage time (AOT) from 7 to 14 days for the dedicated Class IE 
emergency power system and the Unit 1 control room emergency 
ventilation system TSs to permit a one-time increase the AOT from 7 to 
30 days. These one-time extensions are necessary to support 
modifications scheduled to be implemented during the upcoming 1995 Unit 
2 refueling outage.
    Date of issuance: January 11, 1995.
    Effective date: As of the date of issuance to be implemented during 
the 1995 Unit 2 refueling outage.
    Amendment Nos.: Unit 1-202 and Unit 2-180.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 26, 1994 (59 FR 
53835).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated January 11, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of application for amendment: May 15, 1993, as supplemented 
February 17, 1994, February 25, 1994, and November 23, 1994.
    Brief description of amendment: The amendment deletes Section 
2.C.(8) of the Facility Operating License NPF-63, and deletes 
Attachment 1 to the License, in response to your request dated May 15, 
1993, as supplemented February 17, 1994, February 25, 1994, and 
November 23, 1994
    Date of issuance: January 12, 1995.
    Effective date: January 12, 1995.
    Amendment No. 53.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 9, 1993 (58 FR 
32378).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 12, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

    Date of application for amendments: June 3, 1994.
    Brief description of amendments: The amendments revise Byron and 
Braidwood technical specifications (TSs) to reflect a primary-to-
secondary leakage rate of 150 gallons per day through any one steam 
generator and to reflect an inservice inspection of a minimum of 20 
percent of a random sample of the sleeves at the end-of-cycle. The 
amendment also adds a condition to the licenses to conduct additional 
corrosion testing to establish the design life for the sleeved tubes in 
the presence of a crevice. The revised TSs are more conservative than 
the previous TSs and were requested in order to increase the confidence 
in the ability of sleeves to maintain primary-to-secondary integrity.
    Date of issuance: January 6, 1995.
    Effective date: January 6, 1995.
    Amendment Nos.: 67, 67, 57, and 57.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the operating licenses and TSs.
    Date of initial notice in Federal Register: October 12, 1994 (59 FR 
51613). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 6, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: For Byron, the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481. [[Page 6319]] 

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

    Date of application for amendments: November 7, 1994, as 
supplemented December 16, 1994.
    Brief description of amendments: The amendments approve the use and 
storage of fuel with an enrichment not to exceed a nominal 5.0 weight 
percent U-235 in the spent fuel racks.
    Date of issuance: January 20, 1995.
    Effective date: January 20, 1995.
    Amendment Nos.: 68, 68, 58, and 58.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 7, 1994 (59 FR 
63115). The December 16, 1994, submittal provided additional clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in an Environmental 
Assessment dated January 13, 1995, and in a Safety Evaluation dated 
January 20, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: For Byron, the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481.

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: January 10, 1994, as 
supplemented September 15, 1994, January 5 and 10, 1995.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) Table 2.2-1 and TS 4.2.5 to allow a change in the 
method for measuring reactor coolant system (RCS) flow rate from the 
calorimetric heat balance method to a method based on a calibration of 
the RCS cold leg elbow differential pressure taps.
    Date of issuance: January 12, 1995.
    Effective date: To be implemented within 30 days from the date of 
issuance.
    Amendment Nos.: 153 and 135.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 16, 1994 (59 
FR 7688). The September 15, 1994, January 5 and 10, 1995, letters 
provided clarifying information that did not change the scope of the 
January 10, 1994, application, the Federal Register Notice 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 12, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of application for amendments: November 2, 1994.
    Brief description of amendments: These amendments clarify the 
actions required in the event of inoperable equipment associated with 
containment depressurization and cooling systems, and provide 
consistency between Unit 1 and Unit 2 requirements.
    Date of Issuance: January 18, 1995.
    Effective Date: January 18, 1995.
    Amendment Nos.: 131 and 70.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 7, 1994 (59 FR 
63122).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 18, 1995.
    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: April 19, 1994.
    Brief description of amendments: These amendments consist of 
changes to the Technical Specifications relating to surveillance 
requirements for inservice inspection and testing programs.
    Date of issuance: January 11, 1995.
    Effective date: January 11, 1995.
    Amendment Nos. 171 and 165.
    Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27054).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 11, 1995.
    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 No. 50-424, 
Vogtle Electric Generating Plant, Unit 1, Burke County, Georgia

