[Federal Register Volume 60, Number 216 (Wednesday, November 8, 1995)]
[Notices]
[Pages 56361-56378]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-11108]



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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

[[Page 56362]]

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 14, 1995, through October 27, 1995. 
The last biweekly notice was published on Wednesday, October 25, 1995 
(60 FR 54714).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. 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 8, 1995, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no 

[[Page 56363]]
significant hazards consideration. The final determination will serve 
to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendments request: October 3, 1995
    Description of amendments request: The amendment would delete the 
provisions relating to certain previous sale and leaseback transactions 
that were by added by Amendment No. 3 for NPF-51 and Amendment No. 1 
for NPF-74.
    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.
    This amendment request does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated because the proposed change is administrative in nature. 
The proposed change deletes Sections 2.B.(7)(a) and (b) of License 
No. NPF-51, and Sections 2.B.(6)(a) and (b) of License No. NPF-74. 
These sections describe the structure of the financing of El Paso's 
interest in Palo Verde, specifically authorizing sale and leaseback 
transactions. The proposed change does not affect the assumptions 
used in the accident
    analyses, nor does the proposed change result in changes to the 
physical configuration of the facility, design parameters, technical 
specifications, or operation and maintenance of the facility. 
Therefore, the amendment request 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.
    This amendment request does not create the possibility of a new 
or different kind of accident from any accident previously analyzed 
because the proposed change is administrative in nature. The 
proposed change deletes Sections 2.B.(7)(a) and (b) of License No. 
NPF-51, and Sections 2.B.(6)(a) and (b) of License No. NPF-74. These 
sections describe the structure of the financing of El Paso's 
interest in Palo Verde Units 2 and 3, specifically authorizing sale 
and leaseback transitions. The proposed change does not involve 
modifications to any of the existing equipment nor does the change 
affect operation or maintenance of the facility. Therefore, the 
amendment request does not create the possibility of a new or 
different kind of accident not previously analyzed.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    This amendment request does not involve a significant reduction 
in a margin of safety because it is administrative in nature. The 
proposed change deletes Sections 2.B.(7)(a) and (b) of License No. 
NPF-51, and Sections 2.B.(6)(a) and (b) of License No. NPF-74. These 
sections describe the structure of the financing of El Paso's 
interest in Palo Verde, specifically authorizing the sale and 
leaseback transactions. The proposed change does not involve changes 
to any existing plant equipment or accident analyses that provide 
for or establish margins of safety. There is no change to the 
operation or maintenance of the facility and the existing margins of 
safety are not changed by the proposed change. Therefore, the 
amendment request does not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999
    NRC Project Director: William H. Bateman

Baltimore Gas and Electric Company, Docket No. 50-318, Calvert 
Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

    Date of amendment request: October 2, 1995
    Description of amendment request: The proposed amendment would 
revise the Calvert Cliffs Nuclear Power Plant, Unit No. 2, Technical 
Specifications on a one-time basis by increasing the 7 day allowed 
outage time (AOT) of the control room emergency ventilation system 
(CREVS) to an AOT of 30 days. This requested one-time increase in the 
AOT is applicable only for the loss of the emergency power supply to 
one train of the CREVS during the Unit No. 1 spring 1996 refueling 
outage.
    The requested extension in the AOT is necessary to allow the 
licensee to perform modifications to the electrical distribution system 
during the upcoming Unit 1 refueling outage while Unit No. 2 continues 
to operate. The modifications include connecting a fourth safety-
related (SR) emergency diesel generator (EDG) to engineered safety 
features (ESF) Bus No. 11. The work related to this effort will require 
that the bus be deenergized for several days isolating it from its 
normal and 

[[Page 56364]]
emergency EDG power supplies. One train of the CREVS is connected to 
ESF Bus No. 11 and will not have its power supplies available for a 
period of time. The normal (offsite) power is expected to be restored 
in about 3 days, but the emergency power (onsite EDG) may take up to 30 
days.
    The licensee is taking additional actions to assure the 
availability of the normal offsite power source and is also adding a 
nonsafety-related (NSR) EDG as an alternate onsite power source during 
the period that the SR EDG is not available. The licensee expects that 
the tie-in of the NSR EDG will take about 8 days. Thus, even if the 
normal offsite power source is lost, the temporary onsite NSR EDG will 
be available to provide power to the affected train of the CREVS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10CFR 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 previously evaluated.
    The Control Room Emergency Ventilation System (CREVS) is used to 
mitigate the consequences of an accident. It is designed so that the 
Control Room remains habitable for operators and to maintain the 
environment needed for continued equipment operation. The system is 
redundant (two 100% capacity trains) and is powered from both normal 
(offsite) and emergency (emergency diesel generators) power sources. 
We [the licensee] are proposing an amendment which would allow the 
emergency power to be removed from one of the redundant CREVS for an 
additional 23 days (beyond the 7 days allowed by the Technical 
Specifications). Other than the removal of the emergency electrical 
power source, we are not affecting or modifying the operation of the 
CREVS. The CREVS is not an accident initiator for any previously 
evaluated accident. Therefore, the proposed change does not involve 
an increase in the probability of an accident previously evaluated.
    The CREVS is designed to mitigate the consequences of design 
basis accidents. For that purpose, redundant trains are provided to 
protect against a single failure. During the Technical Specification 
seven day Allowed Outage Time (AOT), an operating unit is allowed by 
the Technical Specifications to remove one of the CREVS trains from 
service, thereby eliminating this single failure protection. The 
consequences of a design basis accident coincident with a failure of 
the redundant CREVS train during the additional 23-day period are 
the same as those during the 7-day AOT. Therefore, the proposed 
change does not significantly increase the consequences of an 
accident previously evaluated.
    Therefore, the proposed change does not increase the probability 
or consequences of an accident previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The CREVS is not being modified by this proposed change nor will 
any unusual operator actions be required. The system will continue 
to operate in the same manner. The CREVS is not an initiator to any 
accident, but is designed to respond should an accident occur.
    Therefore, the proposed change does 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 a margin of 
safety.
    The operability of the CREVS during Modes 1 through 4 ensures 
that the Control Room will remain habitable for operators and to 
maintain the environment needed for continued equipment operation 
under all plant conditions. The proposed change does not affect the 
function of the CREVS. During the period of the Technical 
Specifications AOT when one CREVS train is inoperable, the margin of 
safety is reduced. This time period is a temporary relaxation of the 
single failure criteria, which, consistent with overall system 
reliability considerations, provides a limited time to maintain or 
repair the equipment and conduct testing. We are requesting an 
extension of this limited time. The proposed change will allow one 
train of the CREVS to be without an emergency power supply for an 
additional 23 days beyond the 7-day AOT (total of 30 days). This 
train of CREVS will be functional and will have the normal power 
supply available for all but approximately three days to allow work 
and necessary testing on the bus. The other train of the CREVS will 
have both its normal and emergency power supplies during this 
period.
    To provide additional assurance that all reasonable steps have 
been taken to prevent the loss of the normal power supply to the 
CREVS, we will restrict maintenance activities on three of the four 
offsite transmission lines. This restriction will cover the period 
we are in the Action Statement for the CREVS (Action Statement 
3.7.6.1.a and b). To provide an alternative power source during the 
majority of this period, we will connect the Alternate AC power 
source (No. 0C Diesel Generator) to ESF Bus No. 11 and confirm its 
availability as soon as possible after the work on ESF Bus No. 11 
begins (we [the licensee] expect that to take about eight days). 
This power source is independent from the offsite power supplies. In 
addition, we will restrict planned maintenance on the No. 12 CREVS 
during the period we are in the Action Statement to ensure that the 
No. 12 CREVS is not removed from service.
    We believe that the reduction in the margin of safety 
represented by this one-time extension of the AOT is not significant 
based on our management of plant risk, the reliability of the normal 
CREVS power supply, the availability of the redundant CREVS with 
both its normal and emergency power, and the mitigating features 
described above. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Ledyard B. Marsh

