[Federal Register Volume 61, Number 234 (Wednesday, December 4, 1996)]
[Notices]
[Pages 64381-64400]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-21204]


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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 November 8, 1996, through November 21, 1996. 
The last biweekly notice was published on November 19, 1996.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By January 3, 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

[[Page 64382]]

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. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendment request: October 31, 1996
    Description of amendment request: The proposed change would revise 
the maximum allowable water temperature as measured at the respective 
intake structures from 95 deg.F to 94 deg.F and will increase the 
minimum main reservoir level from 205.7 feet mean sea level to 215 feet 
mean sea level in Technical Specification (TS) 3/4.7.5, Ultimate Heat 
Sink.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Since the proposed change does not affect the operation of any 
accident initiating systems, the probability of occurrence of an 
accident previously evaluated will not increase. Also, none of the 
proposed changes will cause plant systems to operate outside their 
design limits or create the likelihood of a radioactive release. 
Therefore, there would be no increase in the consequences of an 
accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    No new component or system level interactions will be created by 
the proposed change in ultimate heat sink level and temperature, and 
no design limits will be exceeded. This change to [Technical] 
Specification 3/4.7.5 is more conservative than the current 
Specification limits and will serve only to restrict operation to a 
higher reservoir level and lower temperature than was previously 
allowed. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed amendment will establish a more conservative 
minimum main reservoir level such that safety-related heat 
exchangers served by Emergency Service Water will continue to remove 
their design-basis accident heat loads. Establishing a higher 
minimum reservoir level, combined with a more conservative reservoir 
temperature assumption, will involve an increase in the margin of 
safety. Also, the proposed change in maximum reservoir temperature 
from 95 deg.F to 94 deg.F will not result in any reduction in the 
margin of safety. A maximum pre-accident initial water temperature 
of 94 deg.F is necessary to yield a post-accident (30-day) 
calculated maximum inlet temperature less than or equal to the 
design basis temperature of 95 deg.F. 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.

[[Page 64383]]

    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
    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: Mark Reinhart, Acting

Duke Power Company, Docket Nos. 50-413 and 50-414, Catawba Nuclear 
Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: November 4, 1996
    Description of amendment request: The proposed amendments would 
eliminate from the Technical Specifications, Section 4.7.13.1, the 
``during shutdown'' restriction pertaining to the 18-month Standby 
Shutdown System (SSS) diesel generator inspection. Unlike Catawba 
Nuclear Station, many nuclear plants do not have an SSS facility and 
associated diesel generator. The requirements in the Technical 
Specifications for the SSS diesel generator (shared between both units) 
were patterned after similar requirements for the emergency diesel 
generators. The current wording requires that both units be shut down 
to perform the subject inspection.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    ... The standard for determining that a Technical Specification 
amendment request involves no significant hazards considerations 
requires that operation of the facility in accordance with the 
requested amendment will 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 the margin of safety.
    Criterion 1
    The proposed amendment seeks to change the surveillance 
requirements to allow the SSS DG [diesel generator] periodic 
inspection with one or both units on line. The surveillance can be 
safely completed as proposed without affecting unit operation. The 
equipment would not be removed from service for a time that would 
exceed the current Limiting Condition For Operation or the 
appropriate action statement would be entered. The probability or 
consequences of any accident previously evaluated will not be 
significantly increased because the removal of the SSS DG from 
service can be safely performed while one or both units are 
operating.
    Criterion 2
    The proposed amendment change does not change any actual 
surveillance requirements. The change simply allows the 18 month SSS 
DG inspection to be performed at different unit conditions. The 
performance of the surveillance with the units operating do not 
require any new component configurations that would reduce the 
ability of any equipment to mitigate an accident. The station is not 
degraded beyond that which has been previously evaluated. Therefore 
the proposed change does not create the possibility of a new or 
different kind of accident.
    Criterion 3
    The allowed outage time for the SSS DG, as specified by the 
Limiting Condition For Operation, defines the required margin of 
safety for equipment operability. Removing the SSS DG from service 
for periodic inspection and returning it to service within the 
allowed outage time 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 
proposed amendments involve no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: October 30, 1996
    Description of amendment request: The proposed changes would (1) 
completely rewrite Technical Specification (TS) 4.4.2 to incorporate a 
prestressed concrete containment surveillance program that is 
consistent with Regulatory Guide 1.35, (2) modify TS 3.6.7 by 
establishing new Limiting Conditions for Operation and required actions 
related to the structural integrity of the reactor buildings, (3) 
incorporate an editorial change to TS 6.6.3 to reference the relocated 
tendon surveillance reporting requirements, and (4) modify TS 3.6.7 
Bases to describe the Reactor Building post-tensioning 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:
    1) Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed amendment to Oconee Technical Specifications 
involves the implementation of an enhanced surveillance program for 
the reactor building prestressed concrete containment and the 
assurance of appropriate station response to abnormal degradation of 
the containment structure. The proposed change will move Oconee into 
a surveillance program which is consistent with accepted industry 
practice and a published NRC regulatory position. The adoption of 
Regulatory Guide 1.35 as a basis for the periodic inspection of the 
reactor building prestressed concrete containment and clearly 
defined station response to any indication of structural 
deterioration will assure acquisition of sufficient data to 
demonstrate that structural integrity is maintained and, if 
necessary, appropriate compensatory action(s) are taken. By assuring 
that any adverse trends in the behavior of the prestressed concrete 
containment are identified and acted upon in a timely manner, this 
change does not increase the probability or consequences of an 
accident previously evaluated.
    2) Will the change create the possibility of a new or different 
kind of accident from any previously evaluated?
    No. The proposed amendment to Oconee Technical Specifications 
involves the implementation of an enhanced surveillance program for 
the reactor building prestressed concrete containment and the 
assurance of appropriate station response to abnormal degradation of 
the containment structure. By adopting Regulatory Guide 1.35 as a 
basis for the surveillance inspection program for the reactor 
building prestressed concrete containment and clearly defining 
required station response to any indication of structural 
deterioration, sufficient data will be obtained to demonstrate that 
structural integrity is being maintained and that any adverse 
behavioral trends are identified and acted upon in a timely manner. 
Therefore, the proposed amendment does not create the possibility of 
any type of accident: new, different or previously evaluated.
    3) Will the change involve a significant reduction in a margin 
of safety?
    No. Margin of safety is associated with confidence in the 
ability of the fission product barriers (i.e., fuel and fuel 
cladding, Reactor Coolant System pressure boundary, and containment 
structure) to limit the level of radiation dose to the public. The 
proposed Technical Specifications amendment will move Oconee into a 
surveillance program which is consistent with accepted industry 
practice and a published regulatory position. By ensuring more 
timely identification of, and response to, any adverse trend in the 
behavior of the reactor building prestressed concrete containment, 
continued maintenance of the structural integrity is enhanced. 
Therefore, the ability of the containment structure to perform the 
intended function of protecting the public

[[Page 64384]]

from radiation dose is further assured, and no reduction in any 
existing margin of safety will occur.
    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: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, 
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport, 
Pennsylvania

    Date of amendment request: September 9, 1996
    Description of amendment request: The proposed amendment would 
modify the design features section (Section 5.0) of the Technical 
Specifications (TSs) to make the design features section consistent 
with the four criteria specified in the Commission's Policy Statement 
on TSs (58 FR 39132) and with the guidance provided in the NRC's 
Standard Technical Specifications, Westinghouse Plants (NUREG-1431, 
Revision 1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change reduces the content of the technical 
specification (TS) design feature section consistent with the 
Improved Standard Technical Specifications (ISTS) of NUREG-1431. The 
information that has been removed is also contained in the UFSAR 
[Updated Final Safety Analysis Report] or offsite dose calculation 
manual (ODCM); therefore, duplication of the information is 
eliminated to improve the use of the TS. Because the information 
removed from the TS is maintained in the UFSAR or ODCM where changes 
are controlled in accordance with regulatory requirements, there is 
no reduction in commitment and adequate control is provided. 
Elimination of information from the design feature section of the TS 
which duplicates information in the UFSAR enhances the usability of 
the TS without reducing commitments. These changes clarify and 
improve the understanding and readability of the TS. Since the 
requirements remain the same, these changes only affect the method 
of presentation and would not affect possible initiating events for 
accidents previously evaluated or any system functional requirement. 
Therefore, the proposed changes would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The relocation of existing requirements, the elimination of 
requirements which duplicate existing information, and making 
administrative improvements are all changes that are administrative 
in nature. The proposed changes will not affect any plant system or 
structure, not [nor] will they affect any system functional or 
operability requirements. Consequently, no new failure modes are 
introduced as a result of the proposed changes. The proposed changes 
are consistent with the ISTS, for the most part, as plant-specific 
information is included in this section. Therefore, the proposed 
change will not create the possibility of a new or different type of 
accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes are administrative in nature in that no 
change to the design features of the facility are being made. The 
design features section is being reformatted to be consistent, for 
the most part, with the ISTS. The proposed changes do not affect the 
UFSAR design bases, accident assumptions, or technical specification 
bases. In addition, the proposed changes do not affect release 
limits, monitoring equipment or practices. Therefore, the proposed 
change will 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: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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: October 24, 1996
    Description of amendment request: The proposed amendment would 
revise the technical specifications to remove accelerated testing 
requirements for the standby diesel generators. The changes implement 
the provisions of Generic Letter (GL) 94-01, ``Removal of Accelerated 
Testing and Special Reporting Requirements For Diesel Generators'', 
dated May 31, 1994.
    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. This request does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    This change will provide flexibility to structure the standby 
diesel generator maintenance program based on the risk significance 
of the structures, systems, and components that are within the scope 
of the Maintenance Rule. The removal of the diesel generator 
accelerated testing is acceptable as the maintenance rule applies 
site and system specific performance criteria to monitor diesel 
generator performance. This criteria includes a running availability 
and reliability goal as well as specific goals to monitor 
maintenance preventable functional failures. The performance 
criteria for the diesel generator reliability and unavailability 
established by the maintenance rule and the causal determinations 
and corrective actions required for maintenance preventable 
functional failures are considered to be an acceptable method for 
monitoring diesel generator performance.
    The proposed change has no effect on the probability of the 
initiation of an accident, because the emergency diesel generators 
do not serve as the initiator of any event. Additionally, as diesel 
generator performance will continue to be assured by the maintenance 
rule, the proposed changes do not affect the ability to mitigate the 
consequences of an accident previously evaluated. The changes do not 
impact the diesel's design sources, operating characteristics, 
system functions, or system interrelationships. The failure 
mechanisms for the accidents previously analyzed are not affected 
and no additional failure modes are created that could cause an 
accident that has been previously evaluated. Since the diesel 
generator performance and reliability will continue to be assured by 
the maintenance rule, the proposed changes cannot involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. This request does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    This proposed change does not involve a change to the plant 
design or operation. As a result, the proposed changes does not 
affect any of the parameters or conditions that could contribute to 
the initiation of any accidents. The proposed changes only affect 
the methods used to monitor and assure diesel generator performance. 
The performance criteria for both the diesel generator reliability 
and unavailability established by the maintenance rule, and the

