[Federal Register Volume 61, Number 216 (Wednesday, November 6, 1996)]
[Notices]
[Pages 57481-57497]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-11106]


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NUCLEAR REGULATORY COMMISSION
Biweekly Notice


Applications and Amendments to Facility Operating 
LicensesInvolving 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 October 11, 1996, through October 25, 1996. 
The last biweekly notice was published on October 23, 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 December 6, 1996, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be

[[Page 57482]]

made a party to the proceeding; (2) the nature and extent of the 
petitioner's property, financial, or other interest in the proceeding; 
and (3) the possible effect of any order which may be entered in the 
proceeding on the petitioner's interest. The petition should also 
identify the specific aspect(s) of the subject matter of the proceeding 
as to which petitioner wishes to intervene. Any person who has filed a 
petition for leave to intervene or who has been admitted as a party may 
amend the petition without requesting leave of the Board up to 15 days 
prior to the first prehearing conference scheduled in the proceeding, 
but such an amended petition must satisfy the specificity requirements 
described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. 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.

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert 
Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

    Date of amendment request: October 3, 1996
    Description of amendment request: The proposed amendment changes 
the provision for receiving, possessing and using byproducts, source 
and special nuclear material at Calvert Cliffs Unit 1.

    Currently, Unit 1 is licensed under 10 CFR Part 30 to receive, 
possess, and use 100 millicuries of byproduct material for sample 
analysis or instrument calibration, 500 millicuries of byproduct 
material in the form of equipment; and 500 millicuries of Sodium-24 for 
steam turbine acceptance testing. In addition, Unit 1 is licensed to 
receive, possess and use 100 milligrams each of source or special 
nuclear material under 10 CFR Parts 40 and 70. Unit 2 is licensed under 
10 CFR Parts 30, 40, and 70 to receive, possess, and use in amounts as 
required any byproduct, source, or special nuclear material for sample 
analysis or instrument calibration or associated with radioactive 
apparatus or components. This proposed amendment would change the Unit 
1 license to be consistent with the Unit 2 license by replacing license 
conditions 2.B.3 and 2.B.4 with the same wording as Unit 2's license 
condition 2.B.4.
    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. Would not involve a significant increase in the probability 
or consequences of an accident.
    Currently, Unit 1 is licensed under 10 CFR Part 30 to receive, 
possess, and use 100 millicuries of byproduct material for sample 
analysis or instrument calibration, 500 millicuries of byproduct 
material in the form of equipment; and 500 millicuries of Sodium-24 
for steam turbine acceptance testing. Unit 1 is also licensed under 
10 CFR parts 40 and 70 to receive, possess, and use 100 milligrams 
of source or special nuclear material. Unit 2 is licensed under 10 
CFR Parts 30, 40, and 70 to receive, possess, and use in amounts as 
required any byproduct, source, or special nuclear material for 
sample analysis or instrument calibration or associated with 
radioactive apparatus or components. This proposed amendment would 
change the Unit 1 license to be consistent with the Unit 2 license. 
The reason for this proposed change is that it is sometimes 
necessary to receive and use byproduct material, sources, or special 
nuclear material with different activity levels, and in different 
quantities than is specified by the Unit 1 license.
    The current licenses for the two units allow radioactive 
materials to be accepted and used at Unit 2, although these same 
materials would not be acceptable for use at Unit 1. These 
byproduct, source, and special nuclear materials are used by the 
same people and for the same function in either

[[Page 57483]]

unit. Training and procedures for handling radioactive material have 
been developed and used at both Units over the last 20 years. These 
procedures are adequate to control the acceptance and use of 
radioactive material at Unit 2 and, therefore, adequate to control 
radioactive material at Unit 1.
    Receiving, possessing, and using byproduct, source, or special 
nuclear material is not related to accident conditions. Therefore, 
changing the Unit 1 license conditions to be the same as the Unit 2 
license condition does not involve a significant increase in the 
probability or consequences of an accident.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    Procedures and training governing the acceptance and use of 
radioactive materials are the same for both Unit 1 and Unit 2. These 
procedures will not be changed as a result of this license change. 
In addition, receiving, possessing, and using radioactive material 
is not related to accident conditions. Therefore, making the Unit 1 
license the same as the Unit 2 license will not create the 
possibility of a new or different type of accident from any accident 
previously evaluated.
    3. Would not involve a significant reduction in the margin of 
safety.
    The margin of safety in this case is exposure to contaminated 
material or equipment. Exposure is controlled by adequate training 
and procedures. Radioactive material is received by personnel 
assigned to the Radiation Safety Section. These personnel are 
trained in receiving and shipping contaminated material. Once the 
material is onsite, it becomes the responsibility of the radiation 
protection staff who are trained in the handling of all levels of 
radioactive material. Training and procedures for handling 
radioactive materials have been developed and used over the 20-year 
life of the plant, and are currently deemed adequate for compliance 
with the Unit 2 license. Therefore, making the Unit 1 license the 
same as the Unit 2 license will not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: S. Singh Bajwa, Acting Director

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

    Date of amendment request: September 20, 1996
    Description of amendment request: The proposed amendments would add 
a footnote to specification 4.3.1.B.4.A.10.a which refers to a letter 
that describes enhancements made to the Combustion Engineering sleeve 
installation process.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment continues to allow the Combustion 
Engineering sleeves to be used as an alternate tube repair method 
for Zion steam generators, along with the process enhancements which 
are described in the letter identified in the proposed Technical 
Specification note. The sleeve configuration, which was designed and 
analyzed in accordance with the criteria of Regulatory Guide (RG) 
1.121 and Section III of the ASME Code, is unaffected by the 
enhancements. Fatigue and stress analyses of the sleeved tube 
assemblies as described in the currently approved Topical Report, 
CEN-331-P, Revision 1-P, are unaffected by the enhancements.
    Mechanical testing which has shown that the structural integrity 
of the sleeves under normal, faulted, and upset conditions is within 
the acceptable limits and is unaffected by the enhancements. Leakage 
rate testing for the tube sleeves which has demonstrated that 
primary to secondary leakage is not expected during any plant 
condition is unaffected by the enhancements. The consequences of 
leakage through the sleeved region of the tube, including the 
enhancements, is bounded by the existing steam generator tube 
rupture (SGTR) analysis included in the Zion Updated Final Safety 
Analysis Report.
    The proposed Technical Specification change reflects 
enhancements to the installation and inspection process identified 
in Topical Report CEN-331-P, Revision 1-P, which is currently 
referenced in the Technical Specifications. These enhancements do 
not increase the probability or consequences of an accident 
previously evaluated. The enhancement which disallows the 
installation of the tube plugs made from Inconel 600 material was 
done so based upon industry information and is addressed by NRC 
Bulletin 89-01. The use of the Plus Point Probe, its associated data 
acquisition equipment, and improved visual inspection equipment, are 
conservative actions and improve the quality of the sleeving 
process. The use of the mechanical plug in lieu of the welded plug 
meets the established design requirements and is advantageous in the 
area of dose reduction, because of reduced time to install. Minor 
changes to the sleeve installation equipment as described in the 
Topical Report, represent equipment enhancements and do not alter 
the sleeve design or qualification testing.
    The proposed Technical Specification change does not adversely 
impact any previously evaluated design basis accident. Installation 
of the sleeves, with the described enhancements, can be used to 
repair degraded tubes by returning the condition of the tubes to 
their original design basis condition for tube integrity and leak 
tightness during all plant conditions. Therefore, the currently 
approved sleeving process with the described enhancements will not 
increase the probability of occurrence of an accident previously 
evaluated.
    Therefore, these proposed changes do 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 implementation of the enhancements to the proposed sleeving 
process will not affect the plant design basis. The current stress 
and fatigue analyses of the repair identified in Topical Report CEN-
331-P, Revision 1-P, has shown the ASME Code and RG 1.121 allowable 
values are met and are unaffected by the described enhancements. The 
current sleeving design, with the described enhancements, will 
continue to maintain overall tube bundle structural integrity and 
leak tightness at a level consistent with that of the originally 
supplied tubing. Leak and mechanical testing of the sleeves, are 
unaffected by the proposed enhancements and continue to support the 
conclusions that the sleeve retains both structural integrity and 
leak tightness during all operating and accident conditions. Repair 
of a tube with a sleeve, utilizing the described enhancements, does 
not provide a mechanism that results in an accident outside of the 
area affected by the sleeve.
    The described change to implement the cited enhancements will 
not create a new or different type of accident. The change only 
reflects enhancements to the currently approved installation/
inspection process and, would not change or impact any hypothetical 
accident previously discussed. Use of improved Non-Destructive 
Examination, data acquisition and visual inspection equipment 
improves the quality of the sleeving process and has no negative 
effect on the margin of safety. The elimination of the use of the 
Inconel 600 plug also improves the margin of safety.
    Any hypothetical accident as a result of potential tube or 
sleeve degradation in the repaired portion of the tube is bounded by 
the existing SGTR analysis. The sleeve design, including described 
enhancements, does not affect any other component, or affect any 
location on the tube outside of the immediate area repaired.
    Therefore, the proposed changes do not create the possibility of 
a new or different type of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.