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

    Date of amendments request: October 23, 1995
    Description of amendments request: The amendments would delete the 
applicability of the primary coolant water chemistry limits when the 
primary system is being chemically decontaminated and the reactor 
vessel is defueled.
    Basis for proposed no significant hazards consideration 
determination: As required by 10CFR 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 will allow the reactor coolant system 
conductivity and chlorides to exceed the limits specified in 
Technical Specification Table 3.4.4-1 in support of performing 
chemical decontamination activities. The reactor coolant system 
water chemistry limits have been established to prevent long-term 
damage to the reactor coolant system materials that are in contact 
with the coolant. Upon concluding the chemical decontamination 
activities, reactor coolant system conductivity and chloride values 
would be restored to within the limits specified in Technical 
Specification Table 3.4.4-1. Existing regulatory requirements, 
specifically a review in accordance with 10 CFR 50.59 to determine 
whether an activity involves an unreviewed safety question, provide 
adequate assurance that solvents selected for use in a chemical 
decontamination activity will not degrade the structural integrity 
of the reactor coolant system. Therefore, since the structural 
integrity of the reactor coolant system will not be adversely 
impacted by the chemical decontamination activities, the proposed 
amendments do not involve a significant increase in the probability 
of an accident previously evaluated.
    As discussed above, the reactor coolant system water chemistry 
limits have been 

[[Page 56365]]
established to prevent long-term damage to the reactor coolant system 
materials that are in contact with the coolant. The solvents being 
used for a chemical decontamination activity are selected to ensure 
their effectiveness and to ensure that damage will not occur to the 
structural materials comprising the reactor coolant pressure 
boundary. As such, the operation of safety equipment used to 
mitigate a design basis accident or transient will not be affected 
by the proposed change of the reactor coolant system water chemistry 
limits during performance of chemical decontamination activities. 
Therefore, the proposed revision to the reactor coolant system 
chemistry limits will not involve a significant increase in the 
consequences of an accident previously evaluated.
    2. The proposed change will allow the reactor coolant system 
conductivity and chlorides to exceed the limits specified in 
Technical Specification Table 3.4.4-1 in order to perform chemical 
decontamination activities. The reactor coolant system water 
chemistry limits have been established to prevent long-term damage 
to the reactor coolant system materials that are in contact with the 
coolant. Even though the solvents used for chemical decontaminations 
may result in reactor coolant system conductivity and chloride 
measurement values in excess of the limits specified in the 
Technical Specifications, the existing regulatory requirements of 10 
CFR 50.59 will continue to ensure that solvents being used for 
performing chemical decontamination have been properly evaluated and 
that these solvents do not adversely affect the material properties 
or structural integrity of the reactor coolant system. Therefore, 
the proposed amendments revising the reactor coolant system water 
chemistry limits during performance of chemical decontamination 
activities will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The reactor coolant system water chemistry limits have been 
established to prevent long-term damage to the reactor coolant 
system materials that are in contact with the coolant. The solvents 
used for chemical decontaminations result in reactor coolant system 
conductivity and chloride measurement values in excess of the limits 
specified in the Technical Specifications; however, the solvents 
being used of performing chemical decontamination have been properly 
evaluated to ensure they will not significantly affect the material 
properties of the reactor coolant system piping (i.e., corrosion) 
nor will they significantly affect the structural integrity (i.e., 
wall thinning) of the reactor coolant system piping. Therefore, the 
proposed license amendments 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: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: David B. Matthews

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: November 2, 1994, as supplemented on 
January 4, 1995
    Description of amendment request: The amendment would revise the 
Technical Specifications (TSs) to make editorial changes, delete 
portions of the TSs that have become unnecessary due to previously 
approved amendments, change managerial titles, update references and 
reporting requirements, revise the Station Nuclear Safety Committee 
(SNSC) composition to specify disciplines rather than specific job 
titles, modify the record keeping requirements of the Nuclear 
Facilities Safety Committee, implement changes referenced in Generic 
Letter 93-07, ``Modification of the Technical Specification 
Administrative Control Requirements for Emergency and Security Plans,'' 
and to correct the shift manning requirements table.
    Basis for proposed no significant hazards consideration 
determination: As required by 10CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. There is no significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed amendments are administrative in nature. They 
involve making editorial changes, deleting portions of the Technical 
Specifications that have become unnecessary due to previously 
approved amendments, changing managerial titles, updating references 
and reporting requirements, revising the SNSC composition to specify 
disciplines rather than specific job titles, implementing changes 
referenced in Generic Letter 93-07, and revising shift manning to 
conform with the requirements of 10 CFR 50.54. These changes do not 
affect possible initiating events for accidents previously evaluated 
or alter the configuration or operation of the facility. The 
Limiting Safety Systems Settings and Safety Limits specified in the 
current Technical Specifications remain unchanged. Therefore, the 
proposed changes to the subject Technical Specifications would not 
increase the probability or consequences of an accident previously 
evaluated.
    2. The possibility of a new or different kind of accident from 
any accident previously evaluated has not been created.
    As stated above, the proposed changes are administrative in 
nature. The safety analysis of the facility remains complete and 
accurate. There are no physical changes to the facility and the 
plant conditions for which the design basis accidents have been 
evaluated are still valid. The operating procedures and emergency 
procedures are unaffected. Consequently, no new failure modes are 
introduced as a result of the proposed changes. Therefore, the 
proposed changes would not initiate any new or different kind of 
accident.
    3. There has been no significant reduction in the margin of 
safety.
    The proposed changes are administrative in nature. Since there 
are no changes to the physical design or operation of the facility, 
the Updated Final Safety Analysis Report (UFSAR) design basis, 
accident assumptions, or Technical Specification Bases are not 
affected. Therefore, the proposed changes would not result in a 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003
    NRC Project Director: Ledyard B. Marsh

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: August 29, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specification Sections 3.1.F and 4.13 to provide for 
appropriate inservice inspection for any steam generator tubes 
containing sleeves and to provide for reduced allowable primary-to-
secondary leakage rates for steam generators containing sleeves. The 
proposed changes are in response to commitments made by Consolidated 
Edison by letter dated April 5, 1995, during the review of an amendment 
which permitted the use of laser welded steam generator tube sleeves as 
a method of tube repair.
    Basis for proposed no significant hazards consideration 
determination: As required by 10CFR 50.91(a), the licensee has provided 
its analysis of the 