[[Page 64385]]

casual determinations and corrective actions required for 
maintenance preventable functional failures, is considered by GL 94-
01 to be an acceptable method for monitoring diesel generator 
performance.
    No [system, structure, or component] SSC, method of operating, 
or system interface is altered by this change. The changes do not 
impact the diesel's design sources, operating characteristics, 
system functions, or system interrelationships. The failure 
mechanisms for the accidents are not affected, and no additional 
failure modes are created. Because the diesel generator performance 
and reliability will continue to be assured by the maintenance rule, 
the proposed changes cannot create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The request does not involve a significant reduction in a 
margin to safety.
    The proposed changes only affect the methods used to monitor and 
assure diesel generator performance and reliability. The performance 
criteria for both the diesel generator reliability and 
unavailability established by the maintenance rule, and the casual 
determinations and corrective actions required for maintenance 
preventable functional failures, is considered by GL 94-01 to be an 
acceptable method for monitoring diesel generator performance. No 
margin to safety as defined in the basis for any technical 
specification is impacted by these changes. This change does not 
impact any uncertainty in the design, construction, or operation of 
any SSC. Diesel generator response to accident initiators is 
unchanged. No SSC, method of operating, or system interface is 
altered by this change. The changes do not impact the diesel's 
design sources, operating characteristics, system functions, or 
system interrelationships. Because the diesel generator performance 
and reliability will continue to be assured by the maintenance rule, 
the proposed changes cannot involve a significant reduction in the 
margin to safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005
    NRC Project Director: William D. Beckner

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: November 6, 1996
    Description of amendment request: The proposed amendment would 
revise the River Bend Station (RBS) Fire Hazards Analysis Report and 
Safety Analysis Report to allow a deviation from 10 CFR Part 50, 
Appendix R, Section III.G.2.c with respect to the requirement for an 
area wide automatic fire suppression system in Fire Area C-16. The 
deviation would allow a 1-hour barrier to separate redundant trains of 
post fire safe shutdown equipment within the fire area and partial 
sprays on the protected train.
    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 request does not involve an increase in the probability 
or consequences of an accident previously evaluated.
    The event of concern is a fire in Fire Area C-16. The low fire 
loading and minimal concentration of exposed combustible material in 
Fire Area C-16 would limit fire spread. However, for this scenario, 
all unprotected equipment in Fire Area C-16 will be assumed lost. 
Fire Area C-16 contains cables for both Division I and Division II 
components required for post fire safe shutdown. The loss of both 
divisions of cables could preclude the ability of the plant to 
achieve post fire safe shutdown. Protection of the required Division 
II cables in a 1-hour fire barrier in conjunction with a partial 
area, automatic suppression system installed above and below the 
protected trays will ensure that post fire safe shutdown can be 
achieved.
    In summary, the probability of a fire occurring in Fire Area C-
16 is not affected. However, if a fire were to occur in Fire Area C-
16 which caused the loss of Division I powered components, Division 
II powered components, by virtue of the 1-hour fire barrier and 
partial area, automatic suppression system, would remain available. 
The low fire loading and minimal concentration of exposed 
combustible material in Fire Area C-16 would limit fire spread. The 
proposed fire protection scheme provides a level of protection 
commensurate with the original design. Therefore, this request does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The request does not create the possibility of occurrence of 
a new or different kind of accident from any accident previously 
evaluated.
    Fire Area C-16 will be protected by a partial area, automatic 
suppression system installed above and below the protected cable 
trays. Fire suppression systems are generally used to limit fire 
spread, once the heat of the fire opens thermally sensitive 
sprinklers. The low fire loading and minimal concentration of 
exposed combustible material in Fire Area C-16 would aid in limiting 
fire spread, and would also limit the severity of any plausible 
fire. The previous analysis assumed all Division I components and 
cables in the area would be lost, and that the installed fire 
barrier would adequately protect the Division II cables routed 
through C-16. The required Division II cables will be enclosed in a 
1-hour fire barrier with a partial area, automatic suppression 
system. These features provide a level of protection commensurate 
with that of the previous design. In addition, the total combustible 
loading in the area results in a maximum theoretical worst case fire 
duration of less than 1-hour.
    In summary, if a fire were to occur in Fire Area C-16 which 
caused the loss of Division I powered components, post fire safe 
shutdown could still be achieved using Division II. Therefore, this 
request does not create the possibility of occurrence of a new or 
different kind of accident from any accident previously evaluated.
    3. The request does not involve a significant reduction in a 
margin of safety.
    In this case, the margin of safety is implicit rather than being 
explicitly expressed as a numerical value. An implicit margin of 
safety involves conditions for NRC acceptance. Since the RBS 
Technical Specification Bases do not specifically address a margin 
of safety for fire protection, the SAR [Safety Analysis Report], the 
NRC's Safety Evaluation Report (SER), and appropriate other 
licensing basis documents were reviewed to determine if the proposed 
change would result in a reduction in a margin of safety. As stated, 
in part, in Attachment 4 to NPF-47 [Facility Operating License; NPF-
47]:
    EOI [Entergy Operations, Inc.] shall implement and maintain in 
effect all provisions of the approved fire protection program as 
described in the Final Safety Analysis Report for the facility 
through Amendment 22 and as approved in the SER dated May 1984 and 
Supplement 3 dated August 1985 subject to provisions 2 and 3 ....
    As discussed in the Reason for Request, SSER [Supplemental 
Safety Evaluation Report] 3 dated August 1985 states, in part:
    On the basis of its evaluation the staff finds that the 
applicant's fire protection program with approved deviations is in 
conformance with the guidelines of BTP CMEB [branch technical 
position, Chemical Materials and Engineering Branch] 9.5-1, 
[S]sections III.G, III.J, and III.O of Appendix R to 10 CFR [Part] 
50, and GDC [General Design Criteria] 3, and is, therefore, 
acceptable.
    Thus, the margin of safety in this case can be defined as 
conformance with the specified fire protection guidelines.
    10 CFR [Part] 50, Appendix R, Section III.G.2, requires, in 
part, that redundant trains of post fire safe shutdown equipment 
located in the same fire area be separated by a 1-hour fire barrier 
and, in addition, that fire detection and an automatic fire 
suppression system be installed in the are under consideration. 
Since Fire Area C-16 will have a partial area, automatic suppression 
system, this fire area would deviate from the requirements of 10 CFR 
[Part] 50, Appendix R, Section III.G.2.c. However, as discussed 
previously, the installed partial area, automatic suppression 
system, the low fire loading and minimal amount of exposed 
combustibles compensate for the lack of a total, area wide, 
automatic fire suppression

[[Page 64386]]

system. There is no adverse impact on the ability to achieve and 
maintain post fire safe shutdown. Therefore, this request 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: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005
    NRC Project Director: William D. Beckner

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

    Dates of amendment requests: October 28, 1996 (Two letters)
    Description of amendment request: The licensee proposed to change 
the St. Lucie Units 1 and 2 Technical Specifications (TS) to implement 
10 CFR 50, Appendix J, Option B, for containment leakage testing by 
referring to Regulatory Guide 1.163, ``Performance-Based Containment 
Leakage-Test Program.'' Changes include relocating the details for 
containment testing to the ``containment leakage rate testing program'' 
and adding the requirements of the containment leakage rate testing 
program to TS 6.8.4, which describes facility programs. Changes are 
also proposed to remove Tables 3.6-1, ``Containment Leakage Paths,'' 
and 3.6-2, ``Containment Isolation Valves'' from TS and relocate the 
information to plant procedures.
    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 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
due to the following reasons:
    a)These proposed changes are all consistent with NRC 
requirements and guidance for implementation of 10 CFR 50, Appendix 
J, Option B, except for the use of Bechtel Topical Report BN-TOP-1 
for type A testing. BN-TOP-1 has been previously approved for use in 
accordance with 10 CFR 50 appendix J.
    b) Based on industry and NRC evaluations performed in support of 
developing Option B, these changes potentially result in a minor 
increase in the consequences of an accident previously evaluated due 
to the increased testing intervals. However, the proposed changes do 
not result in an increase in the core damage frequency since the 
containment system is used for mitigation purposes only.
    c) These changes are expected to result in increased attention 
to components with poor leakage test history as part of the 
performance-based nature of Option B, such that the marginally 
increased consequences from the expanded testing intervals may be 
further reduced or negated.
    Therefore, these changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.(2) Operation of the facility in accordance with the 
proposed amendments would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The use of the modified specifications can not create the 
possibility of a new or different kind of accident from any 
previously evaluated since the proposed amendments will not change 
the physical plant or the modes of plant operation defined in the 
facility operating license. No new failure mode is introduced due to 
the implementation of a performance-based program for containment 
leakage rate testing, since the proposed changes do not involve the 
addition or modification of equipment, nor do they alter the design 
or operation of affected plant systems, structures, or 
components.(3) Operation of the facility in accordance with the 
proposed amendments would not involve a significant reduction in a 
margin of safety.
    The operating limits and functional capabilities of the affected 
systems, structures, and components are basically unchanged by the 
proposed amendments. The increase in intervals between leak-test 
surveillances will not significantly reduce the margin of safety as 
shown by findings in NUREG 1493, ``Performance-Based Containment 
Leak-Test Program'', which was based on implementation of the 
performance-based testing of Option B.
    Therefore these changes do not involve a significant reduction 
in the margin of safety.The NRC staff has reviewed the licensee's 
analysis and, based on thisreview, it appears that the three 
standards of 50.92(c) are satisfied.Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: M. S. Ross, Attorney, Florida Power & Light, 
11770 US Highway 1, North Palm Beach, Fl 33408
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: October 30, 1996
    Description of amendment request: The proposed amendments will 
revise Technical Specification (TS) 3/4.9.10, ``Refueling Operations, 
Water Level-Reactor Vessel.'' The Limiting Condition for Operation 
(LCO) specified for the minimum allowed refueling water level is not 
altered, but the Applicability, Action, and Surveillance Requirements 
are changed to remove inconsistencies with the definition of Core 
Alterations, and to achieve consistency with the generic Standard 
Technical Specifications for Combustion Engineering Plants (NUREG-
1432). An editorial change is proposed for TS 3/4.9.9, ``Refueling 
Operations, Containment Isolation System,'' and, for St. Lucie Unit 1, 
the LCO is modified to conform with other related refueling 
specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.
    (1) 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.
    Certain evolutions performed with the UGS [upper guide 
structure] in place are not Core Alterations, and the revised LCO 3/
4.9.10 will allow these activities to be performed at water levels 
other than prescribed by the existing LCO. Since these activities 
are performed with the UGS in place, the probability that a fuel 
handling accident would occur is not impacted by the proposed 
changes. The minimum water level required for Core Alterations and 
movement of irradiated fuel in containment is not altered by the 
proposed changes, nor are any assumptions or conditions changed that 
were used as inputs to the evaluation of fuel handling accident 
consequences. The changes to Specification 3/4.9.9 are 
administrative in nature and resolve an inconsistency between the 
operability requirements for the containment isolation system and 
the containment penetrations that the system would isolate at PSL1 
[Plant St. Lucie 1]. Therefore, operation of either facility in 
accordance with its proposed amendment would not involve a 
significant increase in the probability or 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