    Date of application for amendment: August 16, 1994.
    Brief description of amendment: The amendment eliminated License 
Condition 2.C.(6) and the associated Attachment 1 of the license. 
License Condition 2.C.(6) referenced Attachment 1 which listed special 
diesel generator maintenance and surveillance requirements.
    Date of issuance: January 20, 1995.
    Effective date: To be implemented within 30 days from the date of 
issuance.
    Amendment No.: 81.
    Facility Operating License Nos. NPF-68: Amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: September 6, 1994 (59 
FR 46071).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 20, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830.

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: March 31, 1994.
    Brief description of amendments: The changes revised TS Table 3.7-1 
by lowering the maximum allowable power range neutron flux high 
setpoint when one or more main steam safety valves (MSSVs) are 
inoperable. The changes also revised the Bases for TS 3/4.7.1.1 
[[Page 6320]] to include the Westinghouse algorithm for determining the 
new setpoint values.
    Date of issuance: January 20, 1995.
    Effective date: To be implemented within 30 days from date of 
issuance.
    Amendment Nos.: 82 and 60.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37071).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 20, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830.

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: May 20, 1994.
    Brief description of amendments: The amendments relocate the heat 
flux hot channel factor, FQ(Z), penalty of 2 percent in 
specification 4.2.2.2.f to the cycle-specific Core Operating Limits 
Report (COLR) to allow for burnup-dependent values of the penalty in 
excess of 2 percent. This amendment also revises the reference in 
specification 6.8.1.6 to the Westinghouse FQ(Z) surveillance 
methodology in order to reflect Revision 1 of WCAP-10216-P, 
``Relaxation of Constant Axial Offset Control--FQ Surveillance 
Technical Specification,'' approved by the NRC on November 26, 1993.
    Date of issuance: January 11, 1995.
    Effective date: To be implemented within 30 days from the date of 
issuance.
    Amendment Nos.: 79 and 58.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37072). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 11, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830.

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: August 16, 1994.
    Brief description of amendments: The amendments change Technical 
Specification 3/4.7.1.1 and its Bases regarding the setpoint tolerance 
for the main steam safety valves.
    Date of issuance: January 12, 1995.
    Effective date: To be implemented within 30 days from the date of 
issuance.
    Amendment Nos.: 80 and 59.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 14, 1994 (59 
FR 47168).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 12, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830.

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

    Date of application for amendment: July 29, 1994.
    Brief description of amendment: The proposed amendment would revise 
the Technical Specifications by deleting reference to written relief 
from ASME Code requirements. The revised Technical Specifications refer 
to the applicable provision of NRC regulations concerning the ASME 
Code.
    Date of issuance: January 6, 1995.
    Effective date: January 6, 1995, to be implemented within 120 days.
    Amendment No.: 206.
    Facility Operating License No. DPR-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 
45026).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 6, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, Iowa 52401.

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

    Date of application for amendment: March 15, 1994, as supplemented 
October 20, 1994.
    Brief description of amendment: This amendment allows the use of 
integral fuel burnable absorbers as a method of controlling core excess 
reactivity and maintaining core power distribution within acceptable 
peaking limitations.
    Date of issuance: January 17, 1995
    Effective date: January 17, 1995.
    Amendment No.: 145.
    Facility Operating License No. DPR-36: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1994 (59 FR 
22010).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 17, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, Maine 04578.