[[Page 57484]]

    The currently approved sleeving repair of degraded steam 
generator tubes has been shown by analysis to restore the integrity 
of the tube bundle to its original design basis condition. By 
implementing the described enhancements, the consistent quality of 
the upper sleeve weld has increased thereby reducing the potential 
for rework and reducing the potential for leaving a weld indication 
in service.
    The proposed change does not involve a reduction in the margin 
of safety. The change reflects enhancements to the installation/
inspection processes which are currently referenced in the Technical 
Specifications. These enhancements would not have any adverse 
effects on the previously evaluated design transient or accident 
analyses. The enhancements represent acceptable industry standards.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: August 21, 1996
    Description of amendment request: The proposed amendment will 
modify Containment Penetrations Nos. 53 and 65 design by modifying the 
design of instrumentation lines for Containment Vacuum Relief (CVR) 
system that pass through these containment penetrations. The proposed 
change will correct the error in previously docketed information that 
was used by NRC during licensing process.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change will not increase the probability of 
previously analyzed accidents. The proposed change seeks to clearly 
document the design and licensing bases for acceptance of the CVR 
sensing instrument lines. The proposed change to the monitoring 
lines will provide greater assurance that containment integrity will 
be maintained following a LOCA concurrent with a single active 
failure. The design change to the non-essential monitoring line will 
reduce the potential bypass leakage from penetrations 53 and 65 by 
adding a redundant automatic containment isolation valve on 
penetration 53 and isolating the non-essential instrument line on 
penetration 65. This design change can be performed at power without 
violating any license/regulatory requirements that ensure 
containment integrity is maintained.
    There is no change in the function of the instrumentation. The 
only difference is that CVR-IDPT-5017B and C non-safety differential 
transmitters that monitor the CVR system will be sensing containment 
pressure from penetration 53. If the non-essential line coming from 
penetration 53 becomes inoperable, containment to annulus 
differential pressure can be obtained from alternate 
instrumentation. The essential sensing line that actuates the CVR 
system to protect containment within design vacuum pressure is not 
affected by the design change.
    Adding a redundant automatic containment isolation valve in 
penetration 53's non-essential instrument line instead of the excess 
flow check valve and isolating the non-essential line in penetration 
65's will significantly reduce the potential bypass leakage. The 
proposed change will credit the essential instrument lines as a 
closed system outside containment. The appropriate testing and 
acceptance criteria will be applied to ensure that any leakage 
associated with these potential bypass leakage paths, will not 
exceed the limits used in the Waterford 3 safety analysis or result 
in a significant increase in analyzed dose consequences. Therefore, 
the proposed change will not involve significant increase in the 
probability or consequences of any accident previously evaluated.
    The proposed change will credit the essential sensing lines 
outside containment as a closed system and will not affect the plant 
or the manner in which the plant [is] operated.
    The failure modes associated with containment isolation remain 
unchanged as a result of the design change to the non-essential 
monitoring lines. The function of the non-safety instrumentation is 
not affected. The only difference is that all of the non-safety 
instrumentation will be sensing containment pressure from 
penetration 53. However, if the non-essential line coming from 
penetration 53 becomes inoperable, containment pressure can be 
obtained from alternate instrumentation. Adding a redundant 
automatic containment isolation valve in series with CVR 401A in the 
non-essential instrument line ensures containment isolation 
following a LOCA with a concurrent a single active failure. 
Therefore, the proposed change will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The addition of a redundant automatic containment isolation 
valve in series with CVR 401A in the non-essential instrument line 
breaching penetration 53 ensures containment isolation postulating a 
single active failure on a Containment Isolation Actuation Signal 
(CIAS). While this modification is performed, administrative 
controls will require containment integrity to be maintained by a 
seismic Category 1, ASME Section III, Class 2, passive containment 
isolation device.
    The essential CVR instrument sensing lines form a seismically 
qualified, closed system outside containment which is designed for 
pressure equal to or greater than containment. The instrument 
cabinets C-3A(B) are seismic Category I and safety related. The 
instruments are Safety Class 1E and have a static pressure rating of 
1000 psig. These lines meet the criteria of BTP CSB 6-3 for 
crediting a closed system as a leakage boundary to preclude bypass 
leakage by being designed, fabricated, erected, and tested to 
standards commensurate with the safety function to be performed. The 
proposed change will apply the appropriate testing and acceptance 
criteria to ensure that any leakage associated with these potential 
bypass leakage paths, will not exceed the limits used in the 
Waterford 3 safety analysis or result in a significant increase in 
analyzed dose consequences. 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: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: October 4, 1996 (TSCR No. 250)
    Description of amendment request: The proposed Technical 
Specification (TS) change reflects a change in the Safety Limit Minimum 
Critical Power Ratio (SLMCPR) and as a result, a change in the 
operating Minimum Critical Power Ratio limit.
    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.
    The derivation of the revised SLMCPR for Oyster Creek for 
incorporation into the TS,

[[Page 57485]]

and its use to determine cycle-specific thermal limits, have been 
performed using NRC-approved methods. Additionally, interim 
implementing procedures, which incorporate cycle-specific 
parameters, have been used. Based on the use of these calculations, 
the revised SLMCPR will not increase the probability or consequences 
of an accident.
    The basis of the MCPR Safety Limit calculation is to ensure that 
greater than 99.9% of all fuel rods in the core avoid transition 
boiling if the limit is not violated. The new SLMCPR preserves the 
existing margin to transition boiling and fuel damage in the event 
of a postulated accident. The probability of fuel damage is not 
increased.
    Revising the operating MCPR limit for stability will ensure that 
adequate margin is retained to the SLMCPR.
    Therefore, the proposed TS change does 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 MCPR Safety Limit is a Technical Specification numerical 
value designed to ensure that fuel damage from transition boiling 
does not occur as a result of the limiting postulated accident. The 
stability MCPR limit ensures an adequate operating MCPR margin to 
the SLMCPR. These revised limits cannot create the possibility of 
any new type of accident. The new SLMCPR has been calculated using 
NRC-approved methods. Additionally, interim procedures, which 
incorporate cycle-specific parameters, have been used. 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 changes do not involve a significant 
reduction in a margin of safety.
    The margin of safety as defined in the TS Bases will remain the 
same. The new SLMCPR is calculated using NRC-approved methods which 
are in accordance with the current fuel design and licensing 
criteria. Additionally, interim implementing procedures, which 
incorporate cycle-specific parameters, have been used. The MCPR 
Safety Limit remains high enough to ensure that greater than 99.9% 
of all fuel rods in the core will avoid transition boiling if the 
limit is not violated, thereby preserving fuel cladding integrity. 
The revised stability MCPR limit retains the existing margin to the 
SLMCPR. Therefore, the proposed TS change does not involve a 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: October 10, 1996 (TSCR No. 203)
    Description of amendment request: The proposed Technical 
Specification revision will extend the instrumentation surveillances 
for Condenser Low Vacuum, High Temperature Main Steamline Tunnel, 
Recirculation Flow, and Reactor Coolant Leakage. Additionally, the 
change will extend the equipment tests/operability checks for 
Containment Vent and Purge Isolation, Electromagnetic Relief Valve 
Operability, and Drywell to Torus Leakage Test. The above change 
extensions conform with the 24 month refueling interval.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated.
    The proposed amendment extends the period between successive 
refueling interval surveillance(s) to once every 24 months for those 
surveillance(s) evaluated herein. The proposed surveillance interval 
changes do not involve any change to the actual surveillance 
requirements, nor does it involve any change to the limits and 
restrictions on plant operations. The reliability of systems and 
components relied upon to prevent or mitigate the consequences of 
accidents previously evaluated is not degraded by the proposed 
change to the surveillance interval. Assurance of system and 
equipment availability is maintained. This change does not involve 
any change to system or equipment configuration. Therefore, this 
change does not increase the probability of occurrence or the 
consequences of an accident previously evaluated.
    Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendment extends the period between successive 
refueling interval surveillance(s) to once every 24 months for those 
surveillance(s) evaluated herein. The proposed surveillance interval 
changes do not involve any change to the actual surveillance 
requirements, nor does it involve any change to the limits and 
restrictions on plant operation. This change does not involve any 
change to system or equipment configuration. Therefore, this change 
is unrelated to the possibility of creating a new or different kind 
of accident from any previously evaluated.
    Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed amendment extends the period between successive 
refueling interval surveillance(s) to once every 24 months (+/-25% 
or 30 months) for the surveillances evaluated herein. The proposed 
surveillance interval changes do not involve any change to the 
actual surveillance requirements, nor does it involve any change to 
the limits and restrictions on plant operation. The reliability of 
systems and components is not degraded by the proposed change to the 
surveillance interval. Assurance of system and equipment 
availability is maintained. Therefore, it is concluded that 
operation of the facility in accordance with the proposed amendment 
does not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of application for amendment: October 10, 1996 (TSCR No. 243)
    Description of amendment: The proposed Technical Specification (TS) 
will change the trip setting for TS Table 3.1.1 Item G.3, Automatic 
Depressurization System (ADS) by clarification of the functional 
requirement to provide an interlock permissive which ensures that a 
source of cooling water is available via the Core Spray System prior to 
depressurization. This will be accomplished by replacing the present 
interlock description ``AC Voltage'' with core spray booster pump 
differential pressure, as the permissive required for initiation of 
ADS. A corresponding surveillance requirement is being added to TS 
Table 4.1.1 which reflects the need to test and calibrate the core 
spray booster pump differential pressure switches pursuant to existing