[[Page 64387]]

kind of accident from any accident previously evaluated.
    The proposed changes are administrative in nature, in that the 
changes do not involve the addition or modification of equipment nor 
do they alter the design of plant systems. New failure modes are not 
introduced, and the physical plant or the modes of plant operation 
defined in the Facility License are not altered. Therefore, 
operation of either facility in accordance with its 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 safety margin associated with a fuel handling accident is 
determined, in part, by the minimum refueling water level allowed 
for conducting Core Alterations and movement of irradiated fuel in 
containment. The minimum water level required by LCO 3/4.9.10, or 
other factors considered as inputs to the safety analysis, is not 
changed by the proposed amendments. The revised applicability 
requirements for LCO 3/4.9.9 at PSL1 will allow the containment 
isolation system to be inoperable only during those Mode 6 
conditions where Core Alterations or irradiated fuel movements 
within containment are not in progress, or each required containment 
penetration is otherwise closed. Therefore, operation of either 
facility in accordance with its 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 
thisreview, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendmentrequest involves no significant hazards consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: M. S. Ross, Attorney, Florida Power & Light, 
11770 US Highway 1, North Palm Beach, Fl 33408
    NRC Project Director: Frederick J. Hebdon

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: October 23, 1996, as supplemented by 
letter dated November 6, 1996.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.4.6.1, regarding reactor coolant 
system leakage detection instrumentation, to adopt the requirements 
found in NUREG-1431, ``Standard Technical Specifications Westinghouse 
Plants,'' for the reactor coolant system leakage detection 
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:
    1. The proposed change does not involved a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change reduces the number of containment 
atmospheric radioactivity channels which must be OPERABLE when 
operating in MODES 1, 2, 3, and 4 from two to one. This change does 
not significantly increase the probability or consequences of a 
previously evaluated accident since the plant will continue to have 
diverse and independent means of detecting significant changes in 
the amount of leakage from the RCS [reactor coolant system]; the 
normal sump level and flow monitoring system, at least one of the 
two containment atmospheric radiation monitors, and the periodic 
precision RCS water inventory balance required by Technical 
Specification surveillance requirement 4.4.6.2.1.c. In addition, STP 
[South Texas Project] design includes advanced trending displays 
which can assist in detecting leakage based on changes in the volume 
control tank or pressurizer level. Other instruments, which are not 
listed in the Technical Specification related to leakage, but which 
can provide indication of leakage, are the containment pressure, 
temperature and humidity indicators. Good operating practice and 
commercial risk associated with long term inoperability of both 
monitors assures that an inoperable containment atmospheric 
radiation monitor will be promptly returned to service.
    The proposed change also revises the limitation on continued 
operation with both containment atmospheric radiation monitors 
inoperable from 72 hours to 30 days. This change is based on the 
continued availability of diverse and redundant instrumentation 
discussed above to detect and indicate RCS leakage.
    The Actions required as a result of this change include analysis 
of a containment atmospheric grab sample or performance of a 
precision RCS water inventory balance in accordance with 
surveillance requirement 4.4.6.2.1.c. The containment normal sump 
level flow monitoring system will also promptly identify changes in 
RCS leakage. Other installed instrumentation, such as containment 
pressure, temperature, and humidity, will provide indications of 
significant increases in leakage. Slower increases will be detected 
by the daily inventory balance or the daily grab samples analysis, 
and the three day inventory balance.
    Inoperability of the on-line automatic containment normal sump 
level and flow monitoring system can be compensated for by the 
performance of a daily manual calculation, a precision RCS inventory 
balance as described in surveillance requirement 4.4.6.2.1.c, or the 
other available indications of increases in leakage such as the 
containment atmospheric radiation monitoring instruments and 
installed containment temperature, pressure and humidity 
instrumentation. The STP control room design also incorporates 
features which allow rapid detection of unexpected changes in the 
volume control tank and pressurizer level through available 
instrument trend displays. The combination of the compensatory 
measures, diverse and separate channels, and non-TS [non-technical 
specification] required instrumentation provides a sufficient level 
of detection to assure prompt identification and quantification of 
leakage with an inoperable containment normal sump level and flow 
monitoring system. The allowable outage time of 30 days provides 
assurance the normal containment sump level and flow monitoring 
system will be returned to service in a reasonable amount of time.
    Based on the continued availability of adequate and redundant 
instrumentation to detect changes in RCS leakage rate, this change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not require the installation of any new 
or different kind of equipment. Nor does the change involve any 
significant new or different MODE of operation of the plant. The 
proposed change reduces the number of required containment 
atmospheric radiation monitors, and provides a 30 day allowed outage 
time for either the containment atmosphere radioactivity monitor or 
the containment normal sump level and flow monitoring system. 
Therefore, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    In addition, as described above, the proposed change does not 
significantly reduce a margin of safety. Small changes in RCS leak 
rates are typically detected over a relatively long period of time. 
Diverse instrumentation continues to be available to plant operators 
which will assist in early detection of any change. The STP design 
provides additional non-Technical Specification human factors which 
assist in assuring any changes in leakage will be quickly detected.
    The proposed change extends the amount of time that the 
containment atmospheric radiation monitors may be inoperable. The 
extension is based on the continued availability of equipment which 
provides a level of detection capability adequate to detect 
increases in RCS leakage and which continues to be diverse and 
independent. This protection is afforded by the continued 
OPERABILITY of the containment normal sump level and flow monitoring 
system, the daily performance of a precision RCS

[[Page 64388]]

inventory balance as described by surveillance requirement 
4.4.6.2.1.c or the daily analysis of containment atmospheric grab 
samples, and other instrumentation such as pressure, temperature and 
humidity indicators.
    The combination of the compensatory measures, diverse and 
separate channels, and non-TS required instrumentation provides a 
sufficient level of detection to assure prompt identification and 
quantification of leakage with an inoperable containment normal sump 
level and flow monitoring system. Additionally, the compensatory 
measure of performing either a daily manual calculation or precision 
RCS inventory balance, provides assurance that the level of safety 
is maintained when the containment normal sump level and flow 
monitoring system is inoperable. The allowable outage time of 30 
days provides assurance the normal containment sump level and flow 
monitoring system will be returned to service in a reasonable amount 
of time.
    Based on the compensatory actions and available installed 
equipment, the proposed changes do not represent 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
    NRC Project Director: William D. Beckner

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

    Date of amendment requests: August 15, 1996
    Description of amendment requests: The proposed amendments would 
revise the Containment Cooling Systems Limiting Conditions for 
Operation Technical Specifications to bring them into conformance with 
recently completed system analyses by no longer permitting both 
containment spray pumps to be inoperable at the same time.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment[s] will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Operation of the Prairie Island plant in accordance with the 
proposed changes does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. 
None of the proposed changes involve a physical modification to the 
plant.
    These changes will require operability of at least one 
containment spray pump at all times and reduces the spray additive 
tank allowable outage time from 72 hours to 24 hours. Both of these 
changes are more conservative and safer than currently required in 
the Prairie Island Technical Specifications. These proposed changes 
do allow one containment fan cooler train out of service for 7 days 
instead of 72 hours as allowed by current Technical Specifications. 
Recent plant analyses confirm that one containment fan cooler train 
with one containment spray train is sufficient to meet the system 
design bases. Since the probability of an accident occurring is low 
while one containment fan cooler train is out of service, the 
probability and consequences of an accident are not significantly 
increased.
    In total these changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed amendment[s] will not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the proposed changes, in themselves, do not introduce a new 
mode of plant operation, surveillance requirement or involve a 
physical modification to the plant.
    The proposed changes do require more restrictive, safer 
containment spray train operability. The proposed changes also allow 
one containment fan cooler train to be out of service for 7 days 
instead of 72 hours as allowed by the current Technical 
Specifications. However, this change does not create the possibility 
of a new kind of accident.
    The proposed changes do no alter the design, function, or 
operation of any plant components and therefore, no new accident 
scenarios are created.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated would not be created 
by these amendments.
    3. The proposed amendment[s] will not involve a significant 
reduction in the margin of safety. This License Amendment Request 
require[s] one containment spray train to be operable at all times 
which is more restrictive than current Technical Specifications and 
thus the margin of safety is not reduced.
    This License Amendment Request will also allow one containment 
fan cooler train to be out of service for 7 days instead of 72 hours 
as allowed by the current Technical Specifications. Since the 
remaining containment cooling components can mitigate an accident 
and the probability of a design basis accident are low during this 
time, this change does not significantly reduce the plant margin of 
safety.
    Therefore, a significant reduction in the margin of safety would 
not be involved with these amendments.
    Based on the evaluation described above, and pursuant to 10 CFR 
Part 50, Section 50.91, Northern States Power Company has determined 
that operation [of] the Prairie Island Nuclear Generating Plant in 
accordance with the proposed license amendment request does not 
involve any significant hazards considerations as defined by Nuclear 
Regulatory Commission regulations in 10 CFR Part 50, Section 50.92.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment requests: September 24, 1996, as supplemented 
October 17, 1996.
    Description of amendment requests: The proposed amendments would 
revise the Technical Specifications (TS) for the Prairie Island Nuclear 
Generating Plant to allow use of an alternate steam generator tube 
repair criteria (elevated F-star or EF*) in the tubesheet region when 
used with the repair method of additional roll expansion. The 
amendments incorporate revised acceptance criteria for tubes with 
degradation in the tubesheet region and enable the licensee to avoid 
unnecessary plugging and sleeving of steam generator tubes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment[s] will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The supporting technical and safety evaluations of the subject 
criterion