[[Page 56366]]
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.
    Technical Specification Amendment No. 183 allowed sleeving as an 
acceptable alternate tube repair method for Indian Point Unit No. 2. 
The steam generator sleeve approved for installation is the 
Westinghouse process (laser welded sleeve). The sleeve configuration 
was designed and analyzed in accordance with the criteria of 
Regulatory Guide (RG) 1.121 and the design requirements of Section 
III of the American Society of Mechanical Engineers (ASME) Code. 
Fatigue and stress analyses of the sleeved tube assembly produced 
acceptable results as documented in the Westinghouse topical report 
submitted in the original sleeving package. Mechanical testing has 
shown that the structural strength of the sleeves under normal, 
faulted, and upset conditions is within acceptable limits. Leakage 
rate testing for the tube sleeves has demonstrated that primary-to-
secondary leakage is not expected during all plant conditions.
    Any leakage through the sleeved region of the tube is fully 
bounded by the leak-before-break considerations and, ultimately, the 
existing steam generator tube rupture analysis included in the 
Updated Final Safety Analysis Report (UFSAR).
    The reduction in TS leakage rate requirements from 0.3 gpm 
[gallons per minute] (432 gpd [gallons per day]) allowable per SG to 
150 gpd per steam generator containing sleeves further ensures that 
SG tube integrity is maintained in the event of a main steam line 
break (MSLB) or under Loss Of Coolant Accident (LOCA) conditions. 
The RG 1.121 criteria for establishing operational leakage rate 
limits require a plant shutdown based upon a leak-before-break 
consideration to detect a free span crack before a potential tube 
rupture. The 150 gpd limit will continue to allow for early leakage 
detection and require a plant shutdown in the event of tube leakage 
that exceeds the revised Technical Specification limit.
    The sleeve sample size has been increased to a minimum of twenty 
(20) percent of the inservice sleeves. Increasing the sample size of 
the sleeves to be inspected will increase the monitoring of tubes 
using sleeves for any further degradation while they remain 
inservice. If the sample identifies a sleeve with an imperfection of 
greater than 23 percent depth an additional 20 percent of the 
sleeves shall be inspected. The sleeves that have identified 
imperfections of greater than 23 percent shall be evaluated and 
removed from service.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Implementation of the proposed amendment will not introduce 
significant or adverse changes to the plant design basis. The 
proposed changes do not involve plant modification or changes to 
equipment, and consist of reducing the allowable steam generator 
leakage limits for steam generators containing sleeves and defining 
the sample size of the steam generator tube sleeve inspection.
    The reduction in TS leakage rate requirements from 0.3 gpm (432 
gpd) allowable per SG to 150 gpd per SG containing sleeves further 
ensures that SG tube integrity is maintained in the event of a MSLB 
or under LOCA conditions. The 150 gpd limit is designed to provide 
for leakage detection and a plant shutdown in the event of the 
concurrence of excessive tube leakage. The limit provides for early 
detection and a plant shutdown prior to a postulated defect reaching 
critical magnitudes for Main Steam Line Break conditions.
    Formalizing the sample size of sleeved tubes inspected during 
each scheduled inservice inspection will ensure increased monitoring 
of these tubes for any further degradation. The improved monitoring 
and evaluation of the tube and the sleeves assures tube structural 
integrity is maintained or the tube is removed from service.
    With these actions the possibility of a new or different type of 
accident from any accident previously evaluated is not created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Implementation of the proposed changes will not reduce the 
margin of safety. This amendment involves the reduction of sleeved 
steam generator tube leakage limit and a formalized inservice 
inspection program for sleeved tubes. These actions will help ensure 
steam generator tube integrity.
    Reduction of the leakage rate requirement from 0.3 gpm (432 gpd) 
to 150 gallons per day (gpd) per sleeved steam generator will 
continue to ensure steam generator tube integrity is maintained in 
the event of main steam line break or under LOCA conditions. 
Reducing this limit will not result in a reduction in the margin of 
safety.
    The portions of the installed sleeve assembly which represent 
the reactor coolant pressure boundary will be monitored for the 
initiation and progression of sleeve/tube wall degradation, thus 
satisfying the requirement of Regulatory Guide 1.83. The portion of 
the tube bridged by the sleeve joints is effectively removed from 
the pressure boundary, and the sleeve then forms the new pressure 
boundary. The sleeve enhances the safety of the plant by increasing 
the protective boundaries of the steam generator. Keeping the tube 
in service with the use of a sleeve, instead of plugging the tube 
and removing it from service, increases the heat transfer efficiency 
of the steam generator. Monitoring for any increased degradation of 
a repaired steam generator tube shall be implemented by sampling 
twenty (20) percent of the sleeves inservice. During each scheduled 
inservice inspection, any sampled sleeve evaluated and found to have 
unacceptable degradation shall be removed from service.
    Based on the preceding analysis it is concluded that operation 
of Indian Point Unit No. 2 in accordance with the proposed amendment 
does not increase the probability of an accident previously 
evaluated, does not create the possibility of a new or different 
kind of accident from any accident previously evaluated, nor reduce 
any margin of plant safety. Therefore, the license amendment does 
not involve a Significant Hazards Consideration as defined in 10 CFR 
50.92.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003
    NRC Project Director: Ledyard B. Marsh

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

    Date of amendment request: September 30, 1994, as supplemented by 
letter dated September 19, 1995
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) related to the replacement of 
the steam generators at McGuire, Units 1 and 2. Currently, the steam 
generators in place at the McGuire units are Westinghouse Model ``D'' 
type preheat steam generators. The tube degradation levels in the 
generators has affected the reliability of the units. Therefore, these 
generators are scheduled to be replaced with feedring steam generators 
designed by Babcock and Wilcox International.
    In the licensee's September 19, 1995, supplement, proposed changes 
were made to TS Table 2.2-1, ``Reactor Trip System Instrumentation Trip 
Setpoints,'' to change the programmed T<INF>AVG from 588.2  deg.F to 
585.1  deg.F. This temperature was chosen based on returning the 
secondary side steam pressure to the original value after replacement 
of the steam generators. The licensee stated that 585.1  deg.F was the 
assumed value for nominal full power T<INF>AVG in all applicable safety 
analyses related to replacement of the steam generators.
    The licensee also requested that the steam line safety valve lift 
settings in Table 3.7-3, which was requested in the September 30, 1994, 
application, be withdrawn. The licensee determined that these changes 
are no longer needed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards 