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

    Date of application for amendment: October 5, 1994.
    Brief description of amendment: The amendment revises the 
applicability requirements of Technical Specification (TS) 3.7.3 to 
require operability of the Control Room Outdoor Air Special Filter 
Train System in Operational Conditions 1, 2, 3 and ** rather than in 
all Operational Conditions and **. The applicability requirements for 
Action Statement b. of TS 3.7.3 and for the Radiation Monitoring 
Instrumentation required operable by TS Tables 3.3.7.1-1 and 4.3.7.1-1 
are being changed in a similar manner. The amendment also adds a 
notation to Action Statement b.1. of TS 3.7.3 stating that the 
provisions of TS 3.0.4 are not applicable for entry into Operational 
Condition ** when one filter train is inoperable provided an operable 
filter train is in operation in the emergency pressurization mode of 
operation.
    Date of issuance: January 18, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 60.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications. [[Page 6321]] 
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55874).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 18, 1995.
    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, Docket No. 50-245, Millstone Nuclear 
Power Station, Unit 1, New London County, Connecticut

    Date of application for amendment: October 4, 1994.
    Brief description of amendment: The amendment relocates the primary 
containment isolation valve list from Technical Specification (TS) 
Section 3.7.D to the Millstone Unit 1 Technical Requirements Manual. 
This change is in accordance with the guidance of Generic Letter 91-08. 
The amendment also makes administrative and editorial changes to TS 
Sections 3.7.D and 4.7.D and makes changes to the associated bases.
    Date of issuance: January 10, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 78.
    Facility Operating License No. DPR-21. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 23, 1994 (59 
FR 60383)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 10, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.

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

    Date of application for amendment: October 14, 1994.
    Brief description of amendment: The amendment clarifies the low 
pressure coolant injection requirements as required by Technical 
Specification 4.5.A.2.
    Date of issuance: January 9, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 77.
    Facility Operating License No. DPR-21. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 7, 1994 (59 FR 
63125).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 9, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resource 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: September 9, 1994, with 
clarifying information provided by letter dated October 5, 1994.
    Brief description of amendment: The amendment revises the Technical 
Specifications to modify surveillance requirements by increasing the 
acceptance criterion for the closure of the main steam isolation valves 
from 5 seconds to 10 seconds.
    Date of issuance: January 10, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 101.
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 19, 1994 (59 
FR 47960). The October 5, 1994, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated Janaury 10, 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 17, 1994, as 
supplemented October 27, 1994.
    Brief description of amendments: The amendments revise the Prairie 
Island Nuclear Generating Plant Technical Specifications to change the 
submittal frequency of the Radioactive Effluent Release Report from 
semiannual to annual in accordance with 10 CFR Part 50.36a.
    Date of issuance: January 11, 1995.
    Effective date: January 11, 1995, with full implementation within 
30 days.
    Amendment Nos.: 114 and 107.
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 7, 1994 (59 FR 
63125) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 11, 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.

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: July 27, 1994.
    Brief description of amendments: These amendments revise the 
Technical Specifications (TS) definition of ``Core Alteration'' to 
conform to the definition approved by the staff for the current boiling 
water reactor (BWR) improved TS in NUREG-1433, ``Standard Technical 
Specifications, General Electric Plants, BWR/4.''
    Date of issuance: January 3, 1995.
    Effective date: January 3, 1995.
    Amendment Nos.: 138 and 108.
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 14, 1994 (59 
FR 47177).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 3, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Osterhout Free Library, 
[[Page 6322]] Reference Department, 71 South Franklin Street, Wilkes-
Barre, Pennsylvania 18701.

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

    Date of application for amendments: August 22, 1994.
    Brief description of amendments: These amendments change Technical 
Specifications 3/4.1.3 to: (1) Extend the scram discharge volume (SDV) 
vent or drain valve restoration time from the current time period of 24 
hours to 7 days; (2) permit the SDV vent and drain valves operability 
check to be performed at shutdown conditions instead of at-least-once-
per-18-months; and (3) delete the SDV float switch response 
surveillance requirement.
    Date of issuance: January 9, 1995.
    Effective date: January 9, 1995.
    Amendment Nos.: 139 and 109.
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 28, 1994 (59 
FR 49433).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 9, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.