[[Page 57486]]

plant procedures. Additionally, allowed outage time (AOT) is addressed 
in the footnote ``i'' for the differential pressure switches based upon 
the currently designed ADS logic trains and footnote ``h'' to parallel 
the ``Low-Low Reactor Water Level'' and ``High Drywell Pressure'' AOTs 
associated with Standard Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated.
    The implementation of this TSCR does not involve an increase in 
the probability of occurrence or the consequences of an accident 
previously evaluated, as no plant modifications are proposed by the 
change request, and no changes in instrument set or reset setpoints 
are required in order to implement the change. This change serves to 
clarify and to incorporate the ``as-built'' ADS system logic 
parameter (core spray booster pump differential pressure) as the 
functional permissive required for initiation of ADS. This 
``interlock'' permissive compares closely with that of the BWR 
[boiling-water reactor] STS [Standard Technical Specifications] 
requirement to monitor core spray discharge pressure for initiation 
of ADS. In addition, the AOTs for the ADS initiation signals are 
being revised to align with the AOTs provided for such signals in 
the STS. The performance and function of the Automatic 
Depressurization System is unchanged by this request. However, by 
implementation of the change the specific functions of the ADS as-
built d/p permissives would then be clearly identified in and 
controlled by T.S. Table 3.1.1, ``Protective Instrumentation 
Requirements,'' including the associated surveillance requirements 
as shown on the revised T.S. Table 4.1.1.
    Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The implementation of this TSCR does not impact upon the 
operation of the facility, and would not create the possibility of a 
new or different kind of accident from any previously evaluated 
because no plant modifications are proposed by this change request, 
and no changes in instrument set or reset setpoints are required in 
order to implement the change. This change clarifies the technical 
specifications by incorporating the ``as-built'' ADS system logic 
parameter (core spray booster pump differential pressure) as the 
functional permissive required for initiation of ADS. This 
``interlock'' permissive compares closely with that of the BWR STS 
requirement to monitor core spay discharge pressure. The revised 
AOTs for ADS initiation signals are also being changed to conform 
with those allowed by and provided in the STS. The performance and 
function of the Automatic Depressurization System (ADS) is unchanged 
by this request.
    OC plant surveillance procedures for both ADS and the Core Spray 
system presently incorporate the calibration requirements and both 
the set and reset setpoints calculated for the core spray booster 
pump d/p switch permissive to the ADS initiation logic. Hence, a new 
or different kind of accident from any previously evaluated is not 
created.
    Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The implementation of this TSCR does not involve a reduction in 
the margin of safety for operation of the ADS or the Core Spray 
system. The Technical Specification Bases which presently define the 
margin of safety are not impacted as the core spray booster pump d/p 
``interlock'' permissives are not described in the specifications 
for ``Protective Instrumentation Requirements'' or its surveillance 
requirements. In addition, the margin of safety for ADS initiation 
is not reduced by this TSCR because the required system response is 
not affected by the proposed changes as no plant modifications are 
required which could create a potential impact upon the margins of 
safety previously established.
    The revision of AOTs associated with ADS actuation signals by 
extension form 72 hours to 4 days is consistent with that presently 
provided in the STS. This does not decrease the margin of safety 
associated with availability of ADS as placement of the initiation 
signals into the ``tripped condition'' maintains the operability of 
the ADS trip systems while in the automatic mode. Additionally, the 
Bases for STS Specification 3.1 provides justifications for AOTs 
using the GE [General Electric] reliability analyses referenced 
therein and therefore 4 days is both justified and conservative. The 
margin of safety with respect to the instrument channels ability to 
perform its intended actuation function is not impacted; therefore, 
there is no reduction in the margin of safety.
    Lastly, the surveillance frequency for the new surveillance 
interval created on Table 4.1.1 for the d/p [s]witches is consistent 
with that established in Reference 2 of the Bases for Technical 
Specification 4.1. Therefore, there is no reduction in the margin of 
safety as a result of this change request.
    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: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

Illinois Power Company and Soyland Power Cooperative, Inc., Docket No. 
50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois

    Date of amendment request: October 17, 1996
    Description of amendment request: The proposed amendment would 
revise Facility Operating License NPF-62 to acknowledge the transfer of 
Soyland Power Cooperative's 13.21% minority ownership interest in the 
Clinton Power Station to Illinova Power Marketing, Inc., the 
unregulated power marketing affiliate of Illinois Power, and a wholly 
owned subsidiary of Illinova Corporation.
    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 because it merely revises the Operating License 
to indicate the transfer of a minority ownership interest to the 
corporate parent of the majority owner and licensee. This proposed 
amendment represents an administrative rather than operational 
change and, therefore, has no impact on accidents previously 
evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated because Illinois Power will continue to be the operator of 
Clinton
    Power Station, and further, there will be no change to the 
plant's physical configuration or operating philosophy as a result 
of this proposed amendment.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety because it is only an 
administrative change and will have no impact on any margin of 
safety related to the design or operation of the facility.
    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: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727
    Attorney for licensee: Leah Manning Stetzner, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th Street, 
Decatur, Illinois 62525
    NRC Project Director: Gail H. Marcus

[[Page 57487]]

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

    Date of amendment request: September 13, 1996
    Description of amendment request: The proposed amendment would 
revise the Maine Yankee containment testing technical specification (TS 
4.4) to implement 10 CFR Part 50, Appendix J, Option B, by referring to 
Regulatory Guide 1.163, ``Performance-Based Containment Leakage-Test 
Program'' dated September 1995.
    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 amendment request does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated, because the proposed changes to the Technical 
Specification do not affect the assumption, parameters or results of 
any FSAR accident analysis.[...] 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 probability of 
an accident previously identified since the containment system is 
used for mitigation purposes only. The changes are also 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. The addition of the 
''...[as modified by approved] exemptions'' phrase is an 
administrative change. Any specific exemptions from the requirements 
of Appendix J will continue to require a submittal under 10 CFR 
50.12 and subsequent review and approval by the NRC prior to 
implementation. Therefore, these changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Operation of Maine Yankee 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) nor alter the 
function of the containment system. The changes only provide for 
additional time between leakage tests and an increase in the test 
pressure value equal to the containment design pressure which bounds 
the containment peak accident pressure. Thus, these changes do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Operation of Maine Yankee in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. The proposed changes do not alter the manner in which safety 
limits, limiting safety system setpoints, or limiting conditions for 
operation are determined. The changes are expected to result in an 
increased focus on components demonstrating poor leakage test 
history without excessive testing of components which continue to 
demonstrate good test history. Therefore, these 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: Wiscasset Public Library, High Street, 
P.O. Box 367, Wiscasset, ME 04578
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, 329 Bath Road, Brunswick, ME 04011NRC Deputy Director: 
John A. Zwolinski