[[Page 56367]]
consideration, which is presented below:
    Operation of McGuire Nuclear Station in accordance with the 
proposed changes to the Technical Specifications will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. The low-low steam generator water 
level reactor trip setpoint, the high-high steam generator water 
level setpoint for turbine trip and feedwater isolation, and the 
low-low steam generator water level setpoint for auxiliary feedwater 
initiation are changing to support operation with the replacement 
steam generators. These setpoints were chosen both to optimize plant 
operation, and ensure that all applicable acceptance criteria are 
met for licensing basis safety analysis. These setpoints do not 
contribute to the initiation of any accident evaluated in the 
McGuire FSAR [Final Safety Analysis Report] and have no adverse 
impact on system operation, therefore it can be concluded that these 
changes will not significantly increase the probability or 
consequences of an accident evaluated in the FSAR.
    The reduction in the primary to secondary leakage rate for 
McGuire will not increase the probability of an accident evaluated 
in the FSAR. This lower limit will require corrective action more 
quickly than is currently required in the event that there is a 
steam generator tube leak. This change will not significantly affect 
the consequences of an accident previously evaluated. The allowable 
leakage is being lowered because this leakage has a major impact on 
the results of the offsite dose calculation for the locked rotor, 
single uncontrolled rod withdrawal, and rod ejection events. The 
taller tube bundle in the replacement steam generators will 
potentially result in a longer period of tube bundle uncovery during 
the above transients. The revised allowable leakages of 0.27 gpm 
through all steam generators and 135 gallons per day through any one 
generator ensure that the dose analysis results are within the 
applicable fraction 10 CFR 100 limits.
    The increase in Reactor Coolant System volume due to the 
replacement steam generators will not increase the probability or 
consequences of an accident previously evaluated. The increase in 
volume has no effect on the probability of occurrence of any 
accident evaluated in the FSAR. The mass and energy release due to 
postulated loss of coolant accidents inside containment has been 
analyzed to ensure that the peak containment pressure limit is not 
exceeded. All Chapter 15 reanalysis which was required due to the 
replacement steam generators assumed the new Reactor Coolant System 
volume. Since the results of these analyses show the applicable 
acceptance criteria continue to be met, it can be concluded that the 
consequences of an accident previously evaluated are not 
significantly increased due to this change.
    * * * *
    Operation of McGuire Nuclear Station in accordance with the 
proposed changes to the Technical Specification will not create the 
possibility of a new or different accident from any accident 
previously evaluated. The proposed changes to revise the low-low 
steam generator water level reactor trip setpoint, high-high steam 
generator water level setpoint for turbine trip and feedwater 
isolation, and low-low steam generator water level setpoint for 
auxiliary feedwater initiation ensure that the appropriate 
acceptance criteria for FSAR Chapter 15 transients which rely on 
these functions are met for operation with the replacement steam 
generators. The proposed change to lower primary to secondary 
leakage for operation with the replacement steam generators will 
require that corrective action be taken more quickly in the event 
that steam generator tube leakage is experienced during operation. 
As discussed in the technical justification, this will cause the 
dose results for transients affected by tube bundle uncovery to be 
within acceptable limits. .... The increase in Reactor Coolant 
System volume is taken into account in the analysis of the mass and 
energy release due to a postulated loss of coolant inside 
containment and Chapter 15 events which have been reanalyzed due to 
replacement of the steam generators. As discussed above, the 
proposed changes will not introduce the possibility of a new or 
different accident from any previously evaluated; they will ensure 
that transients that take credit for these functions and dose 
analyses meet applicable acceptance criteria for operation with the 
replacement steam generators.
    Operation of McGuire Nuclear Station in accordance with the 
proposed changes to the Technical Specifications will not involve a 
significant reduction in a margin of safety. The proposed changes 
are being made to ensure that transients that rely on low-low steam 
generator water level reactor trip setpoint, high-high steam 
generator water level setpoint for turbine trip and feedwater 
isolation, and low-low steam generator water level setpoint for 
auxiliary feedwater actuation meet applicable acceptance criteria. 
The reduction in allowable primary to secondary leak rate will 
ensure that transients with dose analyses which are affected by the 
replacement steam generators meet the current acceptable limits. 
.... The proposed change in the Reactor Coolant System volume will 
not involve a significant reduction in a margin of safety. The 
increased volume affects the mass and energy release due to a 
postulated loss of coolant accident inside containment and the other 
Chapter 15 events which were reanalyzed due to replacement of the 
steam generators. These events have been analyzed and the results 
are within current acceptable limits. As discussed above, the 
acceptance criteria for FSAR transients which are affected by these 
proposed changes continue to be met, therefore there is no 
significant reduction in the margin of safety.
    Changes to the steam generator surveillance requirements will 
simply delete inspection requirements and repair methods which are 
no longer applicable after installation of the replacement steam 
generators. The only exception to this is Surveillance Requirement 
4.4.5.4.a.9. This requirement is modified to clarify that the 
manufacturer will perform the hydrostatic test for the replacement 
steam generators. This change will not affect the probability or 
consequences of an accident previously evaluated, the purpose of the 
preservice inspection is to establish the baseline condition of the 
tubing. The baseline condition of the tubing in the replacement 
steam generators will be established prior to installation. The 
possibility of a new or different accident from any previously 
evaluated will not be created. No new accident initiation mechanisms 
will be introduced by this change, and the intent of the 
requirement, to establish the baseline condition of the tubing, will 
be met. Since the baseline condition of the tubing will be obtained 
for use in the monitoring of tubing degradation, as is currently 
required by the surveillance requirement, there will not be a 
significant reduction in the margin of safety.
    The changes to Technical Specification 6.9.1.9 are 
administrative in nature. These changes are being made to reflect 
the most recent revisions of DPC-NE-3002 and DPC-NE-3000, which 
include changes associated with the replacement steam generators. 
These topical reports revisions will be reviewed and approved for 
use regarding Catawba and McGuire Nuclear Stations. Since these 
changes are administrative in nature, no significant hazards 
considerations are involved.
    Proposed revision to TS Table 2.2-1, Reactor Trip System 
Instrumentation Trip Setpoints:
     proposed change to the Technical Specifications does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated. Changing the value for T<INF>AVG 
in Notes 1 and 2 of Table 2.2-1 will update the value to agree with 
the T<INF>AVG assumed in the applicable safety analyses for 
replacement of the steam generators. Acceptable results were 
obtained for all required reanalyses. The probability of an accident 
will not be significantly affected by operation with the new 
T<INF>AVG value, because all equipment will be operated within 
acceptable design limits. The consequences of previously evaluated 
accidents which are affected by this change have been evaluated, and 
have been determined to be within acceptable limits.
    This proposed change will not create the possibility of a new or 
different kind of accident from any previously evaluated. This 
change does not change the physical configuration of the plant, and 
all analyses which are affected by replacement of the steam 
generators have been determined to have acceptable results assuming 
this value for T<INF>AVG.
    This proposed change to the Technical Specifications will not 
involve a significant reduction in the margin of safety. All safety 
analyses which were affected by replacement of the steam generators 
assumed this value for T<INF>AVG and the results were determined to 
be within previously acceptable limits.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

[[Page 56368]]

