[Federal Register Volume 62, Number 166 (Wednesday, August 27, 1997)]
[Notices]
[Pages 45452-45471]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-10827]


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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 August 4, 1997, through August 15, 1997. The 
last biweekly notice was published on August 13, 1997 (62 FR 43365).

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

[[Page 45453]]

margin of safety. The basis for this proposed determination for each 
amendment request is shown below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By September 26, 1997, the licensee may file a request for a 
hearing with respect to issuance of the amendment to the subject 
facility operating license and any person whose interest may be 
affected by this proceeding and who wishes to participate as a party in 
the proceeding must file a written request for a hearing and a petition 
for leave to intervene. Requests for a hearing and a petition for leave 
to intervene shall be filed in accordance with the Commission's ``Rules 
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. 
Interested persons should consult a current copy of 10 CFR 2.714 which 
is available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for

[[Page 45454]]

amendment which is available for public inspection at the Commission's 
Public Document Room, the Gelman Building, 2120 L Street, NW., 
Washington, DC, and at the local public document room for the 
particular facility involved.

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

    Date of amendments request: June 12, 1997
    Description of amendments request: The proposed amendments would 
revise the Limiting Condition for Operation (LCO) of Technical 
Specification 3.6.1.6 to limit drywell average air temperature instead 
of primary containment average air temperature, which is the volume-
weighted average of both drywell and wetwell atmospheres. This change 
in monitored parameter is consistent with the approach taken in the 
improved standard technical specifications for boiling water reactor 
(BWR) plants of this type (NUREG-1433, Rev. 1, ``Standard Technical 
Specifications General Electric Plants, BWR/4,'' April 1995). The 
proposed amendments would additionally change the temperature limit in 
this LCO from 135 deg.F (primary containment average air temperature) 
to 150 deg.F (drywell average air temperature).
    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 NRC has provided standards in 10 CFR 50.92 for determining 
whether a significant hazards consideration exists. A proposed 
amendment to an operating license for a facility involves no 
significant hazards consideration if operation of the facility in 
accordance with the proposed amendment would not: (1) involve a 
significant increase in the probability or consequences of an 
accident previously evaluated, (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. Carolina Power & Light Company has reviewed these proposed 
license amendment requests and has concluded that their adoption 
would not involve a significant hazards consideration. The basis for 
this determination follows.
    1. The probability of previously evaluated accidents is not a 
function of the ambient drywell air temperature. The revised drywell 
average air temperature limit of 150 deg.F does not affect any 
instrumentation setpoints or allowable values, so [the] likelihood 
of plant instrumentation initiating a plant transient or accident 
has not been increased.
    The design basis accidents were re-evaluated using an initial 
drywell air temperature of 150 deg.F. The evaluation results 
indicate that no containment design requirements are exceeded nor 
are any regulatory requirements exceeded. Analyses demonstrate that 
an initial drywell average air temperature of 150 deg.F will ensure 
that the safety analysis remains valid by ensuring that the peak 
loss-of-coolant accident drywell temperature does not result in the 
drywell structure exceeding the maximum allowable temperature of 
300 deg.F. Indeed, these evaluations indicate that both the peak 
drywell pressure and temperature will be slightly less than the peak 
drywell pressure and temperature resulting from the current 
135 deg.F primary containment air temperature limit. Since the 
drywell temperature and pressure associated with a postulated design 
basis accident remain less than the drywell maximum design allowable 
values, revised drywell average air temperature limit of 150 deg.F 
does not increase the consequences of an accident previously 
evaluated.
    A temporary, one-time exception footnote for the Brunswick Steam 
Electric Plant (BSEP), Unit No. 2 is being deleted because the 
period of the footnote's applicability expired on August 15, 1985. 
Deletion of this footnote is an administrative change that has no 
effect on the probability or consequences of an accident previously 
evaluated.
    Thus, based on the above, the proposed license amendments do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed amendments would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. Revising the primary containment temperature limit basis 
to use the drywell average air temperature and increasing the 
average air temperature limit from 135 deg.F to 150 deg.F does not 
physically modify the facility nor does the proposed revision modify 
the operation of any existing plant equipment. A temporary, one-time 
exception footnote for BSEP Unit No. 2 is being deleted because the 
period of the footnote's applicability expired on August 15, 1985. 
Deletion of this footnote is an administrative change that does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety. The drywell average airspace 
temperature affects the calculated containment response to 
postulated Design Basis Accidents. Analyses demonstrate that an 
initial drywell average air temperature of 150 deg.F will ensure 
that the safety analysis remains valid by ensuring that the peak 
loss-of-coolant accident drywell air temperature does not result in 
the drywell structure exceeding the maximum allowable temperature of 
300 deg.F. Analyses performed using an initial drywell average air 
temperature of 150 deg.F also demonstrate that containment design 
requirements for peak post-accident suppression pool temperature, 
design basis accident related discharge loads for safety-relief 
valve piping, and net positive suction head for residual heat 
removal system and core spray system pumps are met. In addition, 
setpoints for reactor water level instrumentation located in the 
drywell have not been adversely affected, drywell equipment 
environmental qualification is being maintained, and containment 
performance during a postulated station blackout is not being 
adversely affected. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety. The deletion of a 
temporary, one-time exception footnote for BSEP Unit No. 2 is an 
administrative change that also 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 North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: Gordon E. Edison (Acting)

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

    Date of amendments request: July 18, 1997Description of amendments 
request: The proposed amendments would revise two specifications 
included in the Design Features section of the Technical Specifications 
(TS). The value for primary containment suppression chamber design 
temperature (TS 5.2.2.b) would be increased from 200 deg.F to 
220 deg.F. The licensee has determined that the original suppression 
chamber design temperature was 220 deg.F and confirmed that it is still 
the correct design value. Secondly, the specification for reactor 
coolant system volume (TS 5.4.2) would be redefined as the vessel 
volume, rather than the vessel and recirculation system volume, 
resulting in a change in the associated value from 18,670 cubic feet to 
18,320 cubic feet. Additionally, the proposed amendments would correct 
a typographical error in Design Features TS 5.3.2 regarding the reactor 
core control rod assemblies.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 45455]]

licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:
    10 CFR 50.92 provides standards for determining whether a 
significant hazards consideration exists. A proposed amendment to an 
operating license for a facility involves no significant hazards 
consideration if operation of the facility in accordance with the 
proposed amendment would not: (1) involve a significant increase in 
the probability or consequences of an accident previously evaluated, 
(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. Carolina Power & Light Company has 
reviewed these proposed license amendment requests and has concluded 
that their adoption would not involve a significant hazards 
consideration. The basis for this determination follows.
    1. The proposed license amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The proposed amendments correct an inaccurate 
suppression chamber design temperature to reflect the actual design 
temperature used during containment analyses and pressure vessel 
procurement, correct a typographical error, and update the reactor 
coolant system volume to reflect a more accurate volume used in 
current analyses. These changes are administrative in nature and do 
not affect the probability or consequences of any accident 
previously analyzed.
    2. The proposed license amendments will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. These changes are administrative in nature and 
correct the Technical Specifications to accurately represent 
information used during existing accident analyses. These changes do 
not introduce a new initiating event and do not create the 
possibility of a new or different kind of accident previously 
evaluated.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety. As stated above, these changes are 
administrative in nature and correct the Technical Specifications to 
accurately represent information used during existing accident 
analyses. These changes document values currently used in existing 
accident analyses and, therefore, do not reduce the margin of safety 
already established by the analyses.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: Gordon E. Edison (Acting)

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

    Date of amendment request: July 1, 1997
    Description of amendment request: The proposed amendments would 
revise Technical Specification Table 3.3.7.1-1, ``Radiation Monitoring 
Instrumentation,'' to require two channels to be operable per trip 
system as opposed to two per intake. This change reflects a 
modification to the design of the instrument logic to satisfy single 
failure requirements. The amendment would also revise the associated 
action statement to clarify system logic wording.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    The proposed Technical Specification (TS) change clearly defines 
the system logic and the specific actions required for system 
operability. It will not change the probability of occurrence of any 
accidents, because the affected radiation monitoring instrumentation 
is not an accident initiator. UFSAR [Updated Final Safety Analysis 
Report] Section 15.9.3.4 analyzed the effects of the loss of 
ventilation from the Main Control Room in the event of a Station 
Black Out (SBO). The scope of work for the design change associated 
with this TS change does not affect this analysis or any of its 
assumptions The consequences of an accident will not increase, 
because the trip system redundancy is being restored to meet design 
basis requirements. The proposed design change will eliminate the 
potential of exposing main control room personnel to radiation doses 
that exceed the limits specified in General Design Criteria (GDC) 
19. The design change associated with this TS change will comply 
with the redundancy due to two trip systems, either of which will 
actuate the control room emergency makeup train as required and the 
potential for spurious actuations will be reduced due to the logic 
change to require two channels of one trip system to cause 
actuation. The overall control logic for the remaining portions of 
the CREFS [Control Room Emergency Filtration System] is not changed 
by the design change.
    The changes proposed to the actions are intended to clarify 
system logic wording. The actions assure that automatic trip 
capability is maintained and if not, then the CREFS is placed in the 
pressurization mode as in the current TS. This is consistent with 
the current TS.
    Based upon the above, the proposed amendment will not increase 
the probability or consequences of any accident previously 
evaluated.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    The elimination of the electrical connection between the 
redundant trip systems in a given CREFS subsystem will restore trip 
system independence and eliminate the potential of a single failure 
disabling the radiation monitoring instrumentation trip function. 
Specifically, a single failure, resulting from a blown fuse caused 
by a fault in the affected existing circuit, could remove the 
control power to the isolation logic relays in both trip systems. 
These relays require power in order to actuate and perform their 
safety function. A loss of control power to both trip systems due to 
the fault could result in exposing main control room personnel to 
radiation doses that exceed GDC 19 limits.
    In addition, the changes to Action Statement 70 of the 
specification assure that trip capability is maintained.
    Based upon the above, the proposed change will not create the 
possibility of a new or different kind of accident or transient 
previous evaluated.
    3) Involve a significant reduction in the margin of safety 
because:
    The proposed TS change will not prevent the isolation logic 
relays from performing their function or cause false trips. The 
alarm/trip setpoints for the affected monitors (including their 
measurement ranges) remain unchanged. The changes proposed to the 
actions are intended to clarify system logic wording. The actions 
assure that automatic trip capability is maintained and if not, then 
the CREFS is placed in the pressurization mode as in the current TS. 
This is consistent with the current TS.
    Based on the above, the proposed TS change does not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document location:  Jacobs Memorial Library, Illinois 
Valley Community College, Oglesby, Illinois 61348
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

[[Page 45456]]