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

    Date of amendment request: September 13, 1996, as supplemented 
September 25, 1996
    Description of amendment request: The proposed amendment would 
revise TS 5.5.B to eliminate references to the Vice President (YNSD) 
and designate the President, Maine Yankee, as the responsible official 
for matters related to the composition, review and audit 
responsibilities, authority and recordkeeping responsibilities of the 
Nuclear Safety Audit and Review (NSAR) Committee. Minor editorial 
changes are also proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below.
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change is administrative in nature and will not 
have a direct effect on the physical plant or the maintenance of the 
physical plant. The audit and review functions of the NSAR Committee 
will continue to be required. The proposed changes will not, of 
themselves, decrease the effectiveness of these functions. This 
authority and responsibility realignment will continue to assure 
that NSAR Committee has direct access to a level of management 
necessary to perform their audit and review functions.
    Since, the proposed change will not adversely effect the audit 
and review functions of the NSARC and since the proposed change will 
not have a direct effect on the physical plant or maintenance of the 
physical plant, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change is administrative in nature and does not 
introduce any new structures, systems, or components into the plant 
design. This change continues to ensure that the NSAR Committee 
reports to a management level such that there is sufficient 
authority and organizational freedom to execute their audit and 
review functions. Consequently, an unbiased oversight of the 
programs and procedures is not compromised by this proposed change. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed change realigns the authority and responsibility 
relationship of the NSAR Committee. The NSAR Committee will continue 
to maintain effective oversight of programs and procedures. The 
proposed change will continue to ensure that the NSAR Committee is 
sufficiently independent from cost and schedule when opposed to 
safety considerations. Therefore, the proposed change does not 
involve a significant reduction in the margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, 329 Bath Road, Brunswick, ME 04011NRC Deputy Director: 
John A. Zwolinski

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

    Date of amendment request: July 2, 1996
    Description of amendment request: The proposed amendment 
incorporates limiting conditions for operation and surveillance 
requirements for the safety/relief valve (SRV) electrical lift design 
modification. The proposed amendment also makes clarification and 
editorial changes, as well as revising the associated Bases section.

[[Page 57488]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Pursuant to 10CFR50.92, NNECO has reviewed the proposed change 
and concludes that the change does not involve a significant hazards 
consideration (SHC) since the proposed change satisfies the criteria 
in 10 CFR 50.92(c). That is, the proposed change does not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The safety relief valves are considered for two analyzed 
accidents, an overpressure transient (such as MSIV [main steam 
isolation valve] closure with flux scram) and an inadvertent SRV 
opening.
    The new technical specifications do not affect normal operation, 
therefore, they cannot increase the probability of an overpressure 
event. Since the mechanical function will not be affected by the new 
equipment, the new LCOs [limiting conditions for operation], or the 
new surveillance requirements, there is no adverse affect on the 
consequences of an overpressure event. The SRVs will be expected to 
lift mechanically. If they do not open at the design setpoints, the 
electrical actuation, which has the same setpoints, will cause the 
valves to open less than 400 milliseconds later.
    Sufficient redundancy and diversity is established for the 
electrical lift by the use of two sensors in a two-out-of-two-taken-
once configuration. Therefore, the failure of any single component 
cannot result in an inadvertent opening of an SRV. The only proposed 
surveillance performed while at power is the daily instrument check. 
This surveillance does not require the manipulation of any controls 
and, as such, cannot affect the probability of an accident.
    Therefore, based on the above, the proposed change to the 
Technical Specifications does not involve a significant increase in 
the probability or consequences of any previously evaluated 
accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    None of the proposed new LCOs or surveillance requirements has a 
potential for creating a new or different kind of accident. 
Expanding the LCO and surveillance requirements to address both the 
mechanical actuation and the pressure sensor lift does not change 
the type of action that these valves are expected to perform, nor 
does it change the initial ``as-left'' requirements for the valves. 
Plant operating parameters have also not changed.
    Therefore, this change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The margin of safety previously analyzed for the SRVs was based 
on the current nominal setpoints and allowable percent drift. The 
electrical lift system improves the confidence that the SRVs will 
lift within the specified range. The setpoint uncertainty of the 
electrical lift system is similar to the drift allowed for the 
mechanical lift in the Technical Specifications. All existing 
functions that may actuate the SRVs (safety, manual, or automatic 
lift) remain unaffected. The design of the pressure transmitters, 
combined with the logic configuration, minimizes the possibility of 
inadvertently opening the SRVs.
    Therefore, this change has no impact on the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: March 29, 1996
    Description of amendment request: The proposed Technical 
Specifications (TS) changes would revise TS Surveillance Requirement 
(SR) 4.5.1.d.2.b to delete the requirement to perform in-situ 
functional testing of the Automatic Depressurization System (ADS) 
valves once every 24-months as part of start-up testing activities.
    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 proposed TS change does not involve any physical changes to 
plant structures, systems, or components (SSC). The ADS will 
continue to function as designed. The ADS is an Emergency Core 
Cooling System (ECCS) designed to mitigate the consequences of an 
accident, and therefore, can not contribute to the initiation of any 
accident. The ADS utilizes five (5) of the 14 main steam line SRVs 
as the primary method for depressurizing the reactor pressure vessel 
to permit low pressure core cooling capability in the event of a 
small break Loss-of-Coolant-Accident (LOCA) if the high pressure 
cooling systems (i.e., High Pressure Coolant Injection (HPCI) and 
Reactor Core Isolation Cooling (RCIC) systems) fail to maintain 
adequate reactor vessel water level.
    Deleting the TS SR to perform the in-situ testing of the ADS/
SRVs during start-up, as proposed, should reduce the probability of 
an inadvertent opening of an SRV as discussed in Section 15.1.4 of 
the LGS Updated Final Safety Analysis Report (UFSAR) since deleting 
this testing requirement will eliminate a known initiator of SRV 
pilot leakage and subsequent erosion. This proposed TS change will 
have a tendency to increase, rather than decrease, the reliability 
of the ADS/SRVs by eliminating the in-situ ADS functional start-up 
testing. The probability of the ADS/SRVs to open on demand has been 
demonstrated to be extremely high and is not measurably improved 
through the in-situ ADS functional start-up testing.
    This proposed TS change will not increase the probability of 
occurrence of a malfunction of any plant equipment important to 
safety. Alternate testing methods at LGS, Units 1 and 2, and at the 
off-site test facility, adequately demonstrate proper ADS valve 
operation and assure that the valves will continue to function as 
designed. Existing surveillance testing and inspections of the ADS/
SRVs at LGS verify that the ADS initiation logic, solenoid valve 
operation, pneumatic gas supply integrity and air operator assembly 
(including pilot rod) will operate as designed. Offsite testing 
verifies pilot disc operation, setpoint calibration and main valve 
disc operation.
    Deleting the in-situ testing requirement, as proposed, will 
reduce the probability of inflating SRV leakage which should reduce 
the probability of an inadvertent SRV opening. It has been 
documented throughout the BWR industry that pilot disc leakage leads 
to pilot disc and rod erosion, which can ultimately result in an 
inadvertent opening of an SRV. Therefore, any SRV pilot leakage that 
can be eliminated would reduce the probability of occurrence of a 
malfunction of that SRV.
    Deleting the ADS/SRV in-situ functional test will in no way 
increase any consequences of a malfunction of plant equipment 
important to safety. The consequences of a malfunction of an ADS/SRV 
as discussed in the LGS UFSAR remain unchanged.
    In addition, eliminating a known initiator of SRV leakage, as 
proposed in this TS change, would help to reduce operator 
workarounds in the form of suppression pool cooling and letdown 
operation activities. As a result, this will reduce the unnecessary 
operation of the Residual Heat Removal (RHR) and Residual Heat 
Removal Service Water (RHRSW) systems.
    Therefore, the proposed TS change does not involve an increase 
in the probability or consequences of an accident previously 
evaluated.