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

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

    Date of amendment request: September 4, 1993, as supplemented on 
February 16, 1994, and August 4, 1995.
    Description of amendment request: The proposed amendment would 
revise the Arkansas Nuclear One Industrial Security Plan.
    Basis for proposed no significant hazards consideration 
determination: As required by 10CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration. The 
NRC staff has reviewed the licensee's analysis against the standards of 
10 CFR 50.92(c). The NRC staff's review is presented below.
    The accident mitigation features of the plant are not affected by 
the proposed compensatory measures for protecting the site during 
periods when security systems are degraded and therefore no decrease 
occurs in the effectiveness of the security program to protect against 
radiological sabotage or increased risk to the public health and 
safety. This is due to continued compliance with existing regulatory 
requirements and other commitments within the security plan. These 
changes have no impact on the design basis security threat and 
accordingly do not create the possibility of a new or different kind of 
accident. New systems, modes of equipment operation, failure modes or 
other plan situations are not introduced by these changes. The proposed 
changes allow flexibility for the use of compensatory measures and do 
not change any safety limits, LCOs, or surveillance requirements on 
equipment to operate the plant.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location:  Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: October 24, 1995, as supplemented or 
supercedes letters dated May 30, and June 20, 1995
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) on containment systems to 
reflect the adoption of requirements of 10 CFR Part 50, Appendix J, 
Option B, and implementation of a performance-based containment leak 
rate testing program at River Bend Station. The licensee letters dated 
May 20, and June 20, 1995, requested an exemption to Appendix J which 
subsequently became Option B to the appendix. Those letters were 
noticed in the Federal Register on July 5, 1995 (60 FR 35079).
    Basis for proposed no significant hazards consideration 
determination: As required by 10CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. This request does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any of the parameters or conditions that contribute to 
initiation of any accidents previously evaluated. Thus, the proposed 
change cannot increase the probability of any accident previously 
evaluated.
    The proposed change potentially affects the leak-tight integrity 
of the containment structure designed to mitigate the consequences 
of a loss-of coolant accident (LOCA). The function of the 
containment is to maintain functional integrity during and following 
the peak transient pressures and temperatures which result from any 
loss-of-coolant accident (LOCA)[LOCA]. The containment is designed 
to limit fission product leakage following the design basis LOCA. 
Because the proposed change does not alter the plant design, only 
the frequency of measuring Type B and C leakage, the proposed change 
does not directly result in an increase in containment leakage. 
However, decreasing the test frequency can increase the probability 
that a large increase in containment leakage could go undetected for 
an extended period of time. Based upon the results of the periodic 
containment Type A or Integrated Leak Rate Tests (ILRTs) and Type B 
and C or Local Leak Rate Tests (LLRTs) surveillance tests, this is 
not expected during the remaining life of the plant. The risk 
resulting from the proposed changes is as follows:
    Type A Testing
    NUREG/CR-4330 (NRC86) found that the effect of containment 
leakage on overall accident risk is small since risk is dominated by 
accident sequences that result in failure or bypass of the 
containment. It also determined that on an expected individual dose 
basis, the effect of containment leakage is small.
    Industry wide, ILRTs have only found a small fraction of the 
leaks that exceed current acceptance criteria. Only three percent of 
all leaks have a potential for remaining undetected for longer 
periods of time. In addition, when leakage has been detected by 
ILRTs, the leakage rate has been only about two times the allowable 
leakage rate.
    NUREG-1493 found that these observations, together with the 
insensitivity of reactor accident risk to the containment leakage 
rate, show that reducing the Type A leakage test frequency would 
have a minimal impact on public risk.
    Type B and C Testing
    NUREG-1493 found that while Type B and C tests can identify the 
vast majority (greater than 95 percent) of all potential leakage 
paths, performance-based alternatives to current local leakage-
testing requirements are feasible without significant risk impacts. 
The risk model used in NUREG-1493 suggests hat the number of 
components tested would be reduced by about 60 percent with less 
than a three-fold increase in the incremental risk due to 
containment leakage. Since, under existing requirements, leakage 
contributes less than 0.1 percent of overall accident risk, the 
overall impact is very small. NUREG-1493 found that while the 
extended testing intervals for Type B and C tests led to minor 
increases in potential offsite [off-site] dose consequences, the 
actual increase in on-site (worker) doses exceeded (by at least an 
order of magnitude) the potential off-site dose consequences.
    EPRI Research Project Report TR-104285, ``Risk Impact Assessment 
of Revised Containment Leak Rate Testing Intervals,'' also concluded 
that a relaxation of the test intervals for Type B and C 
penetrations results in a negligible increase in total plant risk.
    Based on the above EOI [Entergy Operation, Inc.] has concluded 
that the proposed change will not result in a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. The request does not create the possibility of occurrence of 
a new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any of the parameters or conditions that could contribute to 
initiation of any accidents. This change involves the reduction in 
Type B and C test frequency. The methods of performing the tests are 
not changed. No new accident modes are created by extending the 
testing intervals. No safety-related equipment or safety functions 
are altered as a result of this change. Extending 

[[Page 56369]]
the test frequency has no influence on , nor does it contribute to, the 
possibility of a new or different kind of accident or malfunction 
from those previously analyzed.
    3. The request does not involve a significant reduction in a 
margin to safety.
    The proposed change only affects the frequency of Type A, B, and 
C testing and does not change the methodology for performance of the 
testing. However, the proposed change can increase the probability 
that a large increase in leakage could go undetected for an extended 
period of time. Operational experience has shown that the leak 
tightness of the containment has been maintained significantly below 
the allowable leakage limit. In addition, NUREG-1493 has determined 
that, under several different accident scenarios, the risk of 
radioactivity release from containment is negligible with the 
implementation of these proposed changes.
    The margin of safety that has the potential of being impacted by 
the proposed change involves the offsite [off-site] dose 
consequences of postulated accidents which are directly related to 
containment leakage rate. The containment isolation system is 
designed to limit leakage to L<INF>a which is defined by the RBS 
Technical Specifications to be 0.26 percent by weight of the 
containment air per 24 hours at 7.6 psig (P<INF>a). The limitation 
on containment leakage rate is designed to ensure that total leakage 
volume will not exceed the value assumed in the accident analyses at 
the peak accident pressure (P<INF>a) or 7.6 psig. The margin to 
safety for the offsite [off-site] dose consequences of postulated 
accidents directly related to the containment leakage rate in 
maintained by meeting the 1.0 L<INF>a
    No change in the method of testing is being proposed. The Type B 
and C tests will continue to be done at full pressure (P<INF>a) or 
greater. Other programs are in place to ensure that proper 
maintenance and repairs are performed during the service life of the 
primary containment and systems and components penetrating the 
primary containment.
    As a result, EOI had concluded that the proposed change will not 
result in a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005
    NRC Project Director: William D. Beckner

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

    Date of amendment request: September 28, 1995
    Description of amendment request: The proposed change modifies 
Technical Specification 3/4.8.1.2, ``Electrical Power Sources - 
Shutdown.'' The surveillance requirement 4.8.1.2 is clarified by a Note 
to identify those surveillances which are required to be performed 
during Modes 5 and 6.
    Basis for proposed no significant hazards consideration 
determination: As required by 10CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    No component modification, system realignment, or change in 
operations will occur which could affect the probability of any 
accident or transient. The proposed addition of a Note will provide 
guidance on which surveillances are required to be performed in 
Modes 5 and 6. The Note will preclude rendering operable DGs 
inoperable, and/or preclude de-energizing a required ESF bus or 
disconnecting a required offsite circuit during the performance of 
the surveillance requirement. Proposed changes do not eliminate any 
testing requirements, they simply clarify which tests will be 
performed in Modes 5 and 6, and which are required to be performed 
prior to entry into Mode 4. Therefore, the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously analyzed.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    No component modification, system realignment, or change in 
operating procedure is required to implement the proposed change. 
The proposed change reduces the possibility of a single event 
impacting the operability of an ESF bus or its DG simultaneously. 
Therefore, these changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed change will not alter any assumptions, initial 
conditions, or results of any accident analyses. The Class 1E 
equipment assumed available in the accident analyses and their 
designed capability to mitigate the consequences of any postulated 
accidents will not be changed. The addition of a Note to clarify the 
surveillance requirements 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: Salem Free Public library, 112 
West Broadway, Salem, New Jersey 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