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

    Date of application for amendments: September 26, 1994.
    Brief description of amendments: The amendments change the 
Technical Specifications for each of the units to remove the 
requirement for the average power range monitors (APRMs) to be operable 
while the plant is in Operational Condition 5, refueling status. 
However, the amendment does not change the requirement for the APRMs to 
be operable when the reactor mode switch is in Startup during a 
shutdown margin demonstration.
    Date of issuance: January 9, 1995.
    Effective date: January 9, 1995.
    Amendment Nos.: 140 and 110.
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55880).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 9, 1995.
    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.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: August 25, 1993, as 
supplemented by letter dated August 4, 1994.
    Brief description of amendments: These amendments modify Technical 
Specification (TS) Section 3.3.7.8.2 and associated Bases 3/4.3.7.8 
regarding the Main Control Room (MCR) toxic gas detection system. The 
TS change reflects the implementation of a modification designed to 
eliminate spurious high toxic gas concentration alarms received by the 
MCR.
    Date of issuance: January 19, 1995.
    Effective date: January 19, 1995.
    Amendment Nos. 84 and 45.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 29, 1993 (58 
FR 50971).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 19, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: July 20, 1994, as supplemented 
September 23, 1994.
    Brief description of amendments: The amendments would raise the 
Steam Leakage Detection system set-points that isolate the High 
Pressure Coolant Injection System (HPCI) and Reactor Core Isolation 
Cooling (RCIC) system equipment on high equipment room temperature and 
high delta temperature. The amendments are supported by a Limerick 
Generating Station modification to increase the environmental 
qualifications limits of the HPCI and RCIC systems to allow the systems 
to remain operable when equipment room cooling is unavailable.
    Date of issuance: January 20, 1995.
    Effective date: January 20, 1995.
    Amendment Nos. 85 and 46.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 14, 1994 (59 
FR 47178).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 20, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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: April 18,1994, as supplemented 
October 25, 1994.
    Brief description of amendment: The amendment revises TS Section 
3.14 (Fire Protection and Detection Systems--Limiting Conditions for 
Operation), TS Section 4.12 (Fire Protection and Detection Systems--
Surveillances) and TS Section 6.0 (Administrative Controls) to relocate 
the fire protection requirements from the TSs to the IP3 Operational 
Specifications Manual. In addition, the amendment revised the IP3 
Facility Operating License to include the NRC's standard fire 
protection license condition. These changes were made in accordance 
with the guidance provided in Generic Letter (GL) 86-10, 
``Implementation of Fire Protection Requirements,'' and GL 88-12, 
``Removal of Fire Protection Requirements from Technical 
Specifications.''
    Date of issuance: January 13, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 157.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27065).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 13, 1995.
    No significant hazards consideration comments received: No. 
[[Page 6323]] 
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: December 8, 1994.
    Brief description of amendment: The amendment revises TS section 
4.4.E.1 to allow a one-time extension to the 30-month interval 
requirement for leak rate testing of Residual Heat Removal containment 
isolation valves AC-732, AC-741, AC-MOV-743, AC-MOV-744 and AC-MOV-
1870.
    Date of issuance: January 13, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 158.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 13, 1994 (59 
FR 64223).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 13, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: June 29, 1994, as supplemented 
December 2, 1994.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) Section 6.5, ``Review and Audit,'' and TS Section 
6.8, ``Procedures,'' to establish a new review and approval process for 
nuclear safety-related procedures and to modify membership requirements 
for the Plant Operating Review Committee. The amendment also revised TS 
Section 6.5 to delete review and audit responsibilities for the 
Emergency and Security Plans consistent with Generic Letter 93-07, 
``Modification of the Technical Specification Administrative Control 
Requirements for Emergency and Security Plans.''
    Date of issuance: January 17, 1995.
    Effective date: As of the date of issuance to be implemented within 
90 days.
    Amendment No.: 159.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37081).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 17, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: June 17, 1994, as supplemented 
December 2, 1994.
    Brief description of amendment: The amendment revises Section 6.5, 
``Review and Audit,'' and Section 6.8, ``Procedures,'' of the Technical 
Specifications (TSs) to establish a new review and approval process for 
nuclear safety-related procedures. The amendment also revises Section 
6.5 to modify membership requirements for the Plant Operating Review 
Committee and to delete review and audit responsibilities for the 
Emergency and Security Plans from the TSs consistent with Generic 
Letter 93-07, ``Modification of the Technical Specification 
Administrative Control Requirements for Emergency and Security Plans.''
    Date of issuance: January 18, 1995.
    Effective date: As of the date of issuance to be implemented within 
90 days.
    Amendment No.: 222.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37082).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 18, 1995.
    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.