[[Page 64389]]

demonstrate that the presence of the tubesheet will enhance the tube 
integrity in the region of the hardroll by precluding tube 
deformation beyond its initial expanded outside diameter. The 
resistance to both tube rupture and tube collapse is strengthened by 
the presence of the tubesheet in that region. The results of 
hardrolling of the tube into the tubesheet is an interference fit 
between the tube and the tubesheet. Tube rupture cannot occur 
because the contact between the tube and tubesheet does not permit 
sufficient movement of tube material. The radial preload developed 
by the rolling process will secure a postulated separated tube end 
within the tubesheet during all plant conditions. In a similar 
manner, the tubesheet does not permit sufficient movement of tube 
material to permit buckling collapse of the tube during postulated 
LOCA [loss-of-coolant accident] loadings.
    The EF* length of roll expansion is sufficient to preclude tube 
pullout from tube degradation located below the EF* distance, 
regardless of the extent of the tube degradation. The existing 
Technical Specification leakage rate requirements and accident 
analysis assumptions remain unchanged in the unlikely event that 
significant leakage from this region does occur. As noted above, 
tube rupture and pullout is not expected for tubes using the EF* 
criterion. Any leakage out of the tube from within the tubesheet at 
any elevation in the tubesheet is fully bounded by the existing 
steam generator tube rupture analysis included in the Prairie Island 
Plant USAR [updated safety analysis report]. For plants with partial 
depth roll expansion like Prairie Island, a postulated tube 
separation within the tube near the top of the roll expansion (with 
subsequent limited tube axial displacement) would not be expected to 
result in coolant release rates equal to those assumed in the USAR 
for a steam generator tube rupture event due to the limited gap 
between the tube and tubesheet. The proposed plugging criterion does 
not adversely impact any other previously evaluated design basis 
accident.
    Leakage testing of roll expanded tubes indicates that for roll 
lengths approximately equal to the EF* distance, any postulated 
faulted condition primary to secondary leakage from EF* tubes would 
be insignificant.
    2. The proposed amendment[s] will not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    Implementation of the proposed EF* criterion does not introduce 
any significant changes to the plant design basis. Use of the 
criterion does not provide a mechanism to initiate an accident 
outside of the region of the expanded portion of the tube. Any 
hypothetical accident as a result of any tube degradation in the 
expanded portion of the tube would be bounded by the existing tube 
rupture accident analysis. Tube bundle structural integrity will be 
maintained. Tube bundle leaktightness will be maintained such that 
any postulated accident leakage from EF* tubes will be negligible 
with regard to offsite doses.
    3. The proposed amendment[s] will not involve a significant 
reduction in the margin of safety.
    The use of the EF* criterion has been demonstrated to maintain 
the integrity of the tube bundle commensurate with the requirements 
of Reg Guide 1.121 [<gr-thn-eq>Bases for Plugging Degraded PWR Steam 
Generator Tubes<gr-thn-eq>] (intended for indications in the free 
span of tubes) and the primary to secondary pressure boundary under 
normal and postulated accident conditions. Acceptable tube 
degradation for the EF* criterion is any degradation indication in 
the tubesheet region, more than the EF* distance below the bottom of 
the transition between the roll expansion and the unexpanded tube. 
The safety factors used in the verification of the strength of the 
degraded tube are consistent with the safety factors in the ASME 
[American Society of Mechanical Engineers] Boiler and Pressure 
Vessel Code used in steam generator design. The EF* distance has 
been verified by testing to be greater than the length of roll 
expansion required to preclude both tube pullout and significant 
leakage during normal and postulated accident conditions. Resistance 
to tube pullout is based upon the primary to secondary pressure 
differential as it acts on the surface area of the tube, which 
includes the tube wall cross-section, in addition to the inner 
diameter based area of the tube. The leak testing acceptance 
criteria are based on the primary to secondary leakage limit in the 
Technical Specifications and the leakage assumptions used in the 
USAR accident analyses.
    Implementation of the tubesheet plugging criterion will decrease 
the number of tubes which must be taken out of service with tube 
plugs or repaired with sleeves. Both plugs and sleeves reduce the 
RCS (reactor coolant system) flow margin; thus, implementation of 
the EF* criterion will maintain the margin of flow that would 
otherwise be reduced in the event of increased plugging or sleeving.
    Based on the above, it is concluded that the proposed change 
does not result in a significant reduction in margin with respect to 
plant safety as defined in the USAR or the Technical Specification 
Bases.
    Based on the evaluation described above, and pursuant to 10 CFR 
Part 50, Section 50.91, Northern States Power Company has determined 
that operation of the Prairie Island Nuclear Generating Plant in 
accordance with the proposed license amendment request does not 
involve any significant hazards considerations as defined by NRC 
regulations in 10 CFR Part 50, Section 50.92.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment request: June 10, 1996, as supplemented July 25, 
1996
    Description of amendment request: The proposed amendment would 
change the differential temperature Technical Specification Allowable 
Values and Trip Setpoints for the Reactor Water Cleanup penetration 
room steam leak detection function.
    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 [of occurrence] [sic] or consequences of an 
accident evaluated.
    FSAR section 5.2.5.1.3 addresses the ambient and differential 
room ventilation temperature leakage detection. This section states:
    ``...switch setpoints are based on the temperature rise 
resulting from a leak at system conditions corresponding to full 
reactor power.''
    NRC Safety Evaluation on the RWCU system steam leak detection 
system (related to Amendment Number 123 to License NPF-14 and 
Amendment Number 90 to License NPF-22) reviewed and found acceptable 
the PP&L criteria for calculating the leak detection setpoints for 
the RWCU system, which include:
    1. Setpoints are selected to detect and isolate a leak that is 
normally less than 25 gpm and below the flow rate corresponding for 
the critical crack size for the system piping.
    2. Setpoints are set high enough to avoid inadvertent isolation 
caused by normal temperature transients or abnormal transients 
caused by non-leak conditions (such as loss of ventilation).
    This NRC SER also stated that a leak rate of 25 gpm is less than 
those leak rates associated with the onset of unstable pipe 
ruptures. This fact is also shown in FSAR figure 5.2-10. This value 
of 25 gpm constitutes the design basis for the steam leak detection 
system.
    The mixing and liquid energy addition assumption changes in the 
analysis do not affect this design basis. The analysis calculates 
the resulting room temperature increase from a 25 gpm leak. In fact, 
the new assumptions provide a more accurate yet conservative 
prediction of room temperature increases. Therefore, operation of 
the system is improved.

[[Page 64390]]

    The proposed change leads to higher calculated room temperatures 
to be used in the differential temperature setpoint calculations. 
The engineering study was reviewed to determine if the higher 
calculated temperatures would have a negative impact on the High 
Energy Line Break and Leak Analysis environmental study which 
provides the basis for equipment qualification.
    In determining the room temperatures, the engineering study 
considers ambient temperature setpoints at which the leaks will be 
isolated. The proposed action will not change the ambient 
temperature setpoints, and actuation of these instruments will 
ensure that the results of the engineering study will not be 
adversely affected. Therefore, no impact on equipment qualification 
is being introduced by this change.
    FSAR chapter 15 was reviewed for potential impacts on the 
accident analyses. The 25 gpm leak outside containment is not 
specifically analyzed in FSAR chapter 15. However, other conditions 
which result in coolant leakage outside containment are analyzed in 
section 15.6.2 (Instrument Line Break) and 15.6.4 (Steam System 
Piping Break Outside Containment). As stated in the NRC SER, the 25 
gpm RWCU leak rate is bounded by the analysis in FSAR section 
15.6.4. FSAR section 15.6.2 also states that leak detection 
actuations will initiate operator actions, a fact that is not 
affected by the proposed change. Therefore, based on a review of 
FSAR chapter 15 it was concluded that no impact on the analyzed 
accident scenarios is created by the proposed change.
    Based on the above discussions, it is demonstrated that the 
proposed change will not adversely impact system function or 
equipment. System performance will actually be improved since the 
new setpoints eliminate spurious isolations resulting from a less 
accurate model. The setpoint change has no impact on any equipment 
important to safety or any accidents previously analyzed in the 
FSAR. Therefore, the proposed change does not involve a significant 
increase in the probability of occurrence or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed action does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
Neither the system design basis nor the system function will be 
adversely affected. System performance will be enhanced since 
spurious differential temperature actuations will be reduced as a 
result of using the more accurate, yet conservative, COTTAP model. 
In addition to this, redundant temperature isolation function will 
continue to be provided by the existing high ambient temperature 
detectors.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed action does not involve a significant reduction in 
a margin of safety. The Technical Specification basis for the 
setpoints is to detect a leak below the flow rate corresponding to 
critical crack size for the system piping. As stated previously, the 
25 gpm flow rate is an acceptable flow rate and is used to calculate 
the new temperatures.
    Although the newly calculated RWCU penetration room temperatures 
will be higher (due to the improved model), the isolation actuation 
will be initiated by the high ambient temperature detectors before 
the penetration room temperatures reach the newly calculated values, 
as would happen under the old model. Therefore, system response is 
not adversely affected.
    The current temperature values lead to differential temperature 
setpoints which are too low, causing spurious isolations. The use of 
the new temperature values will reduce the number of spurious 
isolations, reducing unnecessary challenges to safety systems during 
normal plant operations.
    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: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

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

    Date of amendment request: September 18, 1995
    Description of amendment request: The proposed Technical 
Specifications (TS) changes would revise TS Table 4.3.1.1-1, ``Reactor 
Protection System Instrumentation Surveillance Requirements'' to 
reflect the change in the calibration frequency for the Local Power 
Range Monitor (LPRM) signal from every 1000 Effective Full Power Hours 
(EFPH) to every 2000 Megawatt Days per Standard Ton (MWD/ST).
    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 Technical Specifications (TS) change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The change in the calibration frequency of the Local Power Range 
Monitor (LPRM) signal does not make any physical change to the fuel 
or the manner in which the fuel responds to a transient or accident. 
The proposed TS change does not affect the fundamental method by 
which the LPRMs are calibrated. Also, the LPRM calibration frequency 
is not considered an initiator of any events analyzed in the SAR. 
Therefore, calibrating the LPRMs on a different frequency will not 
increase the probability of occurrence of an accident previously 
evaluated in the SAR.
    The resulting nodal power uncertainty does not exceed the nodal 
power uncertainty accounted for in the existing Minimum Critical 
Power Ratio (MCPR) Safety Limit; thus, the MCPR Safety Limit is not 
affected by this TS Change, and, therefore, the initial conditions 
of any accident are unchanged. Since the calibration frequency 
change will not affect the course of any evaluated accident, the 
consequences of an accident previously evaluated in the SAR will not 
be increased.
    Therefore, the proposed TS change does not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The change in the calibration frequency of the Local Power Range 
Monitor (LPRM) signal does not make any physical change to the plant 
or the manner in which the equipment responds to a transient or 
accident. The proposed TS change does not introduce a new mode of 
plant operation and does not involve the installation of any new 
equipment or instrumentation. The fuel will continue to be operated 
to the same safety limits since the Minimum Critical Power Ratio 
(MCPR) Safety Limit remains unchanged due to this TS change.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident, from any 
accident previously evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The following TS Bases were reviewed for potential reduction in 
the margin of safety:
    2.0 Safety Limits and Limiting Safety System Settings;
    3/4.1 Reactivity Control Systems;
    3/4.2.1 Average Planar Linear Heat Generation Rate;
    3/4.2.3 Minimum Critical Power Ratio:
    3/4.2.4 Linear Heat Generation Rate;
    3/4.3.1 Reactor Protection System Instrumentation;
    3/4.3.6 Control Rod Block Instrumentation;
    3/4.3.7.7 Traversing In-Core Probe System;
    The GE Thermal Analysis Basis (GETAB) determination of the 
Minimum Critical Power Ratio (MCPR) Safety Limit allows a maximum 
total nodal uncertainty of the Traversing In-Core Probe (TIP) 
readings of which the Local Power Range Monitor (LPRM).
    Update uncertainty is a part. The change in LPRM calibration 
frequency results in an LPRM Update uncertainty which, when combined 
with the other uncertainties which comprise the total TIP readings 
uncertainty, yields a total TIP readings nodal power uncertainty of 
less than the allowed GETAB uncertainty. Thus the change in LPRM