[[Page 57489]]

    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    This proposed TS change does not involve any physical changes to 
plant SSC. The design and operation of the ADS/SRVs is not changed 
from that currently described in the Safety Analysis Report (SAR). 
The ADS will continue to function as designed to mitigate the 
consequences of an accident. No changes of any kind are being made 
to the valves, auxiliary components, or ADS logic. Deleting the 
requirement to perform the ADS in-situ functional test during plant 
start-up as proposed in this TS Change Request reduces the 
likelihood of a SRV developing a leak and degrading throughout the 
subsequent operating cycle. There is no possibility that 
implementing this proposed TS change would create a different type 
of malfunction to the ADS/SRVs than any previously evaluated.
    Eliminating the requirement to perform the in-situ testing of 
the ADS/SRVs during start-up activities, does not create a new or 
different type of accident than any previously evaluated. There is 
no accident scenario associated with testing the ADS/SRVs other than 
the inadvertent opening of a relief valve which is currently 
discussed in Section 15.1.4 of the LGS UFSAR. This proposed TS 
change does not alter the conclusions described in the UFSAR 
regarding an inadvertent opening of an SRV. No new or different type 
of accident will be created as a result of this proposed TS change.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The proposed TS change does not involve any physical changes to 
plant SSC. The design and functional requirements of the ADS will 
not change. The ADS will still function as designed to mitigate the 
consequences of an accident.
    This proposed TS change involves deleting the requirement to 
perform in-situ functional testing of the ADS/SRVs during start-up 
activities. This testing imposes an unnecessary challenge on the 
ADS/SRVs and has been linked to SRV degradation (e.g., pilot valve 
and/or main valve leakage). This proposed TS change should reduce 
SRV leakage and improve ADS/SRV reliability by reducing the 
potential for spurious SRV actuation. The LGS TS Bases do not 
identify specific testing requirements for ADS. ADS operability can 
be readily demonstrated with extremely high confidence by the 
existing additional surveillance tests and inspections performed for 
the ADS. There will be no reduction in any margin of safety 
resulting from this proposed TS change.
    Therefore, the proposed TS 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: 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: August 8, 1996
    Description of amendment request: The proposed Technical 
Specifications (TS) changes would revise TS Sections 3/4.3.1, ``Reactor 
Protection System Instrumentation,'' 3/4.3.2, ``Isolation Actuation 
Instrumentation,'' 3/4.3.3, ``Emergency Core Cooling System Actuation 
Instrumentation,'' and the associated TS Bases Sections 3/4.3.1 and 3/
4.3.2 to eliminate selected response time testing requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
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 make any physical alterations or 
modifications to the plant systems or equipment. The proposed 
changes do not affect the capability of the associated systems to 
perform their intended functions within their required response 
times, nor do the proposed changes adversely impact the operation of 
any plant equipment. The affected plant systems will continue to 
function as designed. Elimination of the response time testing 
requirements as proposed by this TS change for selected components 
in RPS Instrumentation, Isolation Actuation System Instrumentation, 
and ECCS Actuation Instrumentation will not adversely affect the 
operation of these components.
    The supporting analysis provided in NEDO-32291, demonstrates 
that response time testing is redundant to other TS required 
testing. NEDO-32291 demonstrated that these other required tests 
(i.e., channel checks, channel calibrations, channel functional 
tests, and logic system functional tests), in conjunction with 
actions taken in response to NRC Bulletin 90-01 and NRCB 90-01, 
Supplement 1, are sufficient to identify failure modes or 
degradation in instrument response times, and ensure operation of 
the associated systems within acceptable limits. There are no known 
failure modes that can be detected by response time testing that 
cannot also be detected by other TS required testing. The continued 
application of other existing TS required testing such as channel 
checks, channel calibrations, channel functional tests, and logic 
system functional tests, ensures that the response times for these 
systems will be maintained within the acceptance limits. The 
capability of these systems to perform their intended functions 
within their required response times is not adversely impacted by 
this proposed TS change. NEDO-32291 evaluated the potential failure 
modes of the affected instrumentation loops which could impact the 
instrument loop response times. Industry operating experience was 
also reviewed to identify failures that affect response times and 
how they are detected. The failure modes identified were evaluated 
to determine if other TS required surveillances and actions taken in 
response to NRC Bulletin 90-01, and NRCB 90-01, Supplement 1, would 
detect any effects on response time. There are no failures [sic] 
[failure] modes identified that can be detected by response time 
testing that cannot also be detected by other TS required testing.
    PECO Energy has confirmed the applicability of the generic 
evaluation provided in NEDO-32291 to LGS, Units 1 and 2. By letter 
dated December 28, 1994, the NRC concluded that response time 
testing can be eliminated from the TS for the selected 
instrumentation identified in NEDO-32291, with certain provisions, 
and that NEDO-32291 can be referenced in license amendment requests.
    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 do not involve any physical changes to 
plant systems or equipment. The proposed changes apply only to the 
testing requirements for the selected components involved and do not 
result in any physical modifications to these components, or to 
other plant system components. Elimination of the response time 
testing requirements as proposed by this TS change for selected 
components in RPS Instrumentation, Isolation Actuation System 
Instrumentation, and ECCS Actuation Instrumentation will not 
adversely affect the operation of these components. These components 
will continue to function as designed. Consequently, no new failure 
modes are introduced as a result of the proposed TS changes.
    Eliminating the response time testing requirements as proposed, 
does not create a new or different type of accident than any 
previously evaluated. No new or different type of accident will be 
created as a result of this proposed TS change.
    NEDO-32291 demonstrates that other required tests (i.e., channel 
checks, channel calibrations, channel functional tests, and logic 
system functional tests), in conjunction with actions taken in 
response to NRC Bulletin 90-01 and NRCB 90-01, Supplement

[[Page 57490]]

1, are sufficient to identify failure modes or degradation in 
instrument response times, and ensure operation of the associated 
systems within acceptable limits. There are no known failure modes 
that can be detected by response time testing that cannot also be 
detected by other TS required testing, and therefore, response time 
testing for the selected components is redundant to the other TS 
required testing.
    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 proposed TS changes do not involve any physical changes to 
plant systems or equipment. The proposed TS changes do not affect 
the capability of the associated systems or equipment from 
performing their intended functions. The systems involved will 
continue to respond within their allowed response times. Elimination 
of the response time testing requirements are based on the 
evaluation provided in NEDO-32291 which demonstrates that response 
time degradation can be detected by other TS required testing. The 
evaluation concluded that other TS required tests (i.e., channel 
checks, channel calibrations, channel functional tests, and logic 
system functional tests), in conjunction with actions taken in 
response to NRC Bulletin 90-01 and NRCB 90-01, Supplement 1, are 
sufficient to identify failure modes or degradation in instrument 
response times, and ensure operation of the associated systems 
within acceptable limits.
    In addition, although not specifically evaluated, the proposed 
TS changes will provide an improvement to plant safety and operation 
by reducing the time safety systems are unavailable, reducing the 
potential for safety system actuations, reducing plant operating and 
shutdown risk, limiting radiation exposure to plant personnel, and 
eliminating the diversion of key personnel to conduct unnecessary 
testing. Therefore, PECO Energy considers that the proposed TS 
changes will result in an overall increase in the margin of safety 
and that the changes do not constitute an unreviewed safety 
question.
    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 No. 50-353, Limerick 
Generating Station, Unit 2, Montgomery County, Pennsylvania

    Date of amendment request: August 1, 1996
    Description of amendment request: The proposed Technical 
Specifications (TS) changes would revise TS Section 3/4.4.6 (i.e., 
Figure 3.4.6.1-1) to reflect the addition of two hydrotest curves, 
effective for 6.5 and 8.5 Effective Full Power Years (EFPY), to the 
existing Pressure-Temperature Operating Limit (PTOL) curves for LGS 
Unit 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed Technical Specifications (TS) change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed Technical Specification (TS) change includes 
Pressure-Temperature Operating Limit (PTOL) curves which were 
conservatively generated in accordance with the fracture toughness 
requirements of 10CFR50, Appendix G. The Adjusted Reference 
Temperatures to the initial nil ductility reference temperatures 
(RTNDT) used to evaluate the pressure/temperature limits for the 
beltline materials were based on Regulatory Guide 1.99, Revision 2. 
Future analyses of the Reactor Pressure Vessel (RPV) surveillance 
capsule contents and future revisions to the PTOL curve as required, 
ensure that the reactor pressure boundary will behave in a non-
brittle manner during plant testing, startup, and operation 
throughout the life of the plant. The current schedule for removal 
of the surveillance specimens from Limerick Generating Station (LGS) 
Unit 2 RPV is during 2R05. The proposed change does not impact the 
existing PTOL curves for 10 Effective Full Power Years (EFPY), 
currently shown in the LGS Unit 2 TS. The proposed change only 
provides additional information (i.e., two new curves) related to 
the RPV condition following 6.5 and 8.5 EFPY, in order to facilitate 
hydrostatic testing performed after 2R04 and 2R05, respectively. The 
added PTOL curves are established in compliance with the methodology 
used to calculate the predicted irradiation effects on vessel 
beltline materials as documented in the LGS Updated Final Safety 
Analysis Report (UFSAR). There are no physical changes to the plant 
being introduced by the added PTOL curves.
    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 proposed Technical Specification (TS) change includes 
Pressure-Temperature Operating Limit (PTOL) curves which were 
conservatively generated in accordance with the fracture toughness 
requirements of 10CFR50, Appendix G. The Adjusted Reference 
Temperatures to the initial nil ductility reference temperatures 
(RTNDT) used to evaluate the pressure/temperature limits for the 
beltline materials were based on Regulatory Guide 1.99, Revision 2. 
The proposed changes do not impact the existing PTOL curves for 10 
Effective Full Power Years (EFPY), currently shown in the TS. They 
only provide additional information (i.e., two new curves) related 
to the reactor pressure vessel condition for 6.5 and 8.5 EFPY, in 
order to facilitate hydrostatic testing performed after 2R04 and 
2R05, respectively. The added PTOL curves are established in 
compliance with the previous methodology used to calculate the 
predicted irradiation effects on vessel beltline materials as 
documented in the LGS [Updated Final Safety Analysis Report] UFSAR. 
The proposed TS change does not involve any physical changes to 
safety-related equipment.
    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 proposed change to Technical Specifications (TS) does not 
reduce the margin of safety as defined in the Bases for any TS. The 
added Pressure-Temperature Operating Limit (PTOL) curves for 6.5 and 
8.5 Effective Full Power Years (EFPY) corresponding to 2R04 and 
2R05, respectively, have been calculated in accordance with the 
existing methodology used to calculate the PTOL curves currently 
existing in the LGS Unit 2 TS (i.e., complying with the requirements 
of 10CFR50 Appendix G, and Regulatory Guide 1.99, Revision 2) and 
will more closely reflect the actual required reactor pressure 
vessel condition at the time in which the hydrotest is performed. 
Therefore, the margin of safety is not affected.
    Therefore, the proposed TS change does not involve a reduction 
in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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