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

    Date of amendment request: August 5, 1997
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications for the Safety Limit Minimum 
Critical Power Ratio (SLMCPR) for Cycle 8 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:
    The plant/cycle specific SLMCPRs have been calculated using 
methods identical to those used by GE (General Electric) to assess 
the SLMCPR for other BWRs (boiling water reactors). Similar methods 
were used to determine the value of the SLMCPR for the previous 
cycle. These methods are within the existing design and licensing 
basis and cannot increase the probability or severity of an 
accident. The basis of the SLMCPR calculation is to ensure that 
greater that 99.9% of all fuel rods in the core avoid transition 
boiling and fuel damage in the event of the occurrence of 
Anticipated Operational Occurrences (AOO) or a postulated accident.
    The SLMCPR is used to establish the Operating Limit Minimum 
Critical Power Ratio (OLMCPR). Neither the SLMCPR nor the OLMCPR are 
initiators or affect initiators of an accident previously evaluated 
and therefore changes to the SLMCPR do not increase the probability 
of any accident previously evaluated. The proposed changes involve 
the use of an accepted methodology in calculating the SLMCPR and, 
since there is no change in the definition of the SLMCPR, these 
changes will not affect the consequences of any accident previously 
evaluated. In addition, the proposed changes do not involve any 
change in the way the plant is operated. Existing procedures will 
ensure that the SLMCPR is not violated. Therefore, these changes 
have no effect on the consequences of an accident.
    On these bases, there will be no increase in the probability or 
consequences of an accident previously analyzed as a result the 
proposed changes.
    The proposed changes consist of SLMCPR calculated from an 
accepted method of analysis which has been used by many BWRs. These 
changes do not involve any alteration of the plant and do not affect 
the plant operation. Neither the SLMCPR nor the OLMCPR can initiate 
an event, therefore a change to the SLMCPR does not create the 
possibility of occurrence of a new or different kind of accident 
from any accident previously evaluated.
    The SLMCPR is a Technical Specification numerical value to 
ensure that 99.9% of all fuel rods in the core will avoid transition 
boiling if the limit is not violated. The proposed SLMCPR change 
results from SLMCPR analysis using the accepted methods as 
identified in the Attachment.
    The margin of safety resides between the SLMCPR and the point at 
which fuel fails. Maintaining the MCPR above the proposed SLMCPR 
will maintain the margin of safety associated with GE's SLMCPR 
methodology. Existing plant procedures will continue to ensure that 
the SLMCPR is not violated.
    Therefore, this request does not involve a reduction in the 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document location: Government Documents Department, 
Louisiana State University, Baton Rouge, LA 70803
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005
    NRC Project Director: James W. Clifford, Acting

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

    Date of amendment request: July 28, 1997
    Description of amendment request: This amendment is to modify the 
actions associated with Technical Specifications Table 3.3-1 for the 
Reactor Protective Instrumentation and Table 3.3-3 for the Engineered 
Safety Feature Actuation System Instrumentation.
    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:
    An evaluation of the proposed change has been performed in 
accordance with 10 CFR 50.91(a)(1) regarding no significant hazards 
considerations using the standards in 10 CFR 50.92(c). A discussion 
of these standards as they relate to this amendment request follows:
    1. Does Not Involve a Significant Increase in the Probability or 
Consequences of an Accident Previously Evaluated.
    The proposed change to the ANO-2 Technical Specifications (TS) 
modifies the allowed outage time that a channel of the Refueling 
Water Tank (RWT) Level - Low or Steam Generator differential 
pressure (delta P) can be in the tripped condition from a maximum of 
approximately 18 months when one channel is inoperable, and 31 days 
when two channels are inoperable, to 48 hours for either of these 
conditions.
    If a channel of RWT Level Low is in the tripped condition and a 
single failure occurs that results in one of the other three 
channels of RWT Level - Low to actuate, a Recirculation Actuation 
System (RAS) signal would be generated. This scenario would not be 
considered severe if the condition occurred as a single event. 
However, during the injection phase of a Loss of Coolant Accident 
(LOCA) with a channel of RWT Level - Low in the trip condition with 
the above single failure, a premature RAS actuation would be the 
result. The premature RAS actuation would prevent the contents of 
the RWT from being injected into the reactor coolant system and 
possibly resulting in failure of both trains of Emergency Core 
Cooling System (ECCS) and the Containment Spray System.
    With one channel of Steam Generator delta P in the tripped 
condition, as allowed by the TS, the plant is vulnerable to the 
single failure of a second Steam Generator delta P channel under an 
unisolable Main Steam Line Break condition. The following scenario 
will result in the faulted Steam Generator being supplied feedwater 
by the Emergency Feedwater System during an unisolable Main Steam 
Line Break. One channel of Steam Generator delta P is in the tripped 
condition as allowed by the TS and a Main Steam Line Break occurs 
that is unisolable. During this event one of the remaining channels 
of Steam Generator delta P fails resulting in incorrectly feeding 
the faulted Steam Generator. Reducing the time that a channel of RWT 
Level - Low or Steam Generator delta P can be placed in the tripped 
condition will reduce the probability of these scenarios from 
occurring.
    The consequences of feeding the faulted Steam Generator during a 
main steam line break event or a premature RAS actuation during a 
LOCA are both significant. The proposed change reduces the allowed 
time a channel of RWT Level - Low or Steam Generator delta P can be 
in the tripped condition. Reducing the time the channel can be in 
the tripped condition and thus, the exposure time to this scenario, 
would not be an accident initiator or involve an increase in the 
consequences of any accident previously evaluated.
    The remaining proposed changes are consistent with NUREG-1432, 
``Standard Technical Specifications for Combustion Engineering 
Plants'' and are intended to correct the actions required by TS 
Tables 3.3-1 and 3.3-3 to the current NRC approved guidance.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any 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 the design or configuration 
of the plant. The proposed change provides a more conservative time 
limit for a channel to be in the tripped condition and provides the 
required actions when a channel is out of service. There has been no 
physical change to plant systems, structures or components nor will 
the proposed change reduce the ability of any of the safety related 
equipment required to mitigate anticipated operational

[[Page 45457]]

occurrences or accidents. This change will potentially increase the 
ability of safety related equipment to perform their functions. The 
configuration allowed by the proposed specification is permitted by 
the existing specification.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Does Not Involve a Significant Reduction in the Margin of 
Safety.
    The proposed change provides a more restrictive time limit for a 
channel of RWT Level Low or Steam Generator delta P to be in the 
tripped condition than is currently allowed by the TS. By reducing 
the allowed time, the probability is reduced that a single failure 
of another channel would result in a premature RAS actuation during 
the injection phase of a LOCA or the feeding of a faulted Steam 
Generator. By limiting the vulnerability to these events and their 
consequences, the proposed change will increase the margin of 
safety.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, Entergy Operations has 
determined that the requested change 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 location: Tomlinson Library, Arkansas Tech 
University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: James W. Clifford, Acting

Florida Power and Light Company, Docket No. 50-335, St. Lucie 
Plant, Unit No. 1, St. Lucie County, Florida

    Date of amendment request: July 22, 1997
    Description of amendment request: The proposed amendment will 
incorporate a recent evaluation of a postulated inadvertent opening of 
a Main Steam Safety Valve (MSSV) into the current licensing basis for 
St. Lucie Unit 1. An assessment of the potential consequences of this 
specific transient is not presently contained in the Updated Final 
Safety Analysis Report (UFSAR), and the proposed license amendment is 
required by 10 CFR 50.59(c).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The Unit 1 UFSAR includes analyses for excess load events; 
however, a stuck open MSSV is not specifically evaluated in the 
UFSAR. This proposed amendment will add an evaluation of an 
inadvertent opening of an MSSV to the licensing basis of the plant. 
The probability of occurrence of an excess load event is not 
increased by this amendment since the frequency of initiating events 
has not changed and there is no change to the plant or plant 
operation as a result of this amendment. Thus, there is no 
significant increase in the probability of any accident previously 
analyzed.
    The radiological consequences of an excess load event other than 
steam line ruptures are discussed in UFSAR Section 15.2.11.2.3, and 
are based on the inadvertent opening of an Atmospheric Steam Dump 
Valve (ADV). This proposed amendment revises the radiological 
consequences of the UFSAR excess load event to incorporate the 
results of a recent evaluation of an inadvertent opening of an MSSV. 
The consequences of the postulated MSSV scenario are greater than 
those of an inadvertent opening of an ADV, but the predicted two 
hour site boundary doses remain a small fraction of 10 CFR 100 
limits. In addition, the Unit 1 results are bounded by the St. Lucie 
Unit 2 analysis results which are reported in Section 15.1.3.1.1.3 
of the Unit 2 UFSAR. Therefore, operation of the facility in 
accordance with the proposed amendment will not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendment will add an evaluation of an inadvertent 
opening of an MSSV to the licensing basis of the plant. The 
evaluation addresses an anticipated operational occurrence (AOO) and 
is classified as an Excess Load event under the PSL1 [Plant St. 
Lucie Unit 1] accident classification criteria. Although an analysis 
of this specific transient is not currently provided in the UFSAR, 
analyses of Excess Load events other than steam line ruptures are 
reported in UFSAR Section 15.2.11. The amendment does not change 
plant design or operation and does not introduce new failure modes 
or system interactions. Thus, operation of the facility with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed license amendment adds an engineering evaluation to 
the licensing basis of the plant to address the consequences of a 
postulated stuck open MSSV. A change is not being made to plant 
design or operation. A change is not being made to any Technical 
Specification Limiting Condition for Operation, Action, or 
Surveillance Requirement. The evaluation demonstrates that, post-
trip, the reactor would remain subcritical throughout the transient, 
and that the radiological consequences of a stuck open MSSV are a 
small fraction of 10 CFR 100 limits. Therefore, operation of the 
facility in accordance with the proposed amendment 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document location: Indian River Community College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: August 1, 1997
    Description of amendment request: The proposed amendment will 
extend the semi-annual surveillance interval specified in Table 4.3-2 
of the Technical Specifications for testing the Engineered Safety 
Features Actuation System (ESFAS) subgroup relays to an interval 
consistent with Combustion Engineering Owners Group Report CEN-403, 
Revision 1-A, March 1996. The proposed surveillance interval is at 
least once per 18 months, with testing to be performed 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) Operation of the facility, in accordance with the proposed 
amendment, would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendment revises the testing frequency of ESFAS 
subgroup relays, and is based on demonstrated relay reliability. 
These relays actuate the engineered safety features (ESF) equipment 
which is installed to mitigate design basis accidents. ESF system 
components are not considered initiators of any design basis 
accident. Therefore, operation of the facility