[[Page 64391]]

calibration frequency will not affect the MCPR Safety Limit.
    The LPRMs are utilized as input to the Average Power Range 
Monitor (APRM) and Rod Block Monitor (RBM) systems. The primary 
safety function of the APRM system is to initiate a scram during 
core-wide neutron flux transients before the actual core-wide 
neutron flux level exceeds the safety analysis design basis. This 
prevents fuel damage from single operator errors or equipment 
malfunctions. The APRMs are calibrated at least once per week to the 
plant heat balance, utilize a radially and axially diverse group of 
LPRMs as input and are utilized to detect changes in average, not 
local, power changes. Therefore, the effects of changing the LPRM 
calibration frequency on the APRM system responses will be minimal 
due to any individual LPRM drift being practically canceled out (due 
to diversity of input) and/or due to the frequent recalibration of 
the APRMs to an independent power calculation (the heat balance). 
Thus, changing the LPRM calibration frequency will not impact the 
capability of the APRM system to perform the scram function, and 
there is no impact on transient delta-CPRs.
    The RBM system is utilized in the mitigation of a Rod Withdrawal 
Error (RWE) event. The RBM system is designed to prevent the 
operator from increasing the local power significantly when 
withdrawing a control rod. Under Average Power Range Monitor - Rod 
Block Monitor Technical Specifications/Maximum Extended Load Line 
Limit Analysis (ARTS/MELLLA) on each selection of a control rod, the 
average of the assigned, unbypassed LPRMs is adjusted to equal a 
100% reference signal for each of the two RBM channels. Each RBM 
channel automatically limits the local thermal margin changes by 
limiting the allowable change in local average neutron flux to the 
RBM setpoint. If the local average neutron flux change is greater 
than that allowed by the RBM setpoint, within either RBM channel, 
the rod withdrawal permissive is removed preventing further rod 
movement. Since the change in local neutron flux is calculated from 
the change in the average of the LPRM readings, and calibrated on 
every rod selection to the reference signal, offsets in individual 
LPRM readings due to calibration differences are effectively 
eliminated for a given RBM setpoint. Therefore, the constraints on 
the withdrawal of any given rod are unchanged, and there will not be 
any increase in RWE delta-CPR.
    Since the MCPR Safety Limit is unaffected and the delta-CPR 
values are unchanged, the cycle CPR Operating limits are unchanged 
due to this TS change. Therefore, the proposed change in the 
frequency of LPRM signal calibration does not result in a reduction 
in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, PA 19101
    NRC Project Director: John F. Stolz

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

    Date of amendment request: May 3, 1996
    Description of amendment request: The proposed Technical 
Specifications (TS) changes would revise TS Surveillance Requirements 
4.6.5.3.a and 4.6.5.4.a to modify specific requirements to perform 
surveillance flow testing of the Standby Gas Treatment and Reactor 
Enclosure Recirculation Systems from monthly to quarterly.
    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 Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed TS changes do not involve any physical changes to 
plant systems or equipment. The proposed TS changes only change the 
Surveillance Requirements (SRs) surveillance test frequency 
pertaining to flow testing of the SGTS and RERS from monthly to 
quarterly. The periodic surveillance test frequencies provide 
adequate assurance that the equipment tested will remain in an 
operable condition. The test frequency interval for the flow testing 
of the SGTS and RERS was determined from the regulatory position in 
USNRC Regulatory Guide 1.52, ``Design, Testing, and Maintenance 
Criteria for Post Accident Engineered-Safety-Feature Atmosphere 
Clean-up System Air Filtration and Adsorption Units of Light-Water-
Cooled Nuclear Power Plants''. As stated in Regulatory Position 
C.4.d, ''... each Engineered Safety Feature (ESF) atmosphere cleanup 
train should be operated at least 10 hours per month, with the 
heaters on (if so equipped), in order to reduce the buildup of 
moisture on the absorbers and HEPA filters.''
    System operation on a monthly basis for the purpose of 
preventing moisture buildup on the absorbers as described in R.G. 
1.52 is not required at Limerick due to the continuous dry 
instrument air purge described previously in the Safety Assessment 
section of this submittal. Therefore a change in the interval 
between tests from monthly to quarterly will not result in moisture 
accumulation which would reduce the capability of the absorber to 
remove the iodine species from the exhaust air flow stream.
    The SGTS components are common to both units and must be run 
with the associated RERS for the surveillance test for each unit. 
The currently specified test frequency results in the SGTS being run 
at least twice per month or as many as eight (8) times per quarter 
for this surveillance, in addition to other required system 
surveillance tests which require the use of the components in this 
system. A change in surveillance test frequency from monthly to 
quarterly would reduce the wear on system components and thereby 
reduce the associated system downtime for maintenance and repairs. 
The consequent increased availability provides greater assurance 
that the system will be able to perform its mitigation function 
following any postulated accident.
    Surveillance test frequency on a quarterly interval is 
considered adequate to verify operability, as demonstrated by the 
required quarterly test interval for other equipment important to 
safety which have a similar function, such as the requirement for 
quarterly verification of the isolation time of the secondary 
containment and refueling area isolation valves, as required by LGS 
TS Sections 4.6.5.2.1 and 4.6.5.2.2.
    Therefore, the proposed TS changes do not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS changes only involve changes to the frequency in 
which the specified surveillances tests are performed. The proposed 
TS changes do not physically change the design or intended function 
of the systems, structures, or components associated with the SGTS 
or RERS. There will be no change to the existing redundancy of 
systems and components. The proposed change in surveillance test 
frequency will not introduce the possibility of any failure 
mechanisms of a different type than those already evaluated in the 
SAR. The existing components will not be used in any different 
manner and no new components will be added. Therefore with no 
physical changes and no new or different manner of system operation, 
no new failure mechanisms or equipment failure modes are created.
    Therefore, the proposed TS changes do not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The margin of safety as defined in the LGS TS Bases has not been 
reduced. The specific basis for the 31 day surveillance interval is 
not given in the LGS TS Bases section nor in the LGS UFSAR Sections 
6.5.1 or 9.4.2 which discuss the subject systems. However, 
Regulatory Position C.4.d of Regulatory Guide 1.52, Revision 2, 
relating to maintenance requirements, recommends:
    <gr-thn-eq>Each ESF atmosphere cleanup train should be operated 
at least 10 hours per month, with the heaters on (if so equipped), 
in order to reduce the buildup of moisture on the absorbers and HEPA 
filters.''

[[Page 64392]]

    The Bases for Surveillance Requirements (SR) 3.6.4.3.1 in the 
Standard Technical Specifications for General Electric Plants, BWR/
4, which corresponds to the subject LGS TS test, also notes the need 
for ten (10) hours of operation per month for elimination of 
moisture in the filters.
    The basis for the requirement for a monthly test with the 
heaters energized is clearly related to the desired elimination of 
moisture in the filters and absorbers. However, LGS UFSAR Table 6.5-
2 states that LGS does not conform to R.G. 1.52, Position C.4.d 
because the SGTS and RERS trains are ``continuously purged with dry 
instrumentation air to prevent build-up of moisture.'' UFSAR 
Sections 6.5.1.1.2 and 6.5.1.3.2 provide additional discussion of 
this method of moisture control.
    Therefore, the proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, PA 19101
    NRC Project Director: John F. Stolz