[[Page 57491]]

Philadelphia Electric Company, Docket No. 50-353, Limerick 
Generating Station, Unit 2, Montgomery County, Pennsylvania

    Date of amendment request: August 5, 1996
    Description of amendment request: The proposed Technical 
Specifications (TS) changes would revise TS Section 2.1 and its 
associated TS Basis to reflect the change in the Minimum Critical Power 
Ratio (MCPR) Safety Limit due to the plant specific evaluation 
performed by General Electric Co. (GE), for LGS Unit 2 Cycle 4.
    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 revised Minimum Critical Power Ratio (MCPR) Safety Limit for 
LGS Unit 2 Technical Specifications, and its use to determine cycle-
specific thermal limits have been performed using NRC-approved 
methods within the existing design and licensing basis, and cannot 
increase the probability or severity of an accident.
    The basis of the MCPR Safety Limit calculation is to ensure that 
greater than 99.9% of all fuel rods in the core avoid transition 
boiling if the limit is not violated. The new MCPR Safety Limit 
preserves the existing margin to transition boiling and fuel damage 
in the event of a postulated accident. The probability of fuel 
damage is not 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 MCPR Safety Limit is a Technical Specification numerical 
value, designed to ensure that fuel damage from transition boiling 
does not occur as a result of the limiting postulated accident. It 
cannot create the possibility of any new type of accident. The new 
Minimum Critical Power Ratio (MCPR) Safety Limit is calculated using 
NRC-approved methods and is based on LGS Unit 2 Cycle 4 specific 
inputs.
    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 margin of safety as defined in the TS Bases will remain the 
same. The new Minimum Critical Power Ratio (MCPR) Safety Limit is 
calculated using NRC approved methods which are in accordance with 
the current fuel design and licensing criteria. The MCPR Safety 
Limit remains high enough to ensure that greater than 99.9% of all 
fuel rods in the core will avoid transition boiling if the limit is 
not violated, thereby preserving the fuel cladding integrity.
    Therefore, the proposed TS change does not involve a reduction 
in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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

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

    Date of amendments request: September 30, 1996
    Description of amendments request: The proposed amendments would 
revise Technical Specifications (TSs) 3/4.1.1, 3/4.1.3, 3.1.3.6, 3.2.1, 
3/4.2.2, and 3.2.3 and associated Bases to remove certain cycle-
specific parameter limits from the TSs and relocate them to the Core 
Operating Limits Report (COLR). These changes result from NRC Generic 
Letter (GL) 88-16, dated October 4, 1988, which provided guidance to 
licensees on requests for removal of the values of cycle-specific 
parameter limits from the TSs. The licensee's proposed amendments are 
consistent with the GL.
    The COLR has been included in the Definitions section of the TSs. 
The definition notes that it is the unit-specific document that 
provides these limits for the current operating reload cycle. The 
values of these cycle-specific parameter limits are to be determined in 
accordance with TS 6.9.1.11. This TS requires that the core operating 
limits be determined for each reload cycle in accordance with the 
referenced NRC-approved methodology for these limits and consistent 
with the applicable limits of the safety analysis. The COLR shall be 
provided to the NRC upon issuance.In addition, the above TS changes 
would produce administrative changes to the TS Table of Contents.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The removal of cycle-specific core operating limits from the FNP 
[Farley Nuclear Plant] Technical Specifications has no influence or 
impact on the probability or consequences of a Design Basis Accident 
(DBA) occurrence. The cycle-specific core operating limits, although 
not in Technical Specifications, will be followed in the operation 
of FNP. The proposed amendment retains the same required actions to 
be taken when or if limits are exceeded as stipulated by current 
Technical Specifications. In addition, the associated surveillance 
requirements are not altered by the proposed changes.
    Each accident analysis addressed in the FNP FSAR [Final Safety 
Analysis Report] will be examined with respect to changes in cycle-
dependent parameters, which are obtained from application of the 
NRC-approved reload design methodologies, to ensure that the 
transient evaluation of new reloads are bounded by previously 
accepted analyses. This examination, which will be performed per 
requirements of 10 CFR 50.59, ensures that future reloads will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    As stated earlier, the removal of the cycle-specific variables 
has no influence or impact, nor does it contribute in any way to the 
probability or consequences of an accident. No safety-related 
equipment, safety function, or plant operation will be altered as a 
result of this proposed change. The cycle-specific variables are 
calculated using the NRC-approved methods and submitted to the NRC 
to allow the Staff to continue to trend the values of these limits. 
The Technical Specifications will continue to require operation 
within the required core operating limits and appropriate actions 
will be taken when or if limits are exceeded. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed changes do not result in a significant reduction 
in the margin of safety.
    The margin of safety is not affected by the removal of cycle-
specific core operating limits from the Technical Specifications. 
The margin of safety presently provided by current Technical 
Specifications remains unchanged. Appropriate measures exist to 
control the values of these cycle-specific limits. The proposed 
amendment continues to require operation within the core limits, as 
obtained from the NRC-approved reload design methodologies. The 
required actions to be taken or if limits are violated remain 
unchanged.
    The development of the limits for future reloads will continue 
to conform to those

[[Page 57492]]

methods described in NRC-approved documentation. In addition, each 
future reload involves a 10 CFR 50.59 safety review to assure that 
operation of FNP within the cycle-specific limits will not involve a 
significant reduction in [the] margin of safety. Therefore, the 
proposed changes are administrative in nature and do not impact the 
operation of FNP in a manner that involves a reduction to the margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: Herbert N. Berkow

Previously Published Notices Of Consideration Of Issuance Of 
Amendments To Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, And Opportunity For A Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

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

    Date of amendment request: August 29, 1996
    Description of amendment request: The proposed amendment would 
modify the applicability requirements for certain radiation monitors so 
that the radiation monitors are required to be operable only when 
secondary containment integrity is required to be operable; delineate 
when secondary containment integrity is required; modify standby gas 
treatment operability requirements; make editorial corrections to 
clarify the configuration of the radiation monitors; and revise the 
associated Bases section.
    Date of publication of individual notice in Federal Register: 
October 17, 1996 (61 FR 54242)
    Expiration date of individual notice: November 18, 1996
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut

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

Date of application for amendment: September 6, 1996

    Brief description of amendment request: The proposed amendment 
would change Technical Specification (TS) requirements related to steam 
generator tubes to allow a laser-welded repair of Westinghouse hybrid 
expansion joint (HEJ) sleeved steam generator tubes. Date of individual 
notice in Federal Register: October 15, 1996 (61 FR 53769)

Expiration date of individual notice: November 14, 1996

    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin

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.