[[Page 45458]]

with the proposed amendment would not involve a significant increase 
in the probability of an accident previously evaluated.
    The proposed amendment does not alter the design or operation of 
ESF systems. The mean time between failures demonstrated by the 
ESFAS subgroup relays is significantly greater than the proposed 
surveillance interval, and testing will be performed on a staggered 
test basis. This, in addition to ESF redundancy, provides assurance 
that these systems will continue to function as evaluated to 
mitigate design basis accidents. Therefore, operation of the 
facility, in accordance with the proposed amendment, would not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendment will not change the physical plant or the 
modes of operation defined in the facility license. The changes do 
not involve the addition of new equipment or the modification of 
existing equipment, nor do they alter the design of St. Lucie plant 
systems. Therefore, operation of the facility, in accordance with 
the proposed amendment, would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed amendment revises the surveillance interval for 
testing the ESFAS subgroup relays consistent with the Combustion 
Engineering Owners Group topical report CEN-403, Revision 1-A, and 
conforms to criteria specified in the associated safety evaluation 
issued by the NRC staff. The St. Lucie Unit 2 subgroup relay mean 
time between failures is significantly greater than the proposed 
surveillance interval, and testing will be performed on a staggered 
test basis. ESFAS setpoints, system operation, and plant 
configuration will not be changed, and the subgroup relays are not 
subject to time-related instrument drift. Accident analyses 
assumptions, initial conditions, and conclusions reported in the 
Updated Final Safety Analysis Report are not changed by the revised 
surveillance interval. Therefore, operation of the facility in 
accordance with the proposed amendment 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: July 30, 1997
    Description of amendment request: The purpose of this Technical 
Specification change request (TSCR) is to incorporate additional system 
leakage limits and leak test requirements for systems outside 
containment which were not previously contained in Technical 
Specification 4.5.4 nor considered in the TMI-1 Updated Final Safety 
Analysis Report (UFSAR) design basis accident (DBA) analysis dose 
calculations for 2568 MWt. This TSCR also revises the Technical 
Specification 3.15.3 Bases for the Auxiliary and Fuel Handling Building 
Ventilation System (AFHBVS). The revisions to Technical Specification 
3.15.3 Bases for the AFHBVS serve to clarify system design requirements 
and accident analysis considerations. The revision states that the 
AFHBVS is not credited in reducing off-site dose for the Maximum 
Hypothetical Accident (MHA) or the Waste Gas Tank Rupture (WGTR) 
accident analysis dose calculations.
    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:
    GPUN has determined that this TSCR poses no significant hazards 
consideration as defined by 10 CFR 50.92.
    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated. No physical modifications which would change 
structures, systems, or components are being made or proposed by 
this TSCR. This change has no [effect] on the LOCA [loss-of-coolant 
accident] safety analysis for ECCS [emergency core cooling system] 
performance. The results of revised MHA dose calculation are less 
than that previously evaluated in the UFSAR for the exclusion area 
boundary (EAB). In addition the doses are below the 10 CFR 100 
guideline limits for both the EAB and low population zone (LPZ) ..., 
and below the 10 CFR 50 Appendix A, GDC [General Design Criteria]-19 
limits for the control room. The LPZ increases in dose consequence 
are the result of using more conservative assumptions in the revised 
analyses and the new values remain a small fraction of the 10 CFR 
100 limits. The WGTR dose calculation is not affected by this TSCR. 
The proposed Technical Specification changes ensure that the MHA and 
WGTR accident analysis parameters remain bounded during plant 
operation.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any previously evaluated. This TSCR does not 
involve any physical modifications which would affect structures, 
systems, or components, nor does it involve any changes in plant 
operation. The only changes resulting from this TSCR are revisions 
to leakage limits and testing requirements necessary to reflect the 
revised MHA analysis and to correct discrepancies identified by the 
NRC .... Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. This TSCR does not involve changes to Technical 
Specification defined Safety Limits, Limiting Conditions for 
Operation, and does not involve any change to safety system 
setpoints for operation. Therefore, the proposed change does not 
involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Ronald B. Eaton (Acting)

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

    Date of amendments request:  August 12, 1997
    Description of amendments request: The amendment requests changes 
to the Surveillance Specification of the Technical Specification (TS) 
for the once through steam generator (OTSG) inservice inspection for 
TMI-1 Cycle 12 Refueling (12R) examinations applicable to TMI-1 Cycle 
12 operation. These proposed changes impose axial and circumferential 
extent sizing limitations in addition to TS requirements for

[[Page 45459]]

inside diameter (ID) initiated degradation where bobbin coil eddy 
current test (ECT) signal amplitudes do not permit reliable through 
wall sizing. Editorial changes are being made to improve consistency of 
format, to the Bases which relate to the requested changes in Section 
4.19 of the TS, and to the reporting requirements in Section 4.19.5 of 
the 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:
    GPU Nuclear has determined that this TSCR [Technical 
Specification Change Request] poses no significant hazards 
consideration as defined by 10 CFR 50.92.
    A. These proposed changes do not represent a significant 
increase in the probability of occurrence or consequences of an 
accident previously evaluated. The only accidents previously 
evaluated that could be significantly affected by changes to the 
OTSG tube inservice inspection requirements are the steam generator 
tube rupture (STGR) and the main steam line break (MSLB) accidents.
    The proposed flaw disposition strategy based on measurable eddy 
current parameters of axial and circumferential extent for Inside 
Diameter (ID) Initiated Inter-Granular Attack (IGA) will provide 
high confidence that unacceptable flaws that do not have the 
required structural integrity to withstand the MSLB are removed from 
service. The proposed axial and circumferential length limits for 
eddy current inside diameter degradation indications meet the RG 
[Regulatory Guide] 1.121 acceptance criteria for margin to failure 
for MSLB applied differential pressure and axial tube loads. The 
capability for detection of flaws is unaffected and the 
identification of tubes which should be repaired or removed from 
service is maintained or improved. The operation of the OTSG or 
related structures, systems, or components is otherwise unaffected. 
Therefore, neither the probability nor consequences of a SGTR is 
significantly increased either during normal operation or due to the 
limiting loads of [an] MSLB accident.
    Neither the editorial changes in format, punctuation, or grammar 
nor the administrative changes or changes in reporting requirements, 
as described above, could significantly affect the probability of 
occurrence or consequences of any accident previously evaluated.
    B. These proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
because there are no hardware changes involved nor changes to any 
operating practices. These changes involve only the OTSG tube 
inservice inspection surveillance requirements, which could only 
affect the potential for OTSG primary-to-secondary leakage. The 
proposed changes impose additional flaw length limits for ID IGA 
that go beyond existing requirements to assure tube structural and 
leakage integrity.
    In addition, neither the editorial changes in format, 
punctuation, or grammar nor the administrative changes, as described 
above, could possibly create the possibility of an accident of a new 
or different type from any previously evaluated. These changes are 
included only to improve the clarity and readability of the 
Technical Specifications and comply with the NRC's desire to obtain 
the results of the inspections as soon as practical.
    Therefore, these changes do not create the potential for single 
or multiple tube ruptures or any other kind of accident different 
from those that have been evaluated.
    C. Those proposed changes do not involve a significant reduction 
in a margin of safety because the changes are more restrictive than 
the current technical specification and the margins of safety 
defined in R.G. 1.121 are retained. The probability of detecting 
degradation is unchanged since the bobbin coil eddy current methods 
will continue to be the primary means of initial detection and the 
probability of leakage from any indications left in service remains 
acceptable small. The strategy for dispositioning ID initiated IGA 
will continue to provide a high level of confidence that tubes 
exceeding the allowable limits for tube integrity are repaired or 
removed from service.
    In addition, neither the editorial changes in format, 
punctuation, or grammar nor the administrative changes or changes in 
reporting requirements, as described above, could significantly 
affect a margin of safety and are included only to improve the 
clarity and readability of the Technical Specifications and comply 
with the NRC's desire to obtain the results from tube inspections as 
soon as practical.
    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 
amendments request involves no significant hazards consideration.
    Local Public Document location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Ronald B. Eaton, Acting

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

    Date of amendment request: August 14, 1997
    Description of amendment request: The proposed license amendment, 
if approved, would revise the TMI-1 Updated Final Safety Analysis 
Report (UFSAR) Section 14.1.2.9-Steam Line Break analysis to include 
the environmental dose consequences associated with postulated 
accident-induced steam generator tube leakage not previously analyzed. 
The revised environmental dose consequences for the TMI-1 Steam Line 
Break analysis would be increased above the values previously reviewed 
by the NRC.
    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:
    GPU Nuclear has determined that this License Amendment Request 
poses no significant hazards as defined by 10 CFR 50.92.
    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated. This change has no effect on structures, 
systems or components prior to the postulated steam line break 
accident or any other accident. OTSG [once through steam generator] 
tube loads resulting from other postulated accidents are bounded by 
the calculated steam line break accident tube loads. Other TMI-1 
design basis accidents, which could result in OTSG tube loads and 
environmental dose consequences, involve releases within the reactor 
building. These events generally result in rapid depressurization of 
the primary system which minimizes the differential pressure needed 
to establish a significant primary-to-secondary leak rate and the 
OTSG is isolated. Accordingly, leakage to the environment as a 
result of induced tube loads from postulated accidents other than 
steam line break is insignificant and therefore need not be 
considered. The existing steam line break criteria is maintained in 
that OTSG structural integrity is assured and postulated doses 
remain within 10 CFR 100 limits. The new radiological consequences 
of the revised steam line break dose calculation are below 10 CFR 
100 limits for the exclusion area boundary (EAB) and low population 
zone (LPZ). The 10 CFR 50, Appendix A, GDC [General Design 
Criterion]-19 limits for the control room are not affected by this 
change since the source term assumed for the TMI-1 control room 
habitability analysis remains bounding.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any previously evaluated. This change has no 
impact on any plant structures, systems or components. OTSG tube 
structural integrity is maintained. The only impact is the revised 
radiological consequences of the steam line break analysis to 
account for hypothetical accident induced primary-to-secondary 
leakage.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. This change to the steam line break

[[Page 45460]]

dose consequences does not involve a significant reduction in a 
margin of safety. The new radiological consequences of the revised 
steam line break dose calculation are below 10 CFR 100 limits for 
the EAB and LPZ, and do not affect the TMI-1 control room 
habitability analysis results. This change has no impact on any 
structures, systems or components.
    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 location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Ronald B. Eaton, Acting

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

    Date of amendment request: July 31, 1997
    Description of amendment request: The proposed amendment would 
change Action Statement 36 of Technical Specification (TS) Table 3.3.3-
1, ``Emergency Core Cooling System Actuation Instrumentation,'' so as 
to specify actions to be taken if one or more channels per trip 
function should be inoperable in the high-pressure core spray (HPCS) 
drywell pressure and reactor water level instrumentation. Presently, 
Action 36 only addresses actions for the plant condition of having one 
channel per trip function inoperable. Specifically, Action 36 would be 
changed to require that, with the number of operable channels less than 
required by the minimum operable channels per trip function 
requirement, then (1) with one channel inoperable, the inoperable 
channel is to be placed in the tripped condition within 24 hours or the 
HPCS system is to be declared inoperable, and (2) with more than one 
channel inoperable, the HPCS system is to be declared inoperable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The changes to Table 3.3.3-1, Action 36, will allow Action 36 to 
be in effect for the plant condition where more than one channel is 
inoperable per trip function in the HPCS drywell pressure and 
reactor water level instrumentation and will clarify the actions 
required if more than one channel is inoperable. Specifically, this 
action statement will allow the HPCS to be declared inoperable 
rather than to initiate plant shutdown per TS 3.0.3. None of the 
precursors of previously evaluated accidents are affected and 
therefore, the probability of an accident previously evaluated is 
not increased.
    The HPCS system will continue to perform its safety function to 
automatically initiate and inject water into the vessel. The out of 
service time for the initiating instruments remains bounded by the 
out of service time for HPCS. Therefore, these changes will not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    2. The operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The changes to Table 3.3.3-1, Action 36, will allow Action 36 to 
be in effect for plant conditions where more than one channel is 
inoperable per trip function in the HPCS drywell pressure and 
reactor water level instrumentation and will clarify the actions 
required if more than one channel is inoperable. No physical 
modification of the plant is involved and no changes to the methods 
in which plant systems are operated are required. The changes do not 
introduce any new failure modes or conditions that may create a new 
or different accident. Therefore, the changes do not by themselves 
create the possibility of a new or different kind of accident [from 
any accident] previously evaluated.
    3. The operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not involve a significant reduction in 
a margin of safety.
    The change to Table 3.3.3-1, Action 36, will allow Action 36 to 
be in effect for plant conditions where more than one channel is 
inoperable per trip function in the HPCS drywell pressure and 
reactor water level instrumentation and will clarify the actions 
required if more than one channel is inoperable. The changes do not 
adversely affect any physical barrier to the release of radiation to 
plant personnel or to the public. The proposed change provides 
consistency between the ECCS [emergency core cooling system] 
instrumentation and system TS. The TS also continues to require the 
operability of other injection systems coincidental with HPCS 
inoperability. The change has the benefit of avoiding unnecessary 
challenges to plant systems during an unnecessary plant shutdown. 
Therefore, the changes do not involve a significant reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document location: Reference and Documents Department, 
Penfield Library, State University of New York, Oswego, New York 13126
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: Alexander W. Dromerick, Acting Director