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

    Date of amendment request: September 28, 1995
    Description of amendment request: The proposed changes relocate 
``Reactor Coolant System - Chemistry'' Technical Specification 3/4.4.7 
(Salem Unit 1) and 3/4.4.8 (Salem Unit 2) and their associated Bases to 
the Salem Updated Final Safety Analysis Report (UFSAR) and the 
Surveillance Requirements and Limiting Conditions for Operation to 
applicable plant procedures controlled by the 10 CFR 50.59 process. 
Also, the applicability will be changed from ``At all times'' to 
``Modes 1, 2, 3, 4, 5 and 6.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes involve no hardware changes, no changes to 
the operation of any systems or components, and no changes to 
existing structures. Specifically, changing the Applicability from 
``At all times'' to ``Modes 1, 2, 3, 4, 5 and 6'' by this submittal 
will not alter established chemistry for chlorides, fluorides and 
dissolved oxygen of the Reactor Coolant System. The relocation of 
this Surveillance Requirement/LCOs and Bases to plant procedures and 
the UFSAR respectively, will continue to ensure that the chemistry 
analysis of the Reactor Coolant System water is monitored and 
controlled. Changing the Applicability from ``At all times'' to 
``Modes 1,2,3,4,5 and 6'' represent changes that do not affect plant 
safety and do not alter existing accident analyses.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed changes are procedural in nature concerning the 
location of the descriptive information and surveillance 
requirements for Reactor Coolant System Chemistry. Removing these 
specifications from the Technical Specifications and 

[[Page 56370]]
placing them in the UFSAR and plant procedures will not alter the 
maintenance of the Reactor Coolant System Chemistry or the ability 
to monitor its intended functions. Therefore, these changes will not 
create a new or unevaluated accident or operating condition.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed changes relocate the Reactor Coolant System 
Chemistry Requirements/LCOs from the Technical Specifications to the 
UFSAR and plant procedures in accordance with guidance provided by 
the NRC Final Policy Statement (58 FR 39132) regarding the 
improvement of Technical Specifications. The requirements that will 
reside in the UFSAR and plant procedures for the Reactor Coolant 
System Chemistry will ensure that the ability to determine chloride, 
fluoride and dissolved oxygen concentrations in the Reactor Coolant 
System is properly maintained and that the maintenance of the 
Reactor Coolant System Chemistry will be commensurate with its 
safety significance. Therefore, the proposed changes will not 
involve a significant reduction in any margins of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, New Jersey 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

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 26, 1995
    Description of amendments request: The amendments would revise 
Technical Specification (TS) Section 4.6.1.3 to incorporate 
improvements to containment air lock testing referenced in Chapter 3.6, 
``Containment Systems,'' of NUREG-1431, ``Standard Technical 
Specifications, WestinghousePlants.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10CFR 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 does not involve any change to the 
configuration or method of operation of any plant equipment used to 
mitigate the consequences of an accident. Containment leakage is an 
assumption in the safety analysis of the loss of coolant accident 
and the rod ejection accident. Changes to the containment air lock 
door seal test acceptance criteria will have no impact on the 
radiological consequences of these accidents since the plant safety 
analysis is based on the assumption that the containment leaks at 
its design leak rate of 0.15 percent per day for the first 24 hours 
and 0.075 percent per day thereafter for each of these accidents. 
The change to the surveillance requirement meets the intent of the 
guidance in NUREG-1431. Primary containment integrity ensures that 
the release of radioactive materials from the containment atmosphere 
will be restricted to those leakage paths and associated leak rates 
assumed in the accident analysis. The limitations on closure and 
leak rate for the containment air locks are required to meet these 
restrictions on containment integrity. These changes do not increase 
the probability that the 10 CFR [Part] 100 limits will be exceeded. 
The change to the surveillance requirement does not impose any new 
safety analyses limits or alter the plants ability to detect and 
mitigate events. Therefore, 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 involves a revision to the Technical 
Specifications to meet the intent of the guidance of NUREG-1431, and 
does not necessitate a physical alteration of the plant or change in 
parameters governing normal plant operation. The change has not 
effect on the plant's compliance with the requirements of Appendix 
J. The revision of the acceptance criteria for the air lock door 
seal test will improve the FNP [Farley Nuclear Plant] current 
testing criteria while maintaining an acceptable level of safety. 
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 a margin of safety. The revision of the acceptance criteria of 
the air lock door seal test will decrease the overall test burden 
without decreasing the margin of safety. The overall leakage rate of 
the air lock continues as less than or equal to 0.05L<INF>a and the 
plant safety analysis continues to be based ont he assumption that 
the containment leaks at its design leak rate of 0.15 percent per 
day for the first 24 hours and 0.075 percent per day thereafter for 
each of these accidents. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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

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

    Date of amendment request: July 28, 1995
    Description of amendment request: The proposed amendment would 
clarify the limiting condition for operation for TS 3.8.1.1 and 3.8.1.2 
from ``independent'' circuit to ``qualified'' circuit; explain in the 
Bases the requirements for operability of an offsite circuit; delete 
the STAGGERED TEST BASIS scheduling requirement to perform emergency 
diesel generatorsurveillances; explain in the Bases an acceptable 
method for verification of Emergency Diesel Generator speed for 
surveillance requirements (SR) 4.8.1.1.2.a.4 and 4.8.1.1.2.c.4; remove 
a surveillance test extension that has expired for SR 4.8.1.1.1.b; add 
an exception for SR 4.8.1.1.2.c.5 and 4.8.1.1.2.c.7 to SR 4.8.1.2; and 
revise Bases 3.0.5 to reflect the clarification from ``independent'' 
circuit to ``qualified'' circuit.
    Basis for proposed no significant hazards consideration 
determination: As required by 10CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Toledo Edison has reviewed the proposed changes and determined 
that a significant hazards consideration does not exist because 
operation ofthe Davis-Besse Nuclear Power Station, Unit No. 1 in 
accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because the proposed changes do not 
make a change to any accident initiator, initiating condition or 
assumption. The proposed changes do not involve a significant change 
to the plant design or operation. The proposed changes do not affect 
the safety function of the offsite circuits or the emergency diesel 
generators (EDGs).
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
invalidate assumptions used in evaluating the radiological 
consequences of 