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

    Date of amendment request: September 27, 1996
    Description of amendment request: The proposed Technical 
Specifications (TS) changes would increase the Reactor Enclosure 
Secondary Containment maximum inleakage rate. This change will also 
impact secondary containment drawdown time and system flow rate 
assumptions, thereby, affecting charcoal filter bed efficiency and post 
accident dose analysis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Changing the Reactor Enclosure post drawdown inleakage rate from 
1250 cfm to 2500 cfm does not involve any changes to the function or 
operation of any plant component or safety related system. The 
Reactor Enclosure Recirculation System (RERS) and the Standby Gas 
Treatment System (SGTS) will maintain their design function by 
mitigating the radiological consequences of the analyzed accident 
and mitigating the post LOCA temperatures within the Reactor 
Enclosures. No analyzed accident initiating events are impacted, no 
new accident initiators are created, and no new failure modes are 
created. There are no changes to the redundancy, separation, quality 
assurance or fire protection requirements for the associated 
components and systems.
    The proposed changes to the LGS adsorber bed residence time will 
no longer fully meet the literal design guidance provided in 
Regulatory Guide (RG) 1.52, ``Design, Testing, and Maintenance 
Criteria for Post Accident Engineered-Safety-Feature Atmosphere 
Cleanup System Air Filter and Adsorption Units of Light-Water-Cooled 
Nuclear Power Plants,'' Revision 2, March 1978. This is because 
LGS's unique, yet more conservative, adsorber bed design is not 
addressed by the RG residence time design guidance. However, the LGS 
SGTS charcoal adsorbers still conform to the design function 
described in RG 1.52, based on the following: The LGS design with 
increased inleakage will continue to conform to the three conditions 
specified by RG 1.52, Position C.6.a, in order to maintain an 
assigned decontamination efficiency of 99%; there is a conservative 
amount of charcoal adsorber material provided by the LGS design, 
based on calculations performed in accordance with RG 1.3 
``Assumptions Used For Evaluating The Potential Radiological 
Consequences of a Loss of Coolant Accident For Boiling Water 
Reactors; and the LGS charcoal bed design is more conservative than 
the RG 1.52 design guidance, based on data (i.e., Iodine Penetration 
vs. Air Velocity) published by the charcoal manufacturer.
    Therefore, the probability of occurrence and the consequences of 
a malfunction of equipment important to safety is not increased. 
Also, the probability of occurrence of an accident previously 
evaluated is not increased. However, the proposed changes do affect 
the leak tightness of the Unit 1 and Unit 2 Reactor Enclosure, which 
increases the consequences of a postulated accident previously 
evaluated.
    Changing the Reactor Enclosure post drawdown inleakage rate from 
1250 cfm to 2500 cfm will result in an increase in the calculated 
LOCA/LOOP Design Basis Accident (DBA) off-site and on-site doses. 10 
CFR Part 100, and 10 CFR Part 50 Appendix A, General Design Criteria 
(GDC) 19, establish reference dose values used to determine site 
suitability and provide reasonable assurance that the facility can 
be operated following the analyzed accident without undue risk to 
the health and safety of the public. The proposed TS changes will 
increase the SGTS drawdown time from 2 minutes and 20 seconds to 15 
minutes and 30 seconds. The drawdown time increase will not prevent 
the RERS/SGTS from performing all of their safety related functions. 
However, because it is conservatively assumed that all radioactive 
material released during the drawdown period is unfiltered, and 
because the drawdown period has been extended whereby more 
unfiltered radioactive material is assumed to be released following 
the DBA, there is a corresponding increase in the calculated 
Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and 
Control Room doses. It is also assumed that the SGTS exhausts at the 
maximum inleakage rate throughout the entire DBA evaluation period 
(i.e., 30 days) where an increase in the maximum inleakage rate 
would also contribute to higher postulated EAB, LPZ, and Control 
Room doses. However, the proposed calculated doses do not exceed 10 
CFR Part 100, or 10 CFR Part 50, Appendix A, DGC 19 reference doses.
    Since the proposed doses resulting from the changes remain below 
10 CFR Part 100, and 10 CFR Part 50, Appendix A, these proposed 
changes will not significantly increase the consequences of an 
accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Changing the Reactor Enclosure post drawdown inleakage rate from 
1250 cfm to 2500 cfm is not an accident initiator nor does it result 
in the occurrence of an accident. The changes do not affect the 
function or operation of any plant component or safety related 
system nor do they create any new failure modes.
    In addition, the proposed changes do not involve any changes to 
the function or operation of any plant system or component nor will 
they adversely affect the Reactor Enclosure post LOCA environmental 
conditions. Furthermore, these changes will not create any new or 
different failure modes for the equipment important to safety within 
the Reactor Enclosure Secondary Containment.
    Therefore, the possibility of an accident of a different type or 
a different type of malfunction of equipment important to safety 
than previously evaluated is not created.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    Changing the Reactor Enclosure post drawdown inleakage rate from 
1250 cfm to 2500 cfm will result in reducing the margin of safety as 
defined in the LGS Updated Final Safety Analysis Report (UFSAR) 
relative to the off-site and on-site doses following a LOCA/LOOP 
DBA, and an increase of the UFSAR specified system drawdown time. 
From a system perspective, increasing the SGTS drawdown time from 2 
minutes and 20 seconds to 15 minutes and 30 seconds will not prevent 
the RERS/SGTS from performing all of their safety related functions. 
There will be a postulated increase in the corresponding EAB, LPZ, 
and Control Room doses, since it is assumed that fuel damage occurs 
coincident with the LGS DBA (i.e, at time = 0), all radioactive 
material released during the drawdown time is unfiltered, and the 
drawdown time is proposed to be extended whereby more unfiltered 
radioactive material could be released. It is also assumed that the 
SGTS exhausts at the maximum inleakage rate throughout the entire 
DBA evaluation period (i.e., 30 days) where an increase in the 
maximum inleakage

[[Page 64393]]

rate would also contribute to higher postulated EAB, LPZ, and 
Control Room doses. However, these calculated doses will remain 
below 10 CFR Part 100, and 10 CFR Part 50, Appendix A, GDC 19 
reference doses.
    Therefore, these proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, PA 19101
    NRC Project Director: John F. Stolz

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: October 1, 1996
    Description of amendment request: The proposed amendment would 
allow for a one-time extension of the surveillance intervals for the 
containment isolation valve (CIV) seat leakage test, the isolation 
valve seal water test, the boron injection tank leakage test, the 
containment spray nozzle test, and the city water backup to the 
auxiliary boiler feed pump test. These tests would be performed during 
the refueling outage scheduled to begin in April 1997.
    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:
    Regarding the Containment Isolation Valve seat leakage and 
Isolation Valve Seal Water tests:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response:
    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The probability of a previously evaluated accident will 
not increase because CIV leakage does not provide any role in 
accident initiation. The CIVs provide containment isolation 
following a design basis accident.
    The consequences of an accident previously evaluated will not 
significantly increase because the CIV leakage measurements contain 
significant margin to a more restrictive criteria based on the 
requested surveillance interval extension. As discussed in Section 
II, ``Evaluation of Changes,'' [see application dated October 1, 
1996] based on an evaluation of past CIV leak tests, the proposed 
change will not result in an increase in containment leakage because 
the measured leakage in previous CIV leak tests shows large margin 
to a more restrictive criteria based on the requested surveillance 
interval extension. Also, the latest test of IVSWS [isolation valve 
seal water system] satisfied the established acceptance criteria.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response:
    The proposed license amendment does not create the possibility 
of a new or different kind of accident from any previously 
evaluated. The proposed change only provides for a relatively short, 
one-time extension of the current leak-test interval for certain 
containment isolation valves. The proposed change does not involve 
the addition of any new or different type of equipment, nor does it 
involve operating equipment required for safe operation of the 
facility in a manner different from that addressed in the Final 
Safety Analysis Report. Therefore, the proposed change will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response:
    The proposed amendment does not involve a significant reduction 
in a margin of safety. The proposed change, for a one-time extension 
of the test interval, will not result in a significant reduction in 
a margin of safety because the test interval is being extended by 
only a short period and the measured leakage in previous CIV leak 
tests shows large margin to a more restrictive criteria based on the 
surveillance interval extension. In addition, the online leakage 
monitoring capability of the WCCPPS [weld channel containment 
penetration pressurization system] helps ensure that changes in CIV 
leakage during the extension period will be detected. Therefore, 
this change does not create a significant reduction in a margin of 
safety.
    Regarding the Boron Injection Tank (BIT) leakage test:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response:
    The proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The proposed change will not degrade the 
integrity of the BIT piping outside containment because no time 
dependent failure trends were observed in the review of past test 
results. The probability of a previously evaluated accident will not 
be increased because BIT leakage does not provide any role in 
accident prevention. The BIT leakage test only verifies that the BIT 
and associated piping meet specified leakage limits.
    The consequences of an accident previously evaluated will not 
significantly increase because the BIT leakage test results show 
large margins to the allowable leakage limit.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response:
    The proposed license amendment does not create the possibility 
of a new or different kind of accident from any previously 
evaluated. The proposed change does no[t] involve the addition of 
any new or different type of equipment, nor does it involve 
operating equipment required for safe operation of the facility in a 
manner that's different from that addressed in the Final Safety 
Analysis Report. Also, the increased surveillance interval (one-time 
only) will not adversely affect the integrity of the BIT piping and 
will not result in any new failure modes. Therefore, the proposed 
change will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response:
    The proposed license amendment does not involve a significant 
reduction in a margin of safety. Because of the large margin between 
the previous test and the allowable leak rate limit, the proposed 
change, for a one-time extension of the test interval, for the BIT 
leakage test does not adversely affect the performance of any safety 
related system, component, and does not result in increased severity 
of any of the accidents considered in the Final Safety Analysis 
Report. Based on past test results, the one-time extension of the 
leak test interval does not involve a significant reduction in a 
margin of safety.
    Regarding the Containment Spray Nozzle test:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response:
    The proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. As discussed in Section II, ``Evaluation of 
Changes,'' [see application dated October 1, 1996] based on an 
evaluation of past test results the proposed change will not degrade 
the reliability of the Containment Spray Nozzles because no time 
dependent failure trends were observed in the data review. The 
probability of a previously evaluated accident will not be increased 
because the Containment Spray Nozzles do not provide any role in 
accident prevention. The Containment Spray Nozzles provide a uniform 
spray distribution for containment cooling following postulated 
post-accident conditions.
    The consequences of an accident previously evaluated will not 
increase because the Containment Spray Nozzle reliability is not 
degraded by this change.

[[Page 64394]]

    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response:
    The proposed license amendment does not create the possibility 
of a new or different kind of accident from any previously 
evaluated. The proposed change does not involve the addition of any 
new or different type of equipment, nor does it involve operating 
equipment required for safe operation of the facility in a manner 
that is different from that addressed in the Final Safety Analysis 
Report. Also, the increased surveillance interval (one-time only) 
w[i]ll not adversely affect the functioning of the Containment Spray 
Nozzles and will not result in any new failure modes. Therefore, the 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response:
    The proposed license amendment does not involve a significant 
reduction in a margin of safety. The proposed change, for a one-time 
extension of the test interval, for the Containment Spray Nozzles 
does not adversely affect the performance of any safety related 
system, component, or instrument, or safety system setpoints and 
does not result in increased severity of any of the accidents 
considered in the Final Safety Analysis Report. Based on past test 
results, the one-time extension of the functional test interval will 
not adversely affect the functioning of the Containment Spray 
Nozzles. Therefore, this change does not create a significant 
reduction in a margin of safety.
    Regarding the City Water Backup test:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response:
    The proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The proposed change will not degrade the 
reliability of the City Water Backup Supply Valves for the AFW 
[auxiliary feedwater] System because no time dependent failure 
trends were observed in the review of past test results. The 
probability of a previously evaluated accident will not increase 
because the City Water Backup Supply Valves for the AFW System do 
not provide any role in accident prevention. The City Water Backup 
Supply Valves for the AFW System only provides a diverse source of 
water for the AFW system.
    The consequences of an accident previously evaluated will not 
significantly increase because the City Water Backup Supply Valves 
for the AFW System are not assumed to function to mitigate any 
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:
    The proposed license amendment does not create the possibility 
of a new or different kind of accident from any previously 
evaluated. The proposed change does not involve the addition of any 
new or different type of equipment, nor does it involve operating 
equipment required for safe operation of the facility in a manner 
that is different from that addressed in the Final Safety Analysis 
Report. Also, the increased surveillance interval (one-time only) 
will not adversely affect the functioning of the City Water Backup 
Supply Valves for the ABFP [auxiliary boiler feedpump] and will not 
result in any new failure modes. Therefore, the proposed change will 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response:
    The proposed amendment does not involve a significant reduction 
in a margin of safety. The proposed change, for a one-time extension 
of the test interval, for the City Water Backup Supply Valves for 
the ABFP does not adversely affect the performance of any safety 
related system, component, or instrument, or safety system setpoints 
and does not result in increased severity of any of the accidents 
considered in the Final Safety Analysis Report. Based on past test 
results, the one-time extension of the functional test interval will 
not adversely affect the functioning of the City Water Backup Supply 
Valves for the AFW System. Therefore, this change does not create 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 Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle, 
New York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Acting