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

    Date of amendment request: April 14, 1997
    Description of amendment request: The proposed amendment would 
allow the Safety Review Committee (SRC) to perform a review, rather 
than an audit, of plant staff performance. The proposed amendment also 
involves a title change.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously analyzed?
    Response:
    This amendment application does not involve a significant 
increase in the probability or consequences of an accident 
previously analyzed. The proposed changes allow the SRC to perform a 
review, rather than an audit, of plant staff performance. This 
change does not diminish the SRCs effectiveness. A review of the 
1995 QA [quality assurance] audit of plant staff performance shows 
that no findings were issued. This indicates that the other review 
mechanisms currently in place are sufficient to ensure that plant 
staff performance is monitored.
    The position title change is an administrative change as all 
previously performed functions are being maintained and the 
responsibilities and reporting chain for this position remain the 
same. Therefore, the proposed changes do not affect the probability 
or consequences of any previously analyzed accident.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response:
    This amendment application does not create the possibility of a 
new or different

[[Page 45461]]

kind of accident from any accident previously evaluated. The 
proposed changes affect an SRC audit requirement and a position 
title. These changes do not affect plant equipment or the way the 
plant operates. Therefore, they cannot create a new or different 
kind of accident.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response:
    This amendment application does not involve a significant 
reduction in a margin of safety. The requested Technical 
Specification revisions require the SRC to review rather than audit 
facility staff performance and will not diminish the effectiveness 
of the SRC. A review of the 1995 audit confirms that performance of 
the annual audit is redundant as no findings or recommendations 
concerning plant staff performance were made. The QA/ORG [Operations 
Review Group] quarterly trend reports and SRC review of plant staff 
performance are adequate to ensure that plant staff performance is 
properly monitored.
    The position title change is an administrative change as all 
previously performed functions are being maintained and the 
responsibilities and reporting chain for this position remain the 
same. Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document location: White Plains Public Library, 100 
Martine Avenue, White Plains, New York 10601
    Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle, 
New York, New York 10019
    NRC Project Director: Alexander W. Dromerick, Acting

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

    Date of amendment request: May 29, 1997
    Description of amendment request: The amendment would revise the 
definition of Containment Integrity in Section 1.10, and revise Section 
3.6 and Table 3.6-1 for consistency. Several valves would be added to 
Table 3.6-1 to be consistent with the revised definition in Section 
1.10. The amendment would also add a footnote stating that valves SP-
SOV-506 and SP-SOV-507 in Table 4.4-1, ``Containment Isolation Valves'' 
are sealed from weld channel and containment penetration pressurization 
system (WCCPPS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The revision of the definition of containment integrity in 
Section 1.10, Section 3.6.A.1, the Basis, and the addition of 
existing containment isolation valves into the Table of Containment 
Isolation Valves in the Technical Specifications does not change the 
design, operation or testing of the plant. Section 1.10 is being 
revised to clearly cover all non-automatic containment isolation 
valves, and the valves are being added to be consistent with the 
revised definition. The valves being added are currently identified 
as containment isolation valves and tested as specified in the Final 
Safety Analysis Report. Additionally, valves CB-3, 4, 7 & 8 are 
controlled in accordance with Section 1.10.5 (revised numbering) for 
the airlock doors. Because the design and operation are not being 
changed, the addition of the valves has no effect on the probability 
or consequences of an accident.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    Changing the definition in Section 1.10 and the list of 
containment isolation valves for consistency does not change the 
design, operation or testing of the plant. Section 1.10 is being 
revised to clearly cover all non-automatic containment isolation 
valves, and the valves are being added to be consistent with the 
revised definition. The valves being added are currently identified 
as containment isolation valves and tested as specified in the Final 
Safety Analysis Report. Therefore, without changing design, 
operation or testing of the plant this does not create a new or 
different type of accident.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The proposed changes in the definition for containment integrity 
and the listings of Containment Isolation Valves in the Technical 
Specifications does not involve a significant reduction in the 
margin of safety because the change reflects current design, 
operation and testing of the plant, and will not alter plant 
operation.
    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 location: White Plains Public Library, 100 
Martine Avenue, White Plains, New York 10601
    Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle, 
New York, New York 10019
    NRC Project Director: Alexander W. Dromerick, Acting

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

    Date of amendment request: June 25, 1997
    Description of amendment request: The proposed amendment would 
allow for up to +17/-12 steps of control rod misalignment for core 
power greater than 85% rated thermal power.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response:
    No. Based on the Westinghouse evaluation in WCAP-14668, the 
Authority has determined that all pertinent licensing basis 
acceptance criteria have been met, and the margin of safety as 
defined in the TS [technical specification] Bases is not reduced in 
any of the IP3 licensing basis accident analysis (even for 
misalignments to [plus or minus] 24 steps for core power [less than 
or equal to] 85% of RTP). Increasing the magnitude of allowed 
control rod indicated misalignment is not a contributor to the 
mechanistic cause of an accident evaluated in the FSAR [final safety 
analysis report]. Neither the rod control system nor the rod 
position indicator function is being altered. Therefore, the 
probability of an accident previously evaluated has not 
significantly increased. Because design limitations continue to be 
met, and the integrity of the reactor coolant system pressure 
boundary is not challenged, the assumptions employed in the 
calculation of the offsite radiological doses remain valid. 
Therefore, the consequences of an accident previously evaluated will 
not be significantly increased.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response:
    No. Based on the Westinghouse evaluation in WCAP-14668, the 
Authority has determined that all pertinent licensing basis 
acceptance criteria have been met, and the margin of safety as 
defined in the TS is not reduced in any of the IP3 licensing basis 
accident analysis. Increasing the magnitude of allowed control rod 
indicated misalignment is not a contributor to the mechanistic cause 
of any accident. Neither the rod control system nor the rod position 
indicator function is being altered. Therefore, an accident which is 
new or different than any previously evaluated will not be created.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response:
    No. Based on the Westinghouse evaluation in WCAP-14668, the 
Authority has determined that all pertinent licensing basis

[[Page 45462]]

acceptance criteria have been met, and the margin of safety as 
defined in the TS Bases is not reduced in any of the IP3 [Indian 
Point Unit 3] licensing basis accident analysis based on the changes 
to safety analyses input parameter values as discussed in WCAP-
14668. Since the evaluations in Section 3.0 of WCAP-14668 
demonstrate that all applicable acceptance criteria continue to be 
met, 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document location: White Plains Public Library, 100 
Martine Avenue, White Plains, New York 10601
    Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle, 
New York, New York 10019
    NRC Project Director: Alexander W. Dromerick, Acting

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

    Date of amendment request: June 19, 1997, as supplemented by 
letters dated July 30 and 31, 1997
    Description of amendment request: The proposed amendment would 
provide changes to Technical Specification (TS) 4.1.3.1.2, ``Control 
Rod Operability,'' TS 3.1.3.6, ``Control Rod Drive Coupling,'' TS 
3.1.3.7, ``Control Rod Position Indication'', TS 3.1.4.1, ``Rod Worth 
Minimizer,'' TS 3/4.1.4.2, ``Rod Sequence Control System,'' TS 3/
4.10.2, ``Special Test Exceptions - Rod Sequence Control System,'' the 
Bases for TS 2.2.1.2, ``Average Power Range Monitor,'' the Bases for TS 
3/4.1.4, ``Control Rod Program Controls,'' and the Bases for TS 3/
4.10.2, ``Rod Sequence Control System.'' The changes are proposed in 
order to eliminate the Rod Sequence Control System (RSCS) Limiting 
Condition for Operation and Surveillance Requirements from the TSs and 
reduce the Rod Worth Minimizer (RWM) low power setpoint from 20% to 
10%. Changes are also proposed as necessary to delete reference to the 
RSCS from the TSs and to incorporate additional requirements necessary 
to support the elimination of the RSCS.
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    A. RSCS Deletion
    The RSCS system restricts the pattern of control rods prior to a 
postulated control rod drop accident (RDA) so as to minimize the 
reactivity worth of the dropped rod. The RSCS provides no mitigation 
following the postulated RDA. The ability to restrict the pattern of 
control rods also allows the RSCS to be able to reduce the 
probability of a Continuous Rod Withdrawal During Reactor Startup, 
as described in the Hope Creek UFSAR [Updated Final Safety Analysis 
Report] Section 15.4.1.2 and Appendix 15B. However, to determine the 
consequence of such a rod withdrawal event, the RSCS is not 
credited, and the rod is assumed to be fully withdrawn from the core 
at its maximum rate. The RDA is therefore the only analyzed accident 
impacted by the proposed deletion of the RSCS system. Since the RSCS 
system plays no role in preventing a[n] RDA, it therefore does not 
affect the probability of occurrence of this postulated accident.
    As stated in an NRC Safety Evaluation Report dated December 27, 
1987, the RSCS system is the result of requirements promulgated by 
the NRC staff in the early 1970's in response to unknowns and 
perceived problems relating to the RDA. The GE [General Electric] 
calculational methodology being used at that time produced results 
showing that, even without pattern errors, calculated enthalpies for 
the RDA approached limiting values. In addition, the Rod Worth 
Minimizer (RWM) Technical Specifications were not effective in 
ensuring RWM availability and use, and the system was poorly 
maintained and frequently bypassed thus providing no significant 
protection. Second operator substitution for the RWM was used 
routinely and was providing minimal protection. Finally, no reliable 
study existed to address the probability of exceeding enthalpy 
limits as a result of an RDA.
    Information associated with the above concerns has been 
significantly expanded or modified. Studies using improved 
methodologies have proven significantly lower peak fuel enthalpy 
values compared with methodologies in use when the RSCS was 
originally developed. In addition, a reliable probability study has 
been completed showing that the probability of an RDA exceeding NRC 
limits is very low. As a result, NRC review of the RSCS requirements 
has concluded that the RSCS system is not needed and operation 
without it is acceptable provided: 1) TSs are modified to minimize 
the use of the second operator option, 2) procedures and quality 
control associated with the second operator option are reviewed to 
ensure that this option provides an effective and truly independent 
monitoring process; and 3) rod patterns used are at least equivalent 
to Banked Pattern Withdrawal System (BPWS) patterns. Each of these 
items has been addressed for the Hope Creek Generating Station.
    As a result of the resolution of the original concerns 
associated with the RDA, the RWM system and limited use of the 
second operator option, when properly instituted, are now deemed to 
provide adequate protection to maintain the consequences of the RDA 
at an acceptable level. The remaining concerns regarding operation 
without the RSCS system and proper use of the second operator 
substitution option have been addressed for the Hope Creek 
Generating Station. We therefore conclude that the redundant RSCS 
system is no longer necessary and its deletion from the Technical 
Specifications will not significantly increase the probability or 
consequences of an RDA.
    B. RWM Setpoint Reduction
    The RWM system restricts the pattern of control rods prior to a 
postulated control rod drop accident (RDA) so as to minimize the 
reactivity worth of the dropped rod. The RWM provides no mitigation 
following the postulated RDA. The ability to restrict the pattern of 
control rods also allows the RWM to be able to reduce the 
probability of a Continuous Rod Withdrawal During Reactor Startup, 
as described in the Hope Creek UFSAR Section 15.4.1.2 and Appendix 
15B. However, to determine the consequence of such a rod withdrawal 
event, the RWM is not credited, and the rod is assumed to be fully 
withdrawn from the core at its maximum rate. The RDA is therefore 
the only analyzed accident impacted by the proposed reduction in the 
RWM setpoint. Since the RWM system plays no role in preventing a[n] 
RDA, it therefore does not affect the probability of occurrence of 
this postulated accident.
    Existing calculations have demonstrated that no significant RDA 
can occur above 10% power. Calculations by both General Electric and 
the Brookhaven National Laboratory indicate that, even with 
significant error patterns, peak fuel enthalpy is reduced well below 
required limits at 10% power. The 20% limit was originally required 
as an extreme bound because of the then existing uncertainties in 
the analyses. Based on the current analyses, the 10% level is now 
acceptable and deemed to provide adequate protection to maintain the 
consequences of an RDA at an acceptable level. Changing the RWM 
setpoint from 20% to 10% will therefore not significantly increase 
the consequences of any previously analyzed accident.
    2. Do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    A. RSCS Deletion
    Operation of the RSCS cannot cause or prevent an accident; this 
system functions to minimize the consequences of an RDA. The Bank 
Position Withdrawal Sequence (BPWS) will still be used to ensure 
that rod pull pattern[s] are constrained to those assumed in the 
RDA. The RSCS has no impact on the operation of any other system, 
and therefore its deletion will not contribute to a malfunction in 
any other equipment nor create the possibility of a new or different 
accident from any accident previously evaluated.
    B. RWM Setpoint Reduction
    Operation of the RWM cannot cause or prevent an accident; this 
system functions to minimize the consequences of an RDA. The RWM has 
no impact on the operation of any