[[Page 56371]]
an accident, do not alter the source term or containment isolation and 
do not provide a new radiation release path or alter potential 
radiological releases.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because the proposed 
changes do not introduce a new or different accident initiator or 
introduce a new or different equipment failure mode or mechanism.
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes do not reduce the margin to safety 
which exists in the present Technical Specifications [TS] or Updated 
Safety Analysis Report. The operability requirements of the TS are 
consistent with the initial condition assumptions of the safety 
analyses. Further, the proposed changes do not affect the Action 
statement requirements for the various levels of degradation in the 
offsite [power] circuits or EDGs.
    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 Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: September 29, 1995
    Description of amendment request: The proposed amendment would 
increase the minimum available borated water volume requirement for the 
boric acid addition system, the minimum and maximum boron concentration 
requirements for the borated water storage tank, the minimum boron 
concentration requirement for the core flood tanks; modify the 
surveillance requirements for trisodium phosphate dodecahydrate; and 
modify the refueling boron concentration and the associated Action 
statement. These proposed changes will affect the following Technical 
Specification sections: 3/4.1.2.8, Reactivity Control Systems - Borated 
Water Sources - Shutdown; 3/4.1.2.9, Reactivity Control Systems - 
Operating; 3/4.5.1, Emergency Core Cooling Systems (ECCS) - Core 
Flooding Tanks; 3/4.5.2, Emergency Core Cooling Systems - ECCS 
Subsystems - Tavg [plus or minus] 280  deg.F; 3/4.5.4, ECCS - Borated 
Water Storage Tank; 3/4.9.1, Refueling Operations - Boron 
Concentration; Bases 3/4.1.2, Boration Systems; Bases 3/4.5.2 and 3/
4.5.3, ECCS Subsystems; and Bases 3/4.9.1 Boron Concentration.
    Basis for proposed no significant hazards consideration 
determination: As required by 10CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Toledo Edison has reviewed the proposed changes and determined 
that a significant hazards consideration does not exist because 
operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in 
accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators, 
conditions, or assumptions are significantly affected by the 
proposed changes.
    The proposed changes to the Technical Specifications and their 
Bases increase the minimum volume of the Boric Acid Addition System 
(BAAS), the minimum boron concentration of the Borated Water Storage 
Tank (BWST) and Core Flooding Tanks (CFTs), the maximum boron 
concentration of the BWST, and the minimum volume of trisodium 
phosphate dodecahydrate (TSP) in Containment (CTMT). Administrative 
changes to these Technical Specifications have also been proposed. 
These changes ensure adequate boration capability is maintained for 
normal operations, that adequate Shutdown Margin (SDM) can be 
achieved following an accident, and that the assumed post-Loss of 
Coolant Accident (LOCA) pH can be achieved. Therefore, as stated 
above, these proposed changes do not significantly affect accident 
initiators, conditions, or assumptions.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
change the source term, CTMT isolation, or allowable releases.
    In particular, maintaining the appropriate amount of TSP will 
ensure the assumed pH will be achieved, the assumption of source 
term with respect to iodine retention will be maintained, and the 
radiological consequences of a previously evaluated accident will 
not be increased.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new 
accident initiators or assumptions are introduced by the proposed 
changes.
    These changes ensure that the assumptions used for initial and 
final conditions of SDM, pH, and source term are maintained. Also, 
the Environmental Qualification (EQ) and seismic requirements have 
been verified to be adequate to maintain the adequacy of Structures, 
Systems, and Components (SSCs) during assumed accident conditions.
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes to the minimum volume and boron 
concentration for the BAAS, BWST, and CFTs ensure that the margin of 
safety for reactor subcriticality is maintained at all times for 
future longer fuel cycles, including the upcoming Cycle 11.
    The proposed increase in the BWST maximum boron concentration is 
set at the conservative limit for post-LOCA boron precipitation 
concerns. Therefore, the existing margin of safety with respect to 
post-LOCA boron precipitation is maintained.
    The proposed increase in the minimum TSP volume requirement 
maintains the same margin of safety with respect to post-LOCA pH, 
time for dissolution, iodine retention, and chloride stress 
corrosion of austenitic stainless steels. The TSP capacity margin of 
approximately 40 cubic feet included in the minimum TSP volume 
requirement will not result in increasing the pH above the 
previously approved pH limit of 11. This reserve capacity adds 
margin to ensure adequate minimum pH is achieved.
    The proposed removal of the 1800 ppm refueling boron 
concentration requirement does not reduce the margin of safety 
because the requirement of maintaining keff [less than or equal to] 
0.95 is alone sufficient to ensure that the accident analysis 
assumptions are satisfied.
    The proposed change to the boration rate requirement of the
    LCO 3.9.1 Action statement does not reduce the margin of safety 
because the proposed boration rate of 12 gpm of 7875 ppm boric acid 
solution is equivalent to the present boration rate of
    10 gpm of 8750 ppm boric acid solution.
    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 Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: October 2, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 5.0, ``Design Features,'' 
by adding a site location description, remove site area 

[[Page 56372]]
maps, remove containment and reactor coolant system design parameters, 
remove the description of the meteorological tower location, remove 
component cyclic or transient limits, and revise the fuel assembly 
description to include the use of ZIRLO clad fuel rods.
    Basis for proposed no significant hazards consideration 
determination: As required by 10CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Toledo Edison has reviewed the proposed changes and determined 
that a significant hazards consideration does not exist because 
operation of the Davis-Besse Nuclear Power Station Unit Number 1, in 
accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators, 
conditions or assumptions are affected by the proposed changes to 
Section 5.0, Design Features, of the Technical Specifications. These 
changes are proposed to add a site location description, remove site 
area maps, remove containment and reactor coolant system design 
parameters, remove the description of the meteorological tower 
location, remove component cyclic or transient limits, and revise 
the fuel assembly description to include the use of ZIRLO clad fuel 
rods.
    Under the proposed changes, Technical Specifications (TS) 
Section 5.0 would continue to satisfy the applicable requirements of 
Section 182.a of the Atomic energy Act of 1954, and 10 CFR 
50.36(c)(4). Further, the proposed changes are consistent with 
NUREG-1430, ``Standard Technical Specifications for Babcock and 
Wilcox Plants,'' Revision 1. The information proposed for removal 
from existing TS 5.0 is presently included in the Updated Safety 
Analysis Report (USAR) or is being proposed to be added to the USAR, 
hence sufficient controls exist under 10 CFR 50.59 to ensure that 
future changes to these items are acceptable.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because no accident conditions or 
assumptions are affected by the proposed changes. As described 
above, these changes are consistent with the ``Standard Technical 
specifications for Babcock and Wilcox Plants'' (NUREG-1430) and are 
administrative changes. The proposed changes do not alter the source 
term, containment isolation, or allowable releases. The proposed 
changes, therefore, will not increase the radiological consequences 
of a previously evaluated accident.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new 
accident initiators or assumptions are introduced by the proposed 
changes, which involve only administrative controls. As described 
above, these changes are consistent with the ``Standard Technical 
Specifications for Babcock and Wilcox Plants'' (NUREG-1430) and are 
administrative changes. The proposed changes do not alter any 
accident scenarios.
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes are administrative and do not reduce or 
adversely affect the capabilities of any plant structure, systems or 
components.
    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 Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: September 6, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 5.3.1 to reflect a change in the 
maximum initial enrichment for reload fuel. The amendment would also 
change the maximum reference K<INF>infinity for storage in Region 1 of 
the spent fuel pool and TS Figure 3.9-1 to reflect a change in the 
maximum initial enrichment for storage in Region 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10CFR 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.
    An increase to a maximum initial enrichment of 5.0 w/o U-235 
does not involve an increase in the probability or consequence of an 
accident or other adverse condition over previous evaluations. 
Because of the conservative techniques and assumptions used to 
evaluate the maximum possible neutron multiplication factor, there 
is reasonable assurance that criticality safety is maintained when 
storing fuel assemblies of up to and including 5.0 w/o U-235 in the 
spent fuel storage racks under both normal and postulated accident 
conditions. For example, the calculations for non-accident 
conditions ignore the 2000 ppm soluble boron in the spent fuel pool 
calculations, thus resulting in conservative values of the 
multiplication factor. Storing fuel in the Region 1 configuration 
which meets the IFBA [integral fuel burnable absorber] versus 
enrichment curve (Figure 3 of Attachment 6) results in a maximum 
multiplication factor of 0.9481, including all biases and 
uncertainties.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    An increase to a maximum initial enrichment level of 5.0 w/o U-
235 does not create the possibility of a new or different kind of 
accident or condition over previous evaluations. An increase to the 
enrichment level of 5.0 w/o U-235 involved performing extensive 
evaluations to develop the IFBA versus enrichment curve for V-5 
fuel. Use of dual code packages ensures that the spent fuel pool 
Region 1 criticality limits are not exceeded.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    An increase in the maximum initial enrichment level to 5.0 w/o 
U-235 does not involve a reduction in the margin of safety. As 
discussed above, in all cases the multiplication factors for worst 
case assumptions fall considerably below the criticality limits and 
do not represent any reductions in margin. An increase to the 
initial enrichment level of 5.0 w/o U-235 does not adversely impact 
operation of the various plant systems, i.e. HVAC [heating, 
ventilation, and air conditioning], spent fuel pool cooling, or 
radiological control systems.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: William H. Bateman