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

    Date of application for amendments: June 17, 1996
    Brief description of amendments: The amendment modifies the 
technical specifications (TS) to change (1) the reference method for 
calculating dose conversion factors (DCFs) to be used in dose 
calculations, and (2) the upper and lower limits for operating 
pressurizer pressure to account for new instrument uncertainties and to 
reduce the allowed operating band.
    Date of issuance: October 23, 1996
    Effective date: October 23, 1996, to be implemented within 45 days 
of issuance
    Amendment Nos.: Unit 1 - 109; Unit 2 - 101; Unit 3 - 81
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 11, 1996 (61 
FR 47963). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 23, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004

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

    Date of application for amendments: June 9, 1995

[[Page 57493]]

    Brief description of amendments: The amendments implement changes 
to radiological effluent Technical Specifications in accordance with 
Generic Letter 89-01 ``Implementation of Programmatic for Radiological 
Effluent Technical Specification in the Administrative Controls Section 
of the Technical Specifications and Relocation of Procedural Details of 
RETS to the Offsite Dose Calculation Manual or to the Process Control 
Program.''
    Date of issuance: October 18, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 217 and 194
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 15, 1995 (60 FR 
35062) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated October 18, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendment: May 1, 1996
    Brief description of amendment: The proposed amendment will reflect 
the implementation of 10 CFR Part 50 Appendix J, Option B at the 
Pilgrim Nuclear Power Station.
    Date of issuance: October 4, 1996
    Effective date: October 4, 1996
    Amendment No.: 167
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28606) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 4, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

Carolina Power & Light Company, et al., Docket No. 50-325, 
Brunswick Steam Electric Plant, Unit 1, Brunswick County, North 
Carolina

    Date of amendment request: April 8, 1996, as supplemented on July 
30, 1996, October 4, 1996, October 8, 1996, and October 16, 1996.
    Brief description of amendment: The amendment changes the Technical 
Specifications to (1) reflect the use of a new type of fuel (GE13) and 
(2) modify the minimum critical power ratio safety limit and the 
standby liquid control system sodium pentaborate limits to accommodate 
the GE13 fuel.
    Date of issuance: October 17, 1996
    Effective date: October 17, 1996
    Amendment No.: 182
    Facility Operating License No. DPR-71: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 14, 1996 (61 FR 
42276) which superseded a Federal Register notice published on June 5, 
1996 (61 FR 28607) The Commission's related evaluation of the amendment 
is contained in a Safety Evaluation dated October 17, 1996.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

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

    Date of application for amendment: January 5, 1996, as supplemented 
July 12, 1996
    Brief description of amendment: The amendment revises the shutdown 
cooling (SDC) requirement to allow one train of the SDC system to be 
rendered inoperable for testing or maintenance provided that a filled 
refueling cavity is available to provide backup decay heat removal 
capability in the event that the operating train of SDC becomes 
inoperable.
    Date of issuance: October 10, 1996
    Effective date: October 10, 1996
    Amendment No.: 173
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 28, 1996 (61 FR 
44348) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 10, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423

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

    Date of application for amendments: August 8, 1996
    Brief description of amendments: The amendments revise the 
Technical Specifications, Section 6.9.1.9, to reference updated or 
recently approved topical reports used to calculate cycle-specific 
limits contained in the Core Operating Limits Report.
    Date of issuance: October 24, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 154 and 146
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 11, 1996 (61 
FR 47977) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 24, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Unit Nos. 1 and 2, Pope County, Arkansas

    Date of amendment request: May 9, 1996
    Brief description of amendments: The amendments revised the name 
from Arkansas Power & Light Company to Entergy Arkansas, Inc.
    Date of issuance: October 23, 1996
    Effective date: October 23, 1996
    Amendment Nos.: 187 and 177
    Facility Operating License Nos. DPR-51 and NPF-6. Amendments 
revised the Technical Specifications and the licenses.
    Date of initial notice in Federal Register: August 28, 1996 (61 FR 
44357) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 23, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi 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 8, 1996, as supplemented by 
letters dated July 18 and September 19, 1996
    Brief description of amendment: The amendment modified the 
frequency requirements in Surveillance Requirement 3.6.1.3.5 of the 
Technical Specifications, on the leakage rate testing for each 
containment purge

[[Page 57494]]

isolation valve with resilient seals, to place these purge valves on a 
performance basis in accordance with Appendix J of 10 CFR Part 50, as 
modified by any exemptions to Appendix J. In addition, the purge valves 
would be required to be leakage rate tested every 36 months with at 
least two pairs tested every 18 months and, if any purge valve fails to 
meet the leakage rate acceptance criterion, all remaining valves must 
be tested within 92 days (i.e., a quarter of a year) if not 
successfully tested within the previous 92 days.
    Date if issuance: October 18, 1996
    Effective date: October 18, 1996
    Amendment No.: 128
    Facility Operating License No. NPF-29. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28614) The additional information contained in the supplemental letters 
dated July 18 and September 19, 1996, revised the proposed amendment in 
the application of May 8, 1996; however, the revisions were within the 
scope of the initial notice and did not affect the staff's proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 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.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi 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: June 20, 1996, as supplemented 
by the letter of September 11, 1996
    Brief description of amendment: The amendment redefined the 
secondary containment boundary to allow the enclosure building to be 
inoperable during the upcoming refueling Outage 8 (RFO 8) scheduled to 
begin in October 1996. The amendment added a condition to the license 
that the enclosure building may be inoperable during core alterations 
and movement of non-recently irradiated fuel (i.e., fuel that has not 
occupied part of a critical reactor core for 12 days) during RFO 8 and 
the standby gas treatment (SGT) system may be unable to automatically 
start or achieve and maintain the required vacuum, provided the 
following conditions exist:
    a. All dampers communicating between the auxiliary building and the 
enclosure building are closed.
    b. The access door between the auxiliary building and the enclosure 
building is closed, except when the access opening is being used for 
entry and exit.
    c. The SGT system is blocked from automatic initiation.
    d. The SGT system is available for manual initiation or the actions 
for Limiting Condition for Operation 3.6.4.3 in the Technical 
Specifications for GGNS are complied with.
    The non-recently irradiated fuel is spent fuel that has decayed at 
least 12 days after the reactor was shut down for refueling.
    Date of issuance: October 18, 1996
    Effective date: October 18, 1996
    Amendment No: 129
    Facility Operating License No. NPF-29. Amendment adds a condition 
to the license.
    Date of initial notice in Federal Register: July 17, 1996 (61 FR 
37299) The additional information contained in the supplemental letter 
of September 11, 1996, was clarifying in nature and thus, within the 
scope of the initial notice and did not affect the staff's proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated October 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.

Illinois Power Company and Soyland Power Cooperative, Inc., Docket 
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, 
Illinois

    Date of application for amendment: February 22, 1996, and as 
supplemented by letters dated July 22 and September 20, 1996
    Brief description of amendment: The amendment revises Clinton Power 
Station Technical Specification 3.4.11, ``Reactor Coolant System (RCS) 
Pressure and Temperature (P/T) Limits,'' to incorporate specific P/T 
limits for the bottom head region of the reactor vessel, separate and 
apart from the core beltline region of the reactor vessel.
    Date of issuance: October 23, 1996
    Effective date: October 23, 1996
    Amendment No.: 109
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 24, 1996 (61 FR 
18169) The letters of July 22 and September 20, 1996, provided 
clarifying information and did not alter the staff's initial finding 
that the proposed changes involve no significant hazards consideration. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated October 23, 1996.No significant hazards 
consideration comments received: No
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727

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

    Date of application for amendment: November 21, 1995
    Brief description of amendment: The amendment changes Technical 
Specification Section 5.2.2, ``Design Pressure and Temperature,'' to 
clarify that the reactor containment design temperature is an 
equilibrium liner temperature and not the air temperature. The 
supporting Technical Specification Bases is updated to reflect the 
change and to include the main steam line break accident, in addition 
to the loss-of-coolant accident, as the limiting events affecting the 
containment temperature and pressure.
    Date of issuance: October 21, 1996
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 204
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65684) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 21, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385

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

    Date of amendment requests: July 15, 1996, as supplemented by 
letters dated September 3, 1996, October 22, 1996, October 23, 1996, 
and August 23, 1996
    Brief description of amendment: The amendment revises Technical 
Specifications (TS) Section 4.3.2 to allow the use of zircaloy or ZIRLO 
fuel

[[Page 57495]]

cladding and to use depleted uranium as reactor fuel material. The 
amendment also changes TS Section 5.9.5 to add Westinghouse Topical 
Reports, WCAP-12610-P-A, ``VANTAGE + Fuel Assembly Report,'' and WCAP-
13027-P, ``Westinghouse ECCS Evaluation Model for Analysis of CE-
NSSS,'' to the list of approved analytical methods for determining the 
core operating limits.
    Date of issuance: October 25, 1996
    Effective date: October 25, 1996
    Amendment No.: 178
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
40026) and August 30, 1996 (61 FR 45995). The September 3, 1996, 
October 22, 1996, and October 24, 1996, supplemental letters provided 
additional clarifying and correcting information and did not change the 
initial no significant hazards consideration determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated October 25, 1996.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis 
Obispo County, California