[[Page 45463]]

other system, and therefore changing its setpoint from 20% to 10% 
will not contribute to a malfunction in any other equipment nor 
create the possibility of a new or different accident from any 
accident previously evaluated.
    3. Do not involve a significant reduction in a margin of safety.
    A. RSCS Deletion
    When the original decisions were made regarding the need for the 
RSCS system, numerous perceived problems in the RDA analysis 
existed. As noted in the discussion of the consequences of 
previously analyzed accidents in Item 1 above: 1) the perceived RDA 
problems have been resolved; 2) reviews of the RDA have concluded 
that the RSCS is not needed to mitigate the consequences of an RDA; 
and 3) operation without the RSCS is acceptable. The RWM and limited 
use of second operator substitution, when properly instituted, are 
now deemed adequate to ensure that peak fuel enthalpies remain below 
NRC limits. Therefore, the deletion of the redundant RSCS system 
will not significantly decrease any margin of safety.
    B. RWM Setpoint Reduction
    The Bases for the HCGS TSs state that when thermal power is 
greater than 20%, there is no possible rod worth that, if dropped at 
the design rate of the velocity limiter, could result in a peak 
enthalpy of 280 calories per gram. Existing calculations demonstrate 
that the RDA is not a significant concern above 10% power, and 
therefore, a mitigation system is not needed for higher power level 
operation. Calculations by both General Electric and the Brookhaven 
National Laboratory indicate that, even with significant error 
patterns, peak fuel enthalpy is reduced well below required limits 
(280 calories per gram) at 10% power. The 20% limit was originally 
required as an extreme bound because of the then existing 
uncertainties in the analyses. Based on the current analyses, the 
10% level is now acceptable and deemed to provide adequate assurance 
that the peak fuel enthalpy will remain below the NRC limits during 
a postulated RDA. Changing the RWM setpoint from 20% to 10% will 
therefore not significantly reduce any 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 location: Pennsville Public Library, 190 S. 
Broadway, Pennsville, New Jersey 08070
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit - N21, P. O. Box 236, Hancocks Bridge, New Jersey 08038
    NRC Project Director: John F. Stolz

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

    Date of amendments request: June 30, 1997
    Description of amendments request: The proposed amendments would 
change the Farley Technical Specifications to: revise and clarify the 
requirements for the Control Room Emergency Filtration System (CREFS), 
the Penetration Room Filtration System (PRFS) and the related Storage 
Pool Ventilation System (SPVS); revise the required number of radiation 
monitoring instrumentation channels; and delete the Containment Purge 
Exhaust Filter (CPEF) specification.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Pursuant to 10 CFR 50.92, SNC [Southern Nuclear Operating 
Company, Inc.] has evaluated the proposed amendments and has 
determined that operation of the facility in accordance with the 
proposed amendments would not involve a significant hazards 
consideration. The basis for this determination is as follows:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes to convert from ANSI N510-1980 to ASME 
N510-1989 for specific FNP [Joseph M. Farley Nuclear Plant] 
filtration surveillance testing requirements and related changes do 
not affect the probability of any accident occurring. The 
consequences of any accident will not be affected since the proposed 
changes will continue to ensure that appropriate and required 
surveillance testing for FNP filtration systems will be performed 
consistent with the revised accident analyses. The results of the 
fuel handling accident remain well within the guidelines of I0 CFR 
Part 100 and the doses due to a LOCA [loss-of-coolant accident], 
including ECCS [emergency core cooling system] recirculation loop 
leakage, remain within the guidelines of I0 CFR Part 100 and General 
Design Criterion 19 of Appendix A to I0 CFR Part 50. Relocating 
specific testing requirements to the FNP FSAR [Final Safety Analysis 
Report] has no effect on the probability or consequences of any 
accident previously evaluated since required testing will continue 
to be performed.
    Therefore, the proposed TS [Technical Specification] 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.
    Testing differences between ANSI N510-1980 and ASME N510-1989 
have been evaluated by SNC and none of the proposed changes have the 
potential to create an accident at FNP. ASME N510-1989 has been 
endorsed and approved by the NRC for licensee use in NUREG 1431 
[Standard Technical Specifications Westinghouse Plants]. Testing the 
additional channels of radiation monitoring and verification of 
penetration room boundary integrity do not require the affected 
systems to be placed in configurations different from design. Thus, 
no new system design or testing configuration is required for the 
changes being proposed that could create the possibility of any new 
or different kind of accident from any accident previously 
evaluated. Relocating specific testing requirements to the FSAR has 
no effect on the possibility of creating a new or different kind of 
accident from any accident previously evaluated since it is an 
administrative change in nature.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    Conversion from the testing requirements of ANSI N510-1980 
sections 10, 12, and 13 to ASME N510-1989 sections 10, 11, and 15 
has been previously approved by the NRC at other nuclear facilities. 
ASME N510-1989 has been approved and endorsed by the NRC in NUREG 
1431. The safety factor associated with the conservative charcoal 
adsorber laboratory test methods and dose calculations ensures that 
doses will continue to meet the guidelines of 10 CFR Part 100 and 
GDC [General Design Criterion] 19 of Appendix A to 10 CFR Part 50. 
The enhanced testing of radiation monitoring instrumentation and the 
penetration room boundary integrity provide additional assurance 
that the acceptance criteria of the safety analyses and the 
resultant margins of safety are not reduced. Relocating specific 
testing requirements to the FSAR has no effect on the margin of 
plant safety since required testing will continue to be performed. 
Clarifying the 10 hour run with heaters on is consistent with the 
Improved TS language and accomplishes the purpose for the 
surveillance. Therefore, SNC concludes based on the above, that the 
proposed changes do not result in a significant reduction of margin 
with respect to plant safety as defined in the Final Safety Analysis 
Report or the bases of the FNP technical specifications.
    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 location: Houston-Love Memorial Library, 212 
W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302

[[Page 45464]]