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

    Date of amendment request: October 25, 1996
    Description of amendment request: The proposed change to Hope Creek 
Technical Specification (TS) 3/4.1.3.5, ``Control Rod Scram 
Accumulator'', would: 1) permit a separate entry into a Technical 
Specification action statement for each inoperable control rod; 2) 
provide more specific applicability for required actions in operational 
condition 1 or 2 with one inoperable control rod scram accumulator 
(reactor pressure of greater than or equal to 900 psig would be 
specified); 3) provide more specific actions (verify charging water 
pressure) for two or more inoperable control rod scram accumulators and 
reactor pressure is greater than or equal to 900 psig; 4) provide more 
specific actions when reactor pressure is less than 900 psig and one or 
more control rod scram accumulators are inoperable (verify insertion of 
control rods associated with inoperable accumulators and verify that 
charging water header pressure is greater than or equal to 940 psig); 
and 5) provide specific actions in operational condition 5 with one or 
more withdrawn control rods inoperable; and 6) eliminate the 
requirements to perform a 18-month channel functional test of the leak 
detectors and the 18-month channel calibration of the pressure 
detectors.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The change incorporates the appropriate content of the improved 
BWR/4 Standard Technical Specifications, NUREG-1433, for Control Rod 
Scram Accumulators.
    The proposed Technical Specification and required Action 
completion times are consistent with or more conservative than those 
approved for use in the improved Technical Specifications for 
inoperable control rod scram accumulators. In addition, the proposed 
surveillance requirements for the control rod scram accumulators are 
sufficient to adequately demonstrate operability as stated in the 
Bases for the improved Technical Specifications. Further, the 
proposed changes enhance the current Hope Creek Technical 
Specifications by reflecting improved techniques collectively 
learned by the industry. Therefore, the proposed changes do not 
significantly increase the risk or consequences of any accidents 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Neither the mechanism for initiating or completing a scram is 
modified by this proposed change. There are no physical changes to 
plant equipment proposed in the application. The proposed change 
does not create a means by which the scram function could be impeded 
or prevented. The proposed change is functionally equivalent to the 
current Technical Specifications, but provides additional 
operational flexibility to diagnose and resolve equipment issues 
that do not impact operability of the control rods before taking 
proscriptive actions which

[[Page 64395]]

result in significant plant transients (i.e. full power scram).
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The operability of the accumulators and the scram function of 
the control rod drive system protects the Safety Limit Minimum 
Critical Power Ratio as well as the 1% cladding plastic strain fuel 
design limit. The proposed change does not reduce a margin of safety 
as defined in the Bases of the Technical Specification since the 
proposed change does not affect the maximum allowable scram times 
for control rods, nor does it change the maximum allowable number or 
minimum separation of inoperable control rods. The proposed change 
does not modify any instrument setpoints or functions. The proposed 
change will either maintain the present margins of safety or 
increase them, by reducing the need for unnecessary challenges to 
the reactor protection system and resulting plant shutdowns, while 
still maintaining the capability to complete a reactor scram.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. 
Ginna Nuclear Power Plant, Wayne County, New York

    Date of amendment request: October 29, 1996
    Description of amendment request: The proposed amendment would 
revise the mode of applicability for the motor-driven auxiliary 
feedwater (AFW) pump actuation on opening of the main feedwater (MFW) 
pump breakers to correct an error introduced during Amendment No. 61.
    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 less restrictive changes discussed in Section C.1 [of the 
licensee's application] do not involve a significant hazards 
consideration as discussed below:
    1. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. The proposed 
changes only correct an error which was introduced in Amendment No. 
61 to the Ginna Station technical specifications. The changes revert 
the mode of applicability for the motor-driven AFW pump actuation on 
the opening of the MFW pump breakers to what existed previously. The 
change is essentially correction of a typographical error that was 
caused through use of the electronic version of NUREG-1431 in 
preparation of the Ginna Station ITS [Improved Technical 
Specifications]. There have been no subsequent plant modifications 
or changes to the accident analysis which would invalidate the 
previous NRC acceptance of only requiring this Function above 5% 
power. The accident analyses do not credit automatic initiation of 
AFW on MFW pump trip in MODE 2. As such, these changes do not impact 
initiators or analyzed events or assumed mitigation of accident or 
transient events. Therefore, these changes do not involve a 
significant increase in the probability or consequences of an 
accident previously analyzed.
    2. Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind of 
accident from any accident previously evaluated. The proposed changes 
do not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or changes in the 
methods governing normal plant operation which existed prior to 
Amendment No. 61. The proposed changes will not impose any new or 
different requirements. Thus, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. The proposed changes will not reduce a margin of plant 
safety because there have been no subsequent plant modifications or 
changes to the accident analysis which would invalidate the previous 
NRC acceptance of only requiring this Function above 5% power. As 
such, no question of safety is involved, and the change does not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610
    Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005
    NRC Project Director: S. Singh Bajwa, Acting

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. 
Ginna Nuclear Power Plant, Wayne County, New York

    Date of amendment request: October 29, 1996
    Description of amendment request: The proposed amendment would 
revise the Required Actions for the auxiliary feedwater (AFW) pump 
actuation on Steam Generator Level (SG) - Low Low logic to be 
consistent with those specified in NUREG-1431.
    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 less restrictive changes discussed in Section C.1 [of the 
licensee's application] do not involve a significant hazards 
consideration as discussed below:
    1. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. The proposed 
changes with respect to the Required Actions for AFW actuation on SG 
Level - Low Low logic provide consistency with NUREG-1431 by 
requiring an inoperable channel to be placed in the tripped 
condition within 6 hours. The affected logic then requires 1 of 2 
channels in order to actuate such that there is no impact on any 
initiators or analyzed events or assumed mitigation of accident or 
transient events. Therefore, these changes do not involve a 
significant increase in the probability or consequences of an 
accident previously analyzed.
    2. Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
changes do not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) or changes in 
the methods governing normal plant operation. The proposed changes 
will not impose any new or different requirements. Thus, this change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. The proposed changes will not reduce a margin of plant 
safety because the AFW actuation on SG Level - Low Low still remains 
capable of performing its function with an inoperable channel placed 
in the tripped configuration. These changes are also consistent with 
those provided in NUREG-1431. As such, no question of safety is 
involved, and the change does not involve a significant reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three

[[Page 64396]]

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:  Rochester Public Library, 115 
South Avenue, Rochester, New York 14610
    Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005
    NRC Project Director: S. Singh Bajwa, Acting

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: September 4, 1996
    Description of amendment request: The proposed amendment to the 
Technical Specifications would allow the use of four lead test 
assemblies (advanced zirconium-based alloys) in the North Anna, Units 1 
and 2, reactor cores.
    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:
    Operation of the four FCF [Framatome Cogema Fuels] lead test 
assemblies will not:
    1.Involve a significant increase in the probability of 
occurrence or the consequences of an accident previously evaluated. 
The FCF lead test assemblies are very similar in design to the 
Westinghouse fuel that comprises the remainder of the core. The 
reload core design for North Anna cycles which incorporate the lead 
test assemblies will meet all applicable design criteria. In 
addition, the performance of the ECCS [emergency core cooling 
system] at North Anna Units 1 and 2 will not be affected by the 
insertion of the four lead test assemblies, so the criteria of 10 
CFR 50.46 will be satisfied for use of these assemblies with fuel 
rods, guide thimble tubes, and instrumentation tubes fabricated with 
advanced zirconium-based alloys. The use of these fuel assemblies 
will not result in a change to the North Anna Units 1 and 2 reload 
design and safety analysis limits. The existing safety analyses 
based on the resident Westinghouse fuel will remain applicable for 
cycles which incorporate the lead test assemblies. Therefore, 
neither the probability of occurrence nor the consequences of any 
accident previously evaluated is significantly increased.
    2. Create the possibility for a new or different type of 
accident from any accident previously evaluated. The FCF lead test 
assemblies are very similar in design (both mechanical and 
composition of materials) to the resident Westinghouse fuel. North 
Anna cores which incorporate the lead test assemblies will be 
designed to meet all applicable design criteria and ensure that all 
pertinent licensing basis criteria are met. Demonstrated adherence 
to these standards and criteria precludes new challenges to 
components and systems that could introduce a new type of accident. 
North Anna safety analyses based on the resident Westinghouse fuel 
will remain applicable for cores containing the lead test 
assemblies. All design and performance criteria will continue to be 
met and no single failure mechanisms have been created. In addition, 
the use of these fuel assemblies does not involve any alteration to 
plant equipment or procedures which would introduce any new or 
unique operational modes or accident precursors. Therefore, the 
possibility for a new or different kind of accident from any 
accident previously evaluated is not created.
    3. Involve a significant reduction in the margin of safety. The 
use of the FCF lead test assemblies does not change the performance 
requirements on any system or component such that any design 
criteria will be exceeded, and will not cause the core to operate in 
excess of pertinent design basis operating limits. North Anna reload 
core designs for cycles which incorporate the lead test assemblies 
will specifically evaluate any pertinent differences between the 
lead test assemblies and the resident fuel, and will take into 
consideration the normal core operating conditions allowed in the 
Technical Specifications. Safety analyses based on the resident 
Westinghouse fuel will remain applicable for cores incorporating the 
FCF lead test assemblies. Analyses or evaluations will be performed 
each cycle to confirm that the criteria in 10 CFR 50.46 will be met. 
Therefore, the margin of safety as defined in the Bases to the North 
Anna Units 1 and 2 Technical Specifications is not significantly 
reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Acting Project Director: Mark Reinhart

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: November 6, 1996
    Description of amendment request: The proposed changes will modify 
the requirements for isolated loop startup to permit filling of a 
drained isolated loop via backfill from the reactor coolant system 
through partially open stop loop valves.
    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:
    Specifically, operation of the North Anna Power Station [in] 
accordance with the proposed changes will not:
    1. Involve a significant increase in the probability of 
occurrence or consequences of an accident previously evaluated. The 
probability of occurrence of a positive reactivity addition accident 
is not being increased by the proposed Technical Specification 
change. The proposed restrictions on boron concentration and mixing, 
reactor coolant system inventory and reactivity and count rate 
monitoring provide a level of protection against reactivity addition 
accidents which is equivalent to that currently in place.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated. The proposed change does not 
introduce any new or unique failure modes or accident precursors. 
Eliminating the operability requirements for the loop stop valve 
interlocks does not create any new or different kind of accident 
scenario. Loop startup accidents in the various modes of operation 
have been analyzed. Operation of the loop stop valves will not 
change. New requirements have been imposed for the case of 
backfilling a drained loop from the reactor coolant system to ensure 
that core cooling and reactivity control are preserved throughout 
the backfill evolution.
    3. Involve a significant reduction in any margin of safety. The 
new Technical Specification loop isolation and startup requirements 
for temperature, boron concentration, and shutdown margin fulfill 
the function of the loop stop valve interlocks. Therefore, the 
margin of safety as defined in any Technical Specification bases is 
not reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: Mark Reinhart (Acting)

[[Page 64397]]

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

    Date of amendment request: October 31, 1996
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications 
(TS) by deleting the requirement for an annual submittal of a 
description of changes made pursuant to 10 CFR 50.59. Consistent with 
10 CFR 50.59(b)(2), a description of changes will subsequently be 
included with the KNPP Updated Safety Analysis Report (USAR) update in 
accordance with 10 CFR 50.71(e). Additionally, the proposed amendment 
would correct minor administrative inconsistencies in the TS Table of 
Contents and in a footnote reference.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c)The NRC staff's review is 
presented below:
    On August 31, 1992 (57 FR 39353), the NRC amended 10 CFR 
50.59(b)(2) to reduce the regulatory burden on nuclear licensees. This 
action revised the requirements for the annual submission of reports 
for facility changes under 10 CFR 50.59. This action did not affect the 
substance of the evaluation or the documentation required for 10 CFR 
50.59 type changes. It only affected the interval for submission of the 
information to the NRC. Instead of submitting the information annually, 
the information can be submitted on a refueling cycle basis, provided 
the interval between successive reports does not exceed 24 months.
    In order to take advantage of this reduction in regulatory burden, 
the licensee has proposed an amendment to remove the submittal of a 
report of facility changes under 10 CFR 50.59 from the Technical 
Specification list of annual reporting requirements. Additionally, the 
licensee has proposed corrections to minor administrative 
inconsistencies in the TS Table of Contents and in a footnote 
reference. The proposed changes are administrative only and do not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated; or
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated; or
    3. Involve a significant reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location:  University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497
    NRC Project Director: Gail H. Marcus

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.