    Date of application for amendments: June 7, 1996
    Brief description of amendments: The amendments revised the 
combined Technical Specifications (TS) for the Diablo Canyon Nuclear 
Power Plant, Unit Nos. 1 and 2 by revising Technical Specifications 3/
4.9.14.1, ``Spent Fuel Assembly Storage - Spent Fuel Pool Region 2,'' 
and TS 3/4.9.14.3, ``Spent Fuel Assembly Storage - Spent Fuel Pool 
Region 1,'' to allow storage of fuel assemblies in a checkerboard 
pattern in Region 2 of the spent fuel pool (SFP).
    Date of issuance: October 25, 1996
    Effective date: October 25, 1996, to be implemented within 30 days 
from date of issuance.
    Amendment Nos.: Unit 1 - 116; Unit 2 - 114
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 25, 1996 (61 
FR 50346) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 25, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis 
Obispo County, California

    Date of application for amendments: December 19, 1995, as 
supplemented by letter dated August 8, 1996.
    Brief description of amendments: The amendments revised the 
combined Technical Specifications (TS) for the Diablo Canyon Power 
Plant Unit Nos. 1 and 2 to relocate Technical Specification (TS) 6.5, 
``Review and Audit,'' 6.8, ``Procedures and Programs,'' Sections 
6.8.1c., 6.8.1d., 6.8.2, and 6.8.3, in accordance with guidance in an 
NRC letter dated October 25, 1993, from William T. Russell to the 
chairpersons of industry owners groups and the Commission's Final 
Policy Statement on TS Improvements for Nuclear Power Reactors on 
relocation of TS that do not satisfy the retention criteria. As part of 
the relocation of TS 6.8.2, TS 6.1.1 would be revised to require that 
proposed tests, experiments, or modifications that affect nuclear 
safety be approved by the plant manager or his designee prior to 
implementation.
    Date of issuance: October 25, 1996
    Effective date: October 25, 1996, to be implemented within 90 days 
of issuance.
    Amendment Nos.: Unit 1 - 117; Unit 2 - 115
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1633) The August 8, 1996, supplemental letter provided additional 
clarifying information and did not change the staff's initial no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated October 25, 1996.No significant hazards consideration 
comments received: No.
    Local Public Document Room location:  California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

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

    Date of application for amendments: May 20 and 28, 1996, as 
supplemented by letter dated July 25, 1996
    Brief description of amendments: These amendments, for both units, 
add a reference to the ANF-B critical power correlation to Section 
6.9.3.2 of the Technical Specifications (TSs); change the values of the 
minimum critical power ratio (MCPR) in TS Sections 2.1 and 3.4.1.1.2, 
and make appropriate Bases changes. For Unit 1 only, a reference to ABB 
licensing methodology report CENPD-300 (for lead use assemblies being 
used in the reactor core during the upcoming operating cycle) is added 
to Section 6.9.3.2.
    Date of issuance: October 11, 1996
    Effective date: For both units, as of date of issuance, to be 
implemented within 30 days.
    Amendment Nos.: 161 and 132
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: Unit 2, August 28, 1996 
(61 FR 44362); Unit 1, September 4, 1996 (61 FR 47529)The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated October 11, 1996.No significant hazards consideration 
comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

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

    Date of application for amendment: March 29, 1996, as supplemented 
July 12, 1996, and September 6, 1996.
    Brief description of amendment: The proposed amendment would change 
the Indian Point 3 Technical Specifications (TSs) relating to minimum 
reactor coolant system (RCS) flow and maximum RCS average temperature 
to make these parameters consistent with an assumption of 100% helium 
release from the boron coating of the integral fuel burnable absorber 
rods. The proposed amendment would also add limits associated with 
Departure from Nucleate Boiling to the IP3 Technical Specifications 
TSs.
    Date of issuance: October 22, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days

[[Page 57496]]

    Amendment No.: 170
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 17, 1996 (61 FR 
37301) August 14, 1996 (61 FR 42283)The Commission's related evaluation 
of the amendment is contained in a Safety Evaluation dated October 22, 
1996.No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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 19, 1995, as supplemented 
by letters dated December 22, 1995, and March 26, 1996.
    Brief description of amendments: These amendments modify Technical 
Specification (TS) 3.3.8, ``Containment Purge Isolation Signal 
(CPIS),'' and TS 3.3.9, ``Control Room Isolation Signal (CRIS).'' The 
revisions are needed to (1) support the upgrading or replacement of 
existing radiation monitoring system with state-of-the-art equipment 
that will provide for greater operational flexibility and reliability, 
and (2) incorporate minor editorial changes to improve clarity of these 
TS sections.
    Date of issuance: October 8, 1996
    Effective date: October 8, 1996, to be implemented within 30 days 
of date of issuance
    Amendment Nos.: Unit 2 - Amendment No. 132; Unit 3 - Amendment No. 
121
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49948). The December 22, 1995, and March 26, 1996, letters provided 
additional clarifying information and did not change the initial no 
significant hazards consideration determination.The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated October 8, 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

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

    Date of amendment request: March 29, 1996, as supplemented by 
letters dated June 27, August 29, and September 16, 1996.
    Brief description of amendment: The amendment changes Technical 
Specification 3/4.4.6, ``Steam Generators'' and associated Bases to 
modify the steam generator repair limit to clarify that the appropriate 
method for determining serviceability for tubes with outside diameter 
stress corrosion cracking at the tube support plate is by a methodology 
that more reliably assesses structural integrity.
    Date of issuance: October 11, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment No.: 115
    Facility Operating License No. NPF-8: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25711) The June 27, August 29, and September 16, 1996, letters provided 
additional, clarifying information that did not change the scope of the 
March 29, 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 October 11, 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 Nuclear Operating Company, Inc., Docket No. 50-364, Joseph 
M. Farley Nuclear Plant, Unit 2, Houston County, Alabama

    Date of amendment request: April 22, 1996, as supplemented by 
letters dated May 3, July 15, August 7 and 30, and September 16, 1996
    Brief description of amendment: The amendment changes reflect the 
implementation of a new F* criterion based on maintaining existing 
safety margins for steam generator tube structural integrity concurrent 
with allowances for nondestructive examination eddy current 
uncertainty.
    Date of issuance: October 11, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment No.: 116
    Facility Operating License No. NPF-8: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25713) The May 3, July 15, August 7 and 30, and September 16, 1996, 
letters provided clarifying information that did not change the scope 
of the April 22, 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 October 11, 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

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear 
Plant, Unit 1, Rhea County, Tennessee

    Date of application for amendment: June 29, 1996
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Section 5.2.2.f to delete the sentence, ``The 
Operations Manager shall hold or have held an SRO [Senior Reactor 
Operator] license on a similar unit.'' The revision also indicates that 
the Operations Superintendent will hold a valid SRO license on this 
unit.
    Date of issuance: October 15, 1996
    Effective date: Octber 15, 1996
    Amendment No.: 4
    Facility Operating License No. NPF-90: Amendment revises the TS.
    Date of initial notice in Federal Register: September 11, 1996 (61 
FR 47983)The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 15, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402

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

    Date of amendment request: July 31, 1996 (TXX-96432) as 
supplemented by letters dated August 23 and 27 (TXX-96447 and TXX-
96451), and September 19, 1996 (TXX-96469).
    Brief description of amendments: The amendments (1) change the 
acceptance values for amperes and voltage for the 18 month surveillance 
test of the battery chargers; (2) clarify the meaning of the

[[Page 57497]]

term ``associated inverter'' used in the context of energizing 118-Volt 
AC Instrument Buses during MODES 1 through 6; and (3) delete the 
protection channel and the vital bus ratings for the 118-Volt AC 
Instrument Buses identified for MODES 1 through 4.
    Date of issuance: October 22, 1996
    Effective date: October 22, 1996
    Amendment Nos.: Unit 1 - Amendment No. 53; Unit 2 - Amendment No. 
39
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 28, 1996 (61 FR 
44363) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 22, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019.
    Dated at Rockville, Maryland, this 30th day of October 1996.
    For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II,Office of Nuclear Reactor 
Regulation
[Doc. 96-28372 Filed 11-5-96; 8:45 am]
BILLING CODE 7590-01-F