    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: Herbert N. Berkow

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

    Date of amendments request: June 30, 1997
    Description of amendments request: The proposed amendments would 
change the Farley Technical Specifications to incorporate the 
requirements necessary to change the basis for prevention of 
criticality in the fuel storage pool. This change eliminates the need 
for Boraflex as a neutron absorbing material in the fuel pool 
criticality analysis for both Unit 1 and Unit 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    There is no significant increase in the probability of a fuel 
assembly drop accident in the spent fuel pool when considering the 
presence of soluble boron in the spent fuel pool water for 
criticality control. The handling of the fuel assemblies in the 
spent fuel pool has always been performed in borated water.
    The consequences of a fuel assembly drop accident in the spent 
fuel pool are not affected when considering the presence of soluble 
boron.
    Although the probability of misloading an assembly in the spent 
fuel racks may increase due to new assembly placement constraints, 
there is no significant increase in the probability of an accidental 
misloading of spent fuel assemblies into the spent fuel pool racks 
that will cause a criticality accident when considering the presence 
of soluble boron in the pool water for criticality control. 
Sufficient soluble boron will be maintained in the spent fuel pool 
to maintain keff below 0.95 following a postulated single 
misload. Fuel assembly placement will continue to be controlled 
pursuant to approved fuel handling procedures and will be in 
accordance with the Technical Specification spent fuel rack storage 
configuration limitations. The addition of the spent fuel pool 
storage configuration surveillance in proposed new Technical 
Specifications 3.7.14 for Unit 1 and 3.7.15 for Unit 2 will provide 
increased assurance that a spent fuel pool inventory verification 
will be completed in a timely manner (7 days) after the relocation 
or addition of fuel assemblies in the spent fuel storage pool.
    There is no significant increase in the consequences of the 
accidental misloading of spent fuel assemblies into the spent fuel 
pool racks because criticality analyses demonstrate that the pool 
will remain subcritical following an accidental misloading if the 
pool contains an adequate boron concentration. The proposed new 
Technical Specifications limitations will ensure that an adequate 
spent fuel pool boron concentration will be maintained.
    In the event of failure of a spent fuel pool cooling pump, or 
loss of cooling to a spent fuel pool heat exchanger, the second 
spent fuel pool cooling train provides 100 percent backup 
capability, thus ensuring continued cooling of the spent fuel pool. 
However, even if a loss of spent fuel pool cooling were to occur, 
there is sufficient soluble boron to prevent Keff from 
exceeding 0.95.
    There is no significant increase in the probability of the loss 
of normal cooling to the spent fuel pool water when considering the 
presence of soluble boron in the pool water for subcriticality 
control since a high concentration of soluble boron has always been 
maintained in the spent fuel pool water.
    A loss of normal cooling to the spent fuel pool water causes an 
increase in the temperature of the water passing through the stored 
fuel assemblies. This causes a decrease in water density which would 
result in a decrease in reactivity when Boraflex neutron absorber 
panels are present in the racks.
    However, since Boraflex is not considered to be present, and the 
spent fuel pool water has a high concentration of boron, a density 
decrease causes a positive reactivity addition. However, the 
additional negative reactivity provided by the proposed 2000 ppm 
boron concentration limit, above that provided by the concentration 
required to maintain Keff less than or equal to 0.95 (400 
ppm), will compensate for the increased reactivity which could 
result from a loss of spent fuel pool cooling event. Because 
adequate soluble boron will be maintained in the spent fuel pool 
water, there is no significant increase in the consequences of a 
loss of normal cooling to the spent fuel pool.
    Therefore, based on the conclusions of the above analysis, the 
proposed changes will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    Spent fuel handling accidents are not new or different types of 
accidents, they have been analyzed in Section 15.4.5 of the Final 
Safety Analysis Report.
    Criticality accidents in the spent fuel pool are not new or 
different types of accidents, they have been analyzed in the Final 
Safety Analysis Report and in Criticality Analysis reports 
associated with specific licensing amendments for fuel enrichments 
up to 5.0 weight percent U-235.
    Proposed new Technical Specifications 3.7.13 for Unit 1 and 
3.7.14 for Unit 2 on the spent fuel pool boron concentration do not 
represent new concepts. The boron concentration in the spent fuel 
pool has always been maintained near at the limit of the RWST 
[refueling water storage tank] boron concentration for refueling 
purposes. These new proposed Technical Specifications establish new 
boron concentration requirements for the spent fuel pool water 
consistent with the results of the revised criticality analysis [ ].
    Since soluble boron has always been maintained in the spent fuel 
pool water, the implementation of this new requirement will have 
little effect on normal pool operations and maintenance. The 
implementation of the proposed new limitations on the spent fuel 
pool boron concentration will only result in increased sampling to 
verify boron concentration. This increased sampling will not create 
the possibility of a new or different kind of accident.
    Because soluble boron has always been present in the spent fuel 
pool, a dilution of the spent fuel pool soluble boron has always 
been a possibility. However, it was shown in the spent fuel pool 
dilution evaluation [ ] that a dilution of the Farley spent fuel 
pool which could reduce the spent fuel storage rack Keff 
to less than 0.95 is not a credible event. Therefore, the 
implementation of new limitations on the spent fuel pool boron 
concentration will not result in the possibility of a new kind of 
accident.
    Proposed new Technical Specifications 3.7.14 for Unit 1 and 
3.7.15 for Unit 2, and 5.6.1.1.e., 5.6.1.1.f, and 5.6.1.1.g. (for 
Unit 1) specify the requirements for the spent fuel rack storage 
configurations, and do not represent new concepts. These proposed 
new spent fuel pool storage configuration limitations are consistent 
with the assumptions made in the spent fuel rack criticality 
analysis, and will not have any significant effect on normal spent 
fuel pool operations and maintenance and will not create any 
possibility of a new or different kind of accident. Verifications 
will continue to be performed to ensure that the spent fuel pool 
loading configuration meets specified requirements.
    As discussed above, the proposed changes will not create the 
possibility of a new or different kind of accident. There is no 
significant change in plant configuration, equipment design or 
equipment. The accident analysis in the Final Safety Analysis Report 
remains bounding.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The proposed Technical Specification changes and the resulting 
spent fuel storage operating limits will provide adequate safety 
margin to ensure that the stored fuel assembly array will always 
remain subcritical. Those limits are based on a plant specific 
criticality analysis [ ] performed in accordance the Westinghouse 
spent fuel rack criticality analysis methodology described in [WCAP-
14416-NP-A, ``Westinghouse Spent Fuel Rack Criticality Analysis 
Methodology,'' Revision 1, November 1996].
    The criticality analysis utilized credit for soluble boron to 
ensure Keff will be less than or equal to 0.95 under 
normal circumstances, and storage configurations have been defined 
using a 95/95 Keff calculation to ensure that the spent 
fuel rack Keff will be less than 1.0 with no soluble 
boron.

[[Page 45465]]

    Soluble boron credit is used to provide safety margin by 
maintaining Keff less than or equal to 0.95, including 
uncertainties, tolerances, and accident conditions in the presence 
of spent fuel pool soluble boron.
    The loss of substantial amounts of soluble boron from the spent 
fuel pool which could lead to exceeding a Keff of 0.95 
has been evaluated [ ] and shown to be not credible.
    The evaluations which...show that the dilution of the spent fuel 
pool boron concentration from 2000 ppm to 400 ppm is not credible, 
combined with the 95/95 calculation, which shows that the spent fuel 
rack Keff remain less than 1.0 when flooded with 
unborated water, provide a level of safety comparable to the 
conservative criticality analysis methodology required by [USNRC 
Standard Review Plan for the Review of Safety Analysis Reports for 
Nuclear Power Plants, LWR Edition, NUREG-0800, June 1987, USNRC 
Spent Fuel Storage Facility Design Bases (for comment) Proposed 
Revision 2, 1981, Regulatory Guide 1.13, and ANS, Design 
Requirements for Light Water Reactor Spent Fuel Storage Facilities 
at Nuclear Power Stations, ANSI/ANS-57.2-1983].
    Therefore, the proposed changes in this license amendment will 
not result in a significant reduction in the plant's 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 location: Houston-Love Memorial Library, 212 
W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: July 11, 1997
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) to implement 10 CFR Part 50, 
Appendix J, Option B, by referring to Regulatory Guide 1.163, 
``Performance-Based Containment Leak-Test Program,'' with four 
exceptions as detailed in the licensee's application. Specifically, 
changes are requested for TSs 3.7/4.7, STATION CONTAINMENT SYSTEMS, 
their associated BASES, and changes to TS Table 4.7.2. Included in the 
above changes is a revision to the conservative wording of Surveillance 
Requirement (SR) 4.7.A.3 that is being replaced by wording from the 
Standard Technical Specifications, and the relocation of this SR to the 
Limiting Condition for Operation. The change to TS Table 4.7.2 updates 
the information in the Table to the current operational practices, as 
approved by an NRC letter dated May 3, 1982. In addition, a description 
of Vermont Yankee's Primary Containment Leakage Rate Testing Program 
(PCLRTP) will be added to the Administrative Controls Section (6.0) of 
the TSs. The testing intervals for the containment system and for the 
components that penetrate the primary containment, under Option B of 
Appendix J will be performance-based.
    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:
    Option B
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any of the parameters or conditions that contribute to 
initiation of any accidents previously evaluated. Thus, the proposed 
change cannot increase the probability of any accident previously 
evaluated.
    The proposed change potentially affects the leak-tight integrity 
of the containment structure designed to mitigate the consequences 
of a loss-of-coolant accident (LOCA). The function of the 
containment is to maintain functional integrity during and following 
the peak transient pressures and temperatures which result from any 
LOCA. The containment is designed to limit fission product leakage 
following the design basis LOCA. Because the proposed change does 
not alter the plant design or test method, only the frequency of 
measuring Type A, B and C leakage, the proposed change does not 
directly result in an increase in containment leakage. However, 
decreasing the test frequency can increase the probability that an 
increase in containment leakage could go undetected for an extended 
period of time. Based upon the results of the periodic containment 
Type A or Integrated Leak Rate Tests (ILRTs) and Type B and C or 
Local Leak Rate Tests (LLRTs) surveillance tests, this is not 
expected during the remaining life of the plant. The risk resulting 
from the proposed changes is as follows:
    Type A Testing
    NUREG/CR-4330 (NRC86) found that the effect of containment 
leakage on overall accident risk is small since risk is dominated by 
accident sequences that result in failure or bypass of the 
containment. It is also determined that on an expected individual 
dose basis, the effect of containment leakage is small.
    Industry wide, ILRTs have only found a small fraction of the 
leaks that exceed current acceptance criteria. Only three percent of 
all leaks are detected by ILRTs, and therefore, by extending Type A 
testing intervals, only three percent of all leaks have a potential 
for remaining undetected for longer periods of time. In addition, 
when leakage has been detected by ILRTs, the leakage rate has been 
only about two times the allowable leakage rate.
    NUREG-1493, ``Performance-Based Containment Leakage Test 
Program'', found that these observations, together with the 
insensitivity of reactor accident risk to the containment leakage 
rate, show that reducing the Type A leakage test frequency would 
have a minimal impact on public risk.
    Type B and C Testing
    NUREG-1493 found that while Type B and C tests can identify the 
vast majority (greater than 95 percent) of all potential leakage 
paths, performance-based alternatives are feasible without 
significant risk impacts. The risk model used in NUREG-1493 suggests 
that the number of components tested would be reduced by about 60 
percent with less than a three-fold increase in the incremental risk 
due to containment leakage. Since, under existing requirements, 
leakage contributes less than 0.1 percent of overall accident risk, 
the overall impact is very small. NUREG-1493 found that while the 
extended testing intervals for Type B and C tests led to minor 
increases in potential offsite dose consequences the actual decrease 
of on-site (worker) doses would be reduced in proportion to the 
number of Type B or C tests not performed.
    EPRI Research Project Report TR-104285, ``Risk Impact Assessment 
of Revised Containment Leak Rate Testing Intervals,'' also concluded 
that a relaxation of the test intervals for Type B and C 
penetrations results in a negligible increase in total plant risk.
    Based on the above VYNPC [Vermont Yankee Nuclear Power 
Corporation] has concluded that the proposed change will not result 
in a significant increase in the probability or consequences of any 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any of the parameters or conditions that could contribute to 
initiation of any accidents. This change involves the reduction in 
Type A, B, and C test frequency. The methods of performing the tests 
are not changed. No new accident modes are created by extending the 
testing intervals. No safety-related equipment or safety functions 
are altered as a result of this change. Extending the test frequency 
has no influence over nor does it contribute to, the possibility of 
a new or different kind of accident or malfunction from those 
previously analyzed.
    Based upon the above, VYNPC has concluded that the proposed 
change will not create the possibility of a new or different kind of 
accident from those previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    As stated in the Technical Support Document (TSD) for the NRC's 
Option B to

[[Page 45466]]