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

    Date of application for amendment: September 19, 1994.
    Brief description of amendment: The amendment revised the Technical 
Specifications for the snubber visual inspection schedule.
    Date of issuance: January 20, 1995.
    Effective date: January 20, 1995 and to be implemented within 90 
days.
    Amendment No. 68.
    Facility Operating License No. NPF-58. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 26, 1994 (59 FR 
53843).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 20, 1995.
    No significant hazards consideration comments received: No.
     Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.

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

    Date of application for amendment: May 1, 1992 as clarified by 
facsimile transmission dated January 10, 1995.
    Brief description of amendment: The amendment revises the LIMITING 
CONDITIONS FOR OPERATION and SURVEILLANCE REQUIREMENTS for the 
containment air locks, changes the exception for containment 
penetration status verification to include the annulus, clarifies 
containment air lock testing intervals, and clarifies the definition 
and bases for containment integrity.
    Date of issuance: January 17, 1995.
    Effective date: Date of issuance and to be implemented within 90 
days.
    Amendment No. 194.
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 2, 1992 (57 
FR 40221).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 17, 1995. [[Page 6324]] 
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington

    Date of application for amendment: August 8, 1994.
    Brief description of amendment: The amendment modifies the 
Technical Specifications (TS) to delete the requirement to obtain prior 
written relief from the Commission for inservice inspection (ISI) and 
inservice testing (IST) of components conducts pursuant to 10 CFR 
50.55a. The amendment also adds a definition for the word ``biennial.''
    Date of issuance: January 5, 1995.
    Effective date: January 5, 1995.
    Amendment No.: 133.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 14, 1994 (59 
FR 56558).
    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: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352.

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

    Date of application for amendments: November 23, 1993, as 
supplemented January 10, 12, and 13, 1995.
    Brief description of amendments: These amendments revise the 
operating conditions and limiting conditions for operation for 
containment systems, and revise corresponding definitions and tests. In 
addition, the related bases are updated to ensure consistency and 
clarity.
    Date of issuance: January 18, 1995.
    Effective date: January 18, 1995, to be implemented within 45 days.
    Amendment Nos.: 160 and 164.
    Facility Operating License Nos. DPR-24 and DPR-27. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 19, 1994 (59 FR 
2875).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 18, 1995.
    The January 10, 12, and 13, 1995 submittals provided supplemental 
information that did not change the initial proposed no significant 
hazards consideration determination.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.

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

    Date of amendment request: October 27, 1993.
    Brief description of amendment: The amendment changes Note 5 of 
Technical Specification Table 4.3-1 to reflect the use of integral bias 
curves, rather than detector plateau curves, to calibrate the source 
range instrumentation.
    Date of issuance: January 9, 1995.
    Effective date: January 9, 1995, to be implemented within 30 days.
    Amendment No.: 83.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1993 (58 
FR 62159). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 9, 1995.
    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: October 21, 1994, as supplemented by 
letters dated October 27, 1994 and December 2, 1994.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) Surveillance Requirements 4.7.1.2.1.c.2, operability 
testing of the auxiliary feedwater (AFW) pump auto start feature, and 
4.3.2.2, engineered safety features (ESF) time response testing of the 
AFW pumps to exempt the testing of the turbine-driven AFW pump from the 
provisions of TS 4.0.4 for entry into Mode 3. In addition, TS 
Surveillance Requirement 4.7.1.2.1.c is revised to delete the 
requirement that the 18 month AFW surveillance be performed during 
shutdown.
    Date of issuance: January 20, 1995.
    Effective date: January 20, 1995, to be implemented within 30 days.
    Amendment No.: 84.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 23, 1994 (59 
FR 60389) The December 2, 1994, supplemental letter provided clarifying 
information and did not change the initial no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 20, 1995.
    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 25th 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. 95-2350 Filed 1-31-95; 8:45 am]
BILLING CODE 7590-01-P