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

    Date of application for amendment: May 1, 1996, as supplemented 
August 12, 1996.
    Brief description of amendment: The amendment approves relocation 
of the administrative controls related to the quality assurance review 
and audit requirements of Section 6, Technical Specifications 6.5.B.8, 
``Nuclear Safety Review and Audit Committee-Audits,'' from the Pilgrim 
Station Technical Specifications to the Boston Edison Quality Assurance 
Manual (BEQAM). This change is in accordance with the guidance 
contained in NRC Administrative Letter 95-06, ``Relocation of Technical 
Specification Administrative Controls Related to Quality Assurance.'' 
In addition, the Safety Evaluation includes the NRC staff review and 
approval of the BEQAM changes in support of this amendment.
    Date of issuance: November 12, 1996
    Effective date: November 12, 1996
    Amendment No.: 168
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28605) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 12, 1996. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

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

    Date of application for amendments: August 29, 1996, as 
supplemented on September 20, 1996, and October 4, 1996.
    Brief description of amendments: The amendments change the 
Technical Specifications to implement 10 CFR Part 50, Appendix J, 
Option B, by referring to Regulatory Guide 1.163, ``Performance-Based 
Containment Leakage-Test Program,'' with an exception as detailed in 
the licensee's application.
    Date of issuance: November 12, 1996
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 175 and 162

[[Page 64398]]

    Facility Operating License Nos. DPR-39 and DPR-48: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 9, 1996 (61 FR 
52964). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 12, 1996. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.

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

    Date of application for amendment: August 14, 1996, as supplemented 
October 18, 1996, and related application of January 18, 1996
    Brief description of amendment: The amendment revises the technical 
specifications (TS) to allow one-cycle deferral of the inspection of 
reactor coolant pump (RCP) flywheels.
    Date of issuance: November 7, 1996
    Effective date: November 7, 1996
    Amendment No.: 175
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 24, 1996 (61 
FR 50054) The October 18, 1996, letter provided an updated TS page. 
This change was within the scope of the original application and did 
not change the staff's initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated November 7, 1996.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423

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

    Date of application for amendments: December 14, 1994, as 
supplemented by letters dated May 16 and August 29, 1996
    Brief description of amendments: The amendments will incorporate 
guidance and recommendations for diesel generators contained in NUREG-
1366, ``Improvements to Technical Specifications Surveillance 
Requirements,'' Generic Letter (GL) 93-05, ``Line-Item Technical 
Specifications Improvements to Reduce Surveillance Requirements for 
Testing During Power Operations,'' GL 94-01, ``Removal of Accelerated 
Testing and Reporting Requirements for Emergency Diesel Generators,'' 
and NUREG-1431, ``Revised Standard Technical Specifications for 
Westinghouse PWRs.''
    Date of issuance: November 12, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 170 and 152
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28612) The August 29, 1996, letter provided clarifying information that 
did not change the scope of the December 14, 1996, application and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 12, 1996.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223

Entergy 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 1, 1996
    Brief description of amendment: The amendment revises the technical 
specifications to incorporate requirements for limiting the time that 
the hydrogen mixing isolation valves on the drywell are open. The 
amendment also changes the time from 7 days to 31 days to determine the 
cumulative time the valves are open.
    Date of issuance: November 12, 1996
    Effective date: November 12, 1996
    Amendment No.: 89
    Facility Operating License No. NPF-47. The amendment revised the 
Technical Specifications/operating license.
    Date of initial notice in Federal Register: September 25, 1996 (61 
FR 50343) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 12, 1996.No significant 
hazards consideration comments received. No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803

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

    Date of application for amendment: May 9, 1996, as supplemented by 
letter dated August 27, 1996.
    Brief description of amendment: The amendment changed Surveillance 
Requirements (SRs) 3.4.4.3, Safety/Relief Valves, 3.5.1.7, Automatic 
Depressurization System Valves, and 3.6.1.6.1, Low-Low Set Valves, of 
the Technical Specifications and allows the licensee to perform the 
surveillance of the relief mode of operation of the safety/relief 
valves on the main steam lines without physically lifting the disk of a 
valve off the seat at power. The changes stated that the required 
operation of the valve to verify is that the relief-mode actuator 
strokes when the valve is manually actuated and the frequency of the 
surveillances are in accordance with the inservice testing program for 
the valves.
    Date of issuance: November 18, 1996
    Effective date: November 18, 1996
    Amendment No: 130
    Facility Operating License No. NPF-29. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 11, 1996 (61 
FR 47971) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 18, 1996. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120.

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 3, 1996, as supplemented 
October 23, 1996
    Brief description of amendment: The amendment clarifies a 
restriction on shutdown margin monitor operability while changing 
operational modes, so that it only limits reactivity changes caused by 
boron dilution and rod withdrawal. The amendment also corrects a 
technical specification numerical reference so that the specification 
number cited is in agreement with Amendment 99, dated December 29, 
1994.
    Date of issuance: November 14, 1996

[[Page 64399]]

    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 131
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 20, 1996 (61 FR 
31559) The October 23, 1996, letter provided clarifying information 
that did not change the scope of the June 3, 1996, application and the 
initial proposed no significant hazards consideration determination.The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated November 14, 1996No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385.

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

    Date of application for amendment: May 30, 1996
    Brief description of amendment: The proposed change to the 
anticipated transient without scram recirculation pump trip logic for 
the James A. Fitzpatrick Nuclear Power Plant allows for a high pressure 
trip setpoint which is dependent upon the number of safety/relief 
valves which are out of service.
    Date of issuance: November 7, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 237
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34896) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 7, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: May 30, 1996, as supplemented 
October 17, and November 8, 1996
    Brief description of amendment: The proposed amendment changes the 
FitzPatrick safety limit minimum critical power ratio from its current 
value of 1.07 for two recirculation loop operation to 1.09 and from 
1.08 to 1.10 for single recirculation loop operation for the Cycle 13 
operation.
    Date of issuance: November 14, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 238
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34896) The October 17 and November 8, 1996 letters provided 
supplemental information that did not change the initial no significant 
hazards consideration determination.The Commission's related evaluation 
of the amendment is contained in a Safety Evaluation dated November 14, 
1996.No significant hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph 
M. Farley Nuclear Plant, Unit 2, Houston County, Alabama

    Date of amendment request: August 23, 1996, as supplemented by 
letters dated September 16, November 6, 11 and 14, 1996
    Brief description of amendment: The amendment changes the Technical 
Specifications (TS) to allow installation of laser welded elevated 
tubesheet sleeves. Specifically, the amendment is for one cycle only 
for Farley Unit 2. Permanent, generic TS changes for Westinghouse laser 
welded sleeves for both units will be submitted prior to the next Unit 
1 refueling outage currently scheduled for spring 1997.
    Date of issuance: November 20, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment No.: 117
    Facility Operating License No. NPF-8: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 11, 1996 (61 
FR 47982) The September 16, November 6, 11 and 14, 1996, letters 
provided clarifying information that did not change the scope of the 
August 23, 1996, application and the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 20, 1996.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302

Southern California Edison Company, et al., Docket Nos. 50-361 and 
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, 
San Diego County, California

    Date of application for amendments: July 17, 1995.
    Brief description of amendments: These amendments revise the 
frequency of surveillance requirements for certain plant protective 
system instrumentation contained in Technical Specifications (TS) 
3.3.1, ``Reactor Protective System (RPS) Instrumentation - Operating,'' 
TS 3.3.2, ``Reactor Protective System (RPS) Instrumentation - 
Shutdown,'' TS 3.3.3, ``Control Element Assembly Calculators (CEACs),'' 
TS 3.3.4, ``Reactor Protective System (RPS) Logic and Trip 
Initiation,'' TS 3.3.5, ``Engineered Safety Features Actuation System 
(ESFAS) Instrumentation,'' and TS 3.3.6, ``Engineered Safety Features 
Actuation System (ESFAS) Logic and Manual Trip.''
    Date of issuance: November 18, 1996
    Effective date: November 18, 1996, to be implemented within 30 days 
of the date of issuance.
    Amendment Nos.: Unit 2 - 133 ; Unit 3 - 122
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 30, 1995 (60 FR 
45185) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 18, 1996.No significant 
hazards consideration comments received: No.Temporary
    Local Public Document Room location:  Science Library, University 
of California, P. O. Box 19557, Irvine, California 92713

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

    Date of application for amendment: September 4, 1996
    Brief description of amendment: This amendment revises Technical

[[Page 64400]]

Specification (TS) 6.2.3, ``Facility Staff Overtime,'' by removing 
specific overtime limits and working hours and by adding procedural 
controls to perform a monthly review of overtime hours.
    Date of issuance: November 8, 1996
    Effective date: November 8, 1996, to be implemented not later than 
90 days after issuance
    Amendment No.: 212
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 9, 1996 (61 FR 
52970) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 8, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606

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

    Date of application for amendment: July 18, 1996
    Brief description of amendment: The amendment adopts ASTM D-3803-
1989 as the laboratory testing standard for charcoal samples from the 
charcoal adsorbers in the auxiliary/fuel building emergency exhaust 
system.
    Date of issuance: November 13, 1996
    Effective date: November 13, 1996, to be implemented within 30 days 
of the date of issuance.
    Amendment No.: 118
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 14, 1996 (61 FR 
42285) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 13, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Local Public Document Room location:  Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Dated at Rockville, Maryland, this 26th day of November 1996.
    For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II Office of Nuclear Reactor 
Regulation
[Doc. 96-30714 Filed 12-3-96; 8:45 am]
BILLING CODE 7590-01-F