Appendix J rule change, NUREG-1493 concludes a reduction in the 
frequency of Type A testing from the current three per ten years to 
one per ten years leads to an imperceptible increase in risk. It 
also concludes that a reduction in the frequency of Type B testing 
of electrical penetrations should be possible with no adverse impact 
on risk. A vast majority of leakage paths are identified by Type C 
testing of containment isolation valves and, based on the model of 
component failure with time, performance-based alternatives to the 
current Type C testing intervals are feasible without significant 
risk impacts.
    4.7.A.3
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change does not result in any hardware or operating 
procedure changes. Closed and de-activated automatic valves, closed 
manual valves or blind flanges that serve as primary containment 
isolation valves are not assumed to be initiators of any analyzed 
event. The role of these devices is to isolate containment during 
analyzed events, thereby limiting consequences. The change 
establishes compensatory measures using closed and de-activated 
automatic valves, closed manual valves or blind flanges as an 
isolation barrier which is equivalent to those already included in 
the current Technical Specifications. The proposed change does not 
introduce any new failure modes, such that a single active failure 
could allow a primary containment release through an un-isolated 
path. Therefore, this change will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This change does not result in any changes to equipment design 
or capabilities or the operation of the plant. The change still
    ensures the primary containment boundary is maintained. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    Closed and de-activated automatic valves, closed manual valves 
or blind flanges which are used to satisfy the compensatory measures 
of 4.7.A.3 are primary containment isolation devices will be leak 
tested per the PCLRTP. In addition, the Technical Specification 
establishes these devices as an isolation barrier that cannot be 
adversely affected by a single active failure. As a result, any 
reduction in a margin of safety will be insignificant and offset by 
the benefit gained with equivalent compensatory measures to ensure 
the primary containment boundary is maintained, which reduces 
unnecessary plant shutdown transients.
    Table 4.7.2 Editorial Change
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    This change updates the information presented in this Table to 
reflect current practice. The methods of maintaining an inerted 
containment and differential pressure between the drywell and 
suppression pool have been previously docketed. The valves to now be 
shown normally closed on the Table are large (6'' and 18'') purge 
valves and the valves to be shown as normally open to provide makeup 
nitrogen are both 1'' in size. The probability of an accident is not 
significantly increased, since the subject valves are not considered 
to be initiators of any accident previously evaluated. The 
consequences of an accident are not significantly increased, since 
each of the subject valves receives a close signal from PCIS 
[primary containment isolation system]. In addition, PCIS closure of 
the two one inch valves will terminate the associated release 
pathway more rapidly than the existing valve lineup reflected on the 
Table. Thus it is concluded that this change will not involve any 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from amy previously evaluated?
    All four valves whose listed normal positions are proposed to be 
changed are PCIS valves and receive the same closing signal. All are 
tested in accordance with our Appendix J and IST [inservice testing] 
programs. No changes in equipment design or operation are proposed, 
only the listed normal positions of the subject valves. Thus, this 
change will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The valves to be listed as normally open are significantly 
smaller and faster closing than the purge valves currently listed as 
open. Thus the change in the listed normal position of these four 
valves provides a more conservative initial condition than is 
currently depicted in Table 4.7.2. No changes in equipment design or 
operation are proposed. Thus, it is concluded that 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 location: Brooks Memorial Library, 224 Main 
Street, Brattleboro, VT 05301
    Attorney for licensee: R. K. Gad, III, Ropes and Gray, One 
International Place, Boston, MA 02110-2624
    NRC Project Director: Ronald B. Eaton, Acting

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 amendment request: August 14, 1997 (TSCR 199)
    Description of amendment request: These amendments would revise: TS 
15.4.2.B. ``In-Service Inspection and Testing of Safety Class 
Components Other than Steam Generator Tubes,'' to modify item 2 to 
change the reference from TS 15.4.4 to the Containment Leakage Rate 
Testing Program; TS 15.6.12.A.1, ``Containment Leakage Rate Testing 
Program,'' to eliminate the one-time requirement for Unit 2 Type A 
testing since the testing has been completed; and TS Bases 15.4.4 to 
delete the specific bases for containment purge valve testing and to 
delete a reference that is no longer used.
    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. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not result in a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed administrative changes correct discrepancies in the 
Technical Specifications introduced as a result of Amendment 169 to 
Operating License DPR-24 for Point Beach Nuclear Plant Unit 1 and 
Amendment 173 to Operating License DPR-27 for Point Beach Nuclear 
Plant Unit 2. These changes correct references to containment 
isolation valve testing in the Specifications and Bases. These 
amendments were evaluated as acceptable in a safety evaluation dated 
October 9, 1996. Therefore, these changes do not result in an 
increase in the probability or consequences of any accident 
previously evaluated.
    The Point Beach Nuclear Plant Unit 2 containment was tested and 
found acceptable within the maximum interval defined by a one-time 
Technical Specifications requirement. Subsequent testing will be 
performed in accordance with the approved testing program defined by 
Technical Specifications 15.6.12. Therefore, the Technical 
Specification requirements are met. These requirements are 
established to ensure the containment performs and is maintained as 
designed and assumed in the safety analyses. The removal of the one-
time specific periodicity requirements for the Unit 2, Type A 
containment integrated leak rate test does not result in a 
significant increase in the probability or consequence of any 
accident previously evaluated.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes to the Technical Specifications do not 
change the requirements for the Point Beach Nuclear Plant 
containments to perform as designed and evaluated in the safety 
analyses. Test requirements in the Technical Specifications continue 
to meet the standards evaluated and approved by the NRC to ensure 
the containments continue to perform as

[[Page 45467]]

designed and analyzed. Administrative discrepancies in the 
Specifications and bases are also corrected. Therefore, no new or 
different kind of accident from any accident previously evaluated is 
created.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not involve a significant reduction in 
a margin of safety.
    The proposed changes to the Technical Specifications ensure 
consistency with Amendment 169 to Point Beach Nuclear Plant Unit 1 
Operating License DPR-24 and Amendment 173 to Point Beach Nuclear 
Plant Unit 2 Operating License DPR-27. Testing of the Unit 2 
containment has been performed within the maximum time limit allowed 
by the one-time test requirement of Technical Specification 15.6.12. 
Testing requirements continue to meet NRC requirements and ensure 
the containment continues to operate as designed and analyzed. 
Administrative corrections to the Specifications and bases ensure 
consistency with previously approved amendments. Therefore, a margin 
of safety is not 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 location: The Lester Public Library, 1001 
Adams 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: John N. Hannon

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

    Date of amendment request: July 29, 1997
    Description of amendment request: This license amendment request 
revises the wording of Action Statement 5.a to Technical Specification 
Table 3.3-1. ``Reactor Trip System Instrumentation.'' This action 
statement prescribes a set of actions to be accomplished when a source 
range neutron detector is inoperable with the plant shut down. The 
proposed wording change will clarify the times and order in which these 
actions are to be performed.
    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.
    In MODE 3, 4, or 5 with the rod control system capable of rod 
withdrawal or rods not fully inserted, the source range neutron 
detectors provide a reactor trip signal on high neutron flux to 
provide core protection against an uncontrolled rod cluster control 
assembly bank withdrawal from a subcritical or low power startup 
condition. This trip function is actuated when either of two 
independent source range channels indicates a neutron flux level 
above a preselected manually adjustable setpoint. If the
    rod control system is not capable of rod withdrawal with rods 
fully inserted, the source range detectors are not required to trip 
the reactor.
    NUREG-1431, Revision 1, ``Standard Technical Specifications 
Westinghouse Plants,'' allows one source range neutron detector to 
be out of service for up to 48 hours. One additional hour is allowed 
to open the reactor trip breakers and suspend operations involving 
the addition of positive reactivity. This was the same action 
sequence prescribed for the source range neutron detectors prior to 
the implementation of Amendment No. 96 to the Wolf Creek Technical 
Specifications, which inadvertently resulted in an ambiguous 
rewording of the action. The proposed rewording of the action 
statement clarifies the proper timing of the required actions, and 
is consistent with NUREG-1431, Revision 1.
    The proposed change does not introduce any new potential 
accident initiating conditions and does not alter any plant 
operating procedures or method of operation of any plant components 
or systems. Allowing positive reactivity changes during the 48 hour 
period in which one source range neutron detector is inoperable is 
acceptable since the remaining detector will still provide the 
reactor trip function and control room indication when the reactor 
trip breakers are closed, and control room indication
    when the reactor trip breakers are open. This is consistent with 
the provisions in NUREG-1431, Revision 1. Thus, the proposed change 
does not affect any system's ability to mitigate the consequences of 
an accident and will not increase the probability of occurrence of 
any previously evaluated accident.
    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 does not affect the method of operation of 
any plant component or system, and does not create any new, or alter 
any existing, accident initiators. The proposed change clarifies 
that positive reactivity changes may be allowed during the 48 hour 
period in which a source range neutron detector is inoperable, as 
provided for in NUREG-1431, Revision 1. This action does not affect 
the capability of the remaining source range neutron detector to 
provide a reactor trip signal on high neutron flux during this 
period when the reactor trip breakers are closed, nor does it affect 
the ability of the remaining detector of providing control room 
indication. This function of the source range neutron detectors is 
discussed in Chapter 15 of the Wolf Creek Updated Safety Analysis 
Report. This proposed change does not modify any existing plant 
equipment, add any new plant equipment, or alter any component or 
system operating parameters or procedures. Therefore, this proposed 
change will
    not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The source range neutron detectors provide a reactor trip 
function during shutdown conditions when the reactor trip breakers 
are closed. When the reactor trip breakers are open they provide 
control room alarm/indication, only. The proposed change clarifies 
that positive reactivity changes may be allowed during the 48 hour 
period in which a source range neutron detector is inoperable. This 
is consistent with the provisions in NUREG-1431, Revision 1 and with 
Wolf Creek Technical Specification Table 3.3-1, Action 5.a, prior to 
the implementation of Amendment No. 96. In Amendment No. 96 the 
wording of this action was changed such that this allowance was no 
longer clear. With one source range neutron detector inoperable with 
the reactor trip breakers closed, the reactor trip on high neutron 
flux function is still provided by the remaining source range 
neutron detector. With one source range neutron detector inoperable 
with the reactor trip breakers open, control room indication of high 
neutron flux is still provided. As stated above, this is consistent 
with NUREG-1431, Revision 1, as well as with the action requirements 
prior to the implementation of Amendment No. 96. This proposed 
change, then, does not affect the margin of safety provided by the 
source range neutron detectors.
    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 locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

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

[[Page 45468]]

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.

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

    Date of amendments request: July 25, 1997
    Brief description of amendments: The proposed amendments would 
modify Technical Specification (TS) 4.0.5.f in a manner that would 
allow exceptions to the NRC staff's positions on intergranular stress 
corrosion cracking in boiling water reactor austenitic stainless steel 
piping, where specific written relief has been granted by the NRC. TS 
4.0.5.f now requires that the Brunswick Steam Electric Plant, Units 1 
and 2, Inservice Inspection program be performed in accordance with the 
positions identified in NRC Generic Letter 88-01. Date of publication 
of individual notice in Federal Register: August 12, 1997 (62 FR 43187)
    Expiration date of individual notice: September 11, 1997
    Local Public Document location: University of North Carolina at 
Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

    Date of application for amendment: August 4, 1997
    Brief description of amendment: The proposed amendment would revise 
the Technical Specifications to extend the frequency for certain 
surveillances related to the emergency diesel generators. Date of 
publication of individual notice in the FEDERAL REGISTER:August 12, 
1997 (62 FR 43189)
    Expiration date of individual notice: September 11, 1997
    Local Public Document location: Coastal Region Library, 8619 W. 
Crystal Street, Crystal River, Florida 32629

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

    Date of amendment request: August 6, 1997
    Description of amendment request: The proposed amendment would 
revise Technical Specification Table 2.2-1 and 3/4.2.5 to allow the 
reactor coolant system total flow to be determined using cold leg elbow 
tap differential pressure measurements. Date of individual notice in 
the Federal Register: August 14, 1997 (62 FR 43556)
    Expiration date of individual notice: September 15, 1997
    Local Public Document location:  Wharton County Junior College, J. 
M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488

Notice Of Issuance Of Amendments To Facility Operating Licenses

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

Commonwealth Edison Company, Docket No. 50-455, Byron Station, Unit 
No. 2, Ogle County, Illinois, Docket No. STN 50-457, Braidwood 
Station, Unit No. 2, Will County, Illinois

    Date of application for amendments: May 24, 1997, as supplemented 
by letters dated May 31, June 20 and June 24, 1997
    Brief description of amendments: The amendments revise Technical 
Specification 4.5.2.b.1 to include the use of Ultrasonic Testing (UT) 
to verify that the emergency core cooling system (ECCS) is completely 
filled with water. For the ECCS subsystem with high point vent valves 
in direct communication with the operation system, UT is acceptable in 
lieu of physically opening the vents.
    Date of issuance: August 13, 1997
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 91 and 84
    Facility Operating License Nos. NPF-66 and NPF-77: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 10, 1997 (62 FR 
31633) The May 31, June 20, June 24, and July 18, 1997, submittals 
provided additional clarifying information that did not change the 
proposed initial no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated August 13, 1997. No significant hazards 
consideration comments received: No
    Local Public Document location: For Byron, the Byron Public Library 
District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481

Commonwealth Edison Company, Docket Nos. 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 9, 1997
    Brief description of amendments: The amendments authorize a change 
to the realistic dose values for the process gas system rupture in 
Section 15.0 of the

[[Page 45469]]

Byron/Braidwood (B/B) Updated Final Safety Analysis Report (UFSAR). 
During preparation of a UFSAR change package, ComEd discovered that the 
Final Safety Analysis Report (FSAR) had not been updated to correct an 
error from the previous revision of the dose calculation. Since the 
correct dose value is greater than that previously reported, the 
consequences of the accident had increased, and an unreviewed safety 
question resulted.
    Date of issuance: August 13, 1997
    Effective date: August 13, 1997
    Amendment Nos.: 92, 92, 85, 85
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments authorize a change to the Byron/Braidwood UFSAR.
    Date of initial notice in Federal Register: July 10, 1997 (62 FR 
37079). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 13, 1997. No significant 
hazards consideration comments received: No
    Local Public Document location: For Byron, the Byron Public Library 
District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481

Consumers Energy Company, Docket No. 50-155, Big Rock Point Plant, 
Charlevoix County, Michigan

    Date of application for amendment: April 30, 1997
    Brief description of amendment: The amendment revises the Big Rock 
Point Plant license and technical specifications to reflect the 
licensee's name change from ``Consumers Power Company'' to ``Consumers 
Energy Company.''
    Date of issuance: August 14, 1997
    Effective date: August 14, 1997
    Amendment No.: 119
    Facility Operating License No. DPR-6: Amendment revised the license 
and the Technical Specifications.
    Date of initial notice in Federal Register: June 4, 1997 (62 FR 
30630) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 14, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document location: North Central Michigan College, 
1515 Howard Street, Petoskey, Michigan 49770

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

    Date of application for amendment: May 8, 1997, as supplemented 
June 10, and July 25, 1997
    Brief description of amendment: The amendment incorporates 
additional NRC-approved topical reports into the Technical 
Specifications (TS).
    Date of issuance: August 12, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 202
    Facility Operating License No. DPR-50: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 4, 1997 (62 FR 
30633) The June 10 and July 25, 1997, letters provided clarifying 
information that did not change the scope of the May 8, 1997, 
application or the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated August 12, 1997. No 
significant hazards consideration comments received: No
    Local Public Document location: Law/Government Publications 
Section, State Library of Pennsylvania (REGIONAL DEPOSITORY), Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105

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

    Date of application for amendments: February 29, 1996 
(AEP:NRC:1232), and supplemented November 15, 1996 (AEP:NRC:1232A), and 
February 4, 1997 (AEP:NRC:1232B)
    Brief description of amendments: The amendments revise the 
Technical Specifications and associated bases to increase the minimum 
borated water volume in the boric acid storage system and decrease the 
required boron concentration.
    Date of issuance: August 7, 1997
    Effective date: August 7, 1997, with full implementation when the 
required plant modifications are completed, but not later than August 
31, 1998.
    Amendment Nos.: 216 and 200
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 24, 1996 (61 FR 
18172) The November 15, 1996, and February 4, 1997, supplements only 
provided the schedule for the plant modifications and procedure changes 
associated with this amendment and did not change the staff's proposed 
determination of no significant hazards consideration. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated August 7, 1997.No significant hazards consideration 
comments received: No.
    Local Public Document location:  Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085

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

    Date of application for amendments: December 20, 1996
    Brief description of amendments: The amendments reduce the 
frequency and scope of reactor coolant pump flywheel inspections.
    Date of issuance: August 8, 1997
    Effective date: August 8, 1997, with full implementation within 45 
days.
    Amendment Nos.: 217 and 201
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33126) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 8, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document location:  Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085

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

    Date of application for amendment: September 13, 1996, as 
supplemented by letter dated September 25, 1996
    Brief description of amendment: The amendment revised Technical 
Specification 5.5.B to designate the President, Maine Yankee as the 
responsible official for matters related to the Nuclear Safety Audit 
and Review (NSAR) Committee. The amendment includes some minor 
editorial changes to the same technical specification.
    Date of issuance: August 8, 1997
    Effective date: August 8, 1997, to be implemented within 30 days of 
the date of issuance.
    Amendment No.: 159
    Facility Operating License No. DPR-36: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 6, 1996 (61 FR

[[Page 45470]]

57487) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 8, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578

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

    Date of application for amendment: June 13, 1997
    Brief description of amendment: The amendment modifies Technical 
Specification (TS) Surveillance Requirement 4.4.1.3.3 to be consistent 
with the requirements of TS 3.4.1.3. Specifically, the change brings TS 
4.4.1.3.3 into agreement with TS 3.4.1.3 by requiring that the 
specified reactor coolant and/or residual heat removal system loops be 
verified in operation and circulating reactor coolant at least once per 
12 hours during Mode 4.
    Date of issuance: August 5, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 145
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 2, 1997 (62 FR 
35850) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 5, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document location: Learning Resources Center, Three 
Rivers Community-Technical College, 574 New London Turnpike, Norwich, 
Connecticut 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, Connecticut 06385

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: January 27, 1997, as 
supplemented May 16, 1997
    Brief description of amendment: The amendment changes the Technical 
Specifications to permit control rod misalignment of up to plus or 
minus 18 steps when the core thermal power is less than 85% of rated 
power.
    Date of issuance:  August 11, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 176
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 19, 1997 (62 FR 
33445) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 11, 1997. No significant 
hazards consideration comments received: No
    Local Public Document location: White Plains Public Library, 100 
Martine Avenue, White Plains, New York 10610

South Carolina Electric & Gas Company, South Carolina Public 
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear 
Station, Unit No. 1, Fairfield County, South Carolina

    Date of application for amendment: March 26, 1997
    Brief description of amendment: The amendment revises TS 4.5.2.a 
for the two charging/high head safety injection (HHSI) pump cross 
connect valves (XVG-8133A and XVG-8133B) and charging pump mini-flow 
header isolation valve (XVG-8106) in the emergency core cooling system 
(ECCS). The proposed amendment adds these valves to the list of valves 
in TS Surveillance Requirement 4.5.2.a on page 3/4 5-4, consequently 
these valves will be verified once every 12 hours to indicate that they 
are in the required position with power to the valve operators removed.
    Date of issuance: August 8, 1997
    Effective date: August 8, 1997
    Amendment No.: 136
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 21, 1997 (62 FR 
27801) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 8, 1997. No significant 
hazards consideration comments received: No
    Local Public Document location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180

South Carolina Electric & Gas Company, South Carolina Public 
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear 
Station, Unit No. 1, Fairfield County, South Carolina

    Date of application for amendment: November 14, 1995, as 
supplemented July 11, 1996 and July 24, 1997
    Brief description of amendment: The amendment revises Technical 
Specification 3/4.8.4.2 for motor-operated valves thermal overload 
protection and bypass devices at Virgil C. Summer Nuclear Station.
    Date of issuance: August 13, 1997
    Effective date: August 13, 1997
    Amendment No.: 137
    Facility Operating License No. NPF-12: Amendment adds a new License 
Condition and revises the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65684) The July 11, 1996, and July 24, 1997 submittals contained 
clarifying information only and did not change the proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated August 13, 1997. No significant hazards consideration comments 
received: No
    Local Public Document location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee Date of 
application for amendments: September 26, 1996, as supplemented on 
August 12, 1997 (TS 96-04)

    Brief description of amendments: The amendments change the 
Technical Specifications (TS) by relocating the fire protection program 
details to the Updated Final Safety Analysis Report and Fire Protection 
Plan in accordance with Generic Letters 86-10 and 88-12.
    Date of issuance: August 12, 1996
    Effective date: August 12, 1996
    Amendment Nos.: 227 and 218
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TS.
    Date of initial notice in Federal Register: July 2, 1997 (62 FR 
35843) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 12, 1997. No significant 
hazards consideration comments received: No
    Local Public Document location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402

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: August 22, 1996, as revised 
July 14, 1997
    Brief description of amendments: These amendments revise Section 
3.A of Facility Operating Licenses DPR-24 and

[[Page 45471]]

DPR-27 from a licensed power level of 1518 megawatts thermal to 1518.5 
megawatts thermal. A similar revision is made in the bases of Technical 
Specification 15.3.1.B, ``Pressure/Temperature Limits.''
    Date of issuance: August 6, 1997
    Effective date: August 6, 1997
    Amendment Nos.: 175 and 179
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the licenses.
    Date of initial notice in Federal Register: October 9, 1996 (61 FR 
52972) The July 14, 1997, supplement provided a corrected bases page 
and did not affect the staff's no significant hazards considerations 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 6, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document location: The Lester Public Library, 1001 
Adams Street, Two Rivers, Wisconsin 54241

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: February 12, 1997, as 
supplemented on March 11, 1997 (TSCR 196)
    Brief description of amendments: These amendments revise Point 
Beach Nuclear Plant's (PBNP) Technical Specifications (TSs) to relocate 
turbine overspeed protection specifications, limiting conditions for 
operation, surveillance requirements, and associated bases from TS 
Section 15.3.4, ``Steam and Power Conversion System,'' and Section 
15.4.1, ``Operational Safety Review,'' to the Final Safety Analysis 
Report (FSAR) in accordance with Generic Letter 95-10.
    Date of issuance: August 6, 1997
    Effective date: These license amendments are effective as of the 
date of issuance and shall be implemented by incorporating the turbine 
overspeed protection specifications, limiting conditions for operation, 
surveillance requirements, and associated bases into the FSAR by June 
30, 1998.
    Amendment Nos.: 176 and 180
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 23, 1997 (62 FR 
19838) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 6, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241
    Dated at Rockville, Maryland this 20th day of August 1997.
    For the Nuclear Regulatory Commission
John A. Zwolinski,
Acting Director, Division of Reactor Projects - I/II, Office of Nuclear 
Reactor Regulation.
[Doc. 97-22635 Filed 8-26-97; 8:45 am]
BILLING CODE 7590-01-F