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

    Date of amendment request: October 6, 1995
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 
4.2.b, ``Steam Generator Tubes,'' its associated bases, and Figure TS 
4.2-1 by redefining the pressure boundary for Westinghouse mechanical 
hybrid expansion joint (HEJ) steam generator (SG) tube sleeves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10CFR 50.91(a), the 

[[Page 56373]]
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:
    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist.
    1. Operation of the KNPP in accordance with the proposed license 
amendment does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    Mechanical testing has shown that the inherent structural 
strength of the HEJ joint provides sufficient integrity such that 
the tube rupture capability recommendations of RG [Regulatory Guide] 
1.121 are met, even for instances of 100 percent throughwall, 
360 deg. circumferentially oriented degradation in the HEJ HRLT 
[hardroll lower transition] region. Structural integrity 
recommendations consistent with RG 1.121 are supplied
    for all tube degradation 1.1 inch or greater below the bottom of 
the HEJ HRUT [hardroll upper transition]. Based on test data, a 
bounding SLB [steam line break] leak rate of 0.033 gpm for 
indications between 1.1 and 1.3 inch below the bottom of the HRUT is 
applied. As the leakage data base is expanded and statistical basis 
established, this SLB leakage allowance may be reduced. For 
indications existing greater than 1.3 inch below the bottom of the 
HRUT, SLB event leakage can be neglected.
    Additional prevention from tube rupture is inherently provided 
by the HEJ geometry. For RCS [reactor coolant system] release rates 
to exceed the normal makeup capacity of the plant, the tube must be 
postulated to experience a complete circumferential separation at 
the lower transition, and become axially displaced by 3 to 3.25 
inches, resulting in complete geometric disassociation between the 
tube and sleeve resulting in sufficient flow area to support leakage 
in excess of makeup capacity. During the 1989 plug top release event 
at North Anna Unit 1, primary to secondary release rates were 
calculated to be less than 80 gpm, for a flow area approximately 
four times larger than the flow area created by a tube which was 
axially displaced by about 1.25 to 1.5 inch. Analysis of the steam 
generator indicates that at a 95 percent cumulative probability, the 
tube would experience an axial displacement of less than the 1.1 
inch boundary. At this level of axial displacement, a ring of metal 
to metal contact would remain between the tube and sleeve, and 
leakage would be far less than makeup. Projected leakage at this 
point is expected to be less than 2.5 gpm. Therefore, implementation 
of the proposed repair boundary will not result in tube rupture, 
even for a tube postulated to not behave as predicted by the 
available test and pulled tube data.
    The proposed technical specification change to support the 
implementation of the HEJ sleeve tube pressure boundary for parent 
tube degradation in the HEJ HRLT region does not adversely impact 
any other previously evaluated design basis accident or the results 
of accident analyses for the current technical specification minimum 
reactor coolant system flow rate. Plugging limit criteria are 
established using the guidance of RG 1.121. Furthermore, per RG 1.83 
recommendations, the sleeved tube assembly can be monitored through 
periodic inspections with present eddy current techniques.
    2. The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Implementation of the revised pressure boundary will not 
introduce significant or adverse changes to the plant design basis. 
Mechanical testing of degraded sleeve joints supports the 
conclusions of the calculations that the sleeve retains structural 
(tube burst) capability consistent with RG 1.121. As with initial 
installation of sleeves, implementation of the relocated pressure 
boundary cannot interact with other portions of the RCS. Any 
hypothetical accident as a result of potential tube degradation in 
the HEJ HRLT region of the tube is bounded by the existing tube 
rupture accident analysis. Neither the sleeve design nor 
implementation of the tube repair boundary defined on Figure TS 4.2-
1 affects any other component or location of the tube outside of the 
immediate area repaired.
    3. The proposed license amendment does not involve a significant 
reduction in a margin of safety.
    The safety factors used in the establishment of the HEJ sleeved 
tube pressure boundary are consistent with the safety factors in the 
ASME [American Society of Mechanical Engineers] Boiler and Pressure 
Vessel Code used in steam generator design. Based on the sleeved 
tube geometry, it is unrealistic to consider that application of the 
revised pressure boundary could result in single tube leak rates 
exceeding the normal makeup capacity during normal operating 
conditions. The pressure boundary established ... has been developed 
using the methodology of RG 1.121. The performance characteristics 
of postulated degraded parent tubes of HEJ tube/sleeve joints have 
been verified by testing to retain structural integrity and preclude 
significant leakage during normal and postulated accident 
conditions. Testing indicates that postulated circumferentially 
separated tubes which the repair boundary addresses would not 
experience axial displacement during either normal operation or SLB 
conditions. The existing offsite dose evaluation performed for KNPP 
in support of the voltage based plugging criteria for axial ODSCC 
[outside diameter stress corrosion cracking] at TSP [tube support 
plate] intersections established a faulted loop primary to secondary 
leak rate of 34.0 gpm using technical specification dose equivalent 
Iodine-131 activity levels. Following implementation of the 
criteria, postulated leakage from all sources must not exceed 34.0 
gpm in the faulted loop. Maintenance of this limit will ensure that 
offsite doses would not exceed the currently accepted limit of a 
small fraction of the 10 CFR 100 guidelines. The repair boundary 
uses a conservatively established ``per indication'' leak rate for 
estimation of SLB leakage. This leak rate is applied to all 
indications left in service as a result of the tube repair boundary, 
including non-throughwall indications and a limited number of 
indications of circumferential throughwall extent.
    For a postulated indication whose performance is not 
characteristic of the test and pulled tube data, and which would 
experience axial displacement at the 95 percent cumulative 
probability value following a postulated SLB event with no operator 
intervention, leakage would not be expected to result in an 
uncontrolled release of reactor coolant in excess of normal makeup 
capacity.
    For the three removed tube sleeve samples and nearly 1,000 PTIs 
[parent tube indications] detected in the field, there were no 
instances of degradation of elevations (multiple expansion 
transitions) on either side of the hardroll expansion in the same 
tube. This includes no instances on non-detected degradation in the 
upper hydraulic and hardroll upper expansion transitions for the 
removed tubes. One tube was identified in the most recent KNPP 
inspection with two separate circumferential crack elevations within 
the HRLT. Rapidly occurring degradation would not be expected at the 
upper transitions, based partly on the field inspection results. The 
available inspection results include two inspection programs (1994 
and 1995) at Kewaunee and one at Point Beach Unit 2 (1994). Through 
these three inspection programs, approximately 11,000 HEJ sleeved 
tubes have been inspected using advanced ET [eddy current testing] 
techniques.
    The portions of the installed sleeve assembly which represent 
the reactor coolant pressure boundary can be monitored for the 
initiation and progression of sleeve/tube wall degradation, thus 
satisfying the requirements of Regulatory Guide 1.83.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497
    NRC Project Director: Gail H. Marcus

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 

[[Page 56374]]
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.

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: October 6, 1995
    Description of amendment request: Revise the Technical 
Specifications to change the definition of the F* distance.
    Date of publication of individual notice in Federal Register: 
October 16, 1995 (60 FR 53648)
    Expiration date of individual notice notice: November 15, 1995
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085

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

    Date of application for amendment: September 13, 1995, as 
supplemented by letter dated October 19, 1995
    Brief description of amendment request: The proposed amendments 
would revise Technical Specification (TS) Section 15.1, 
``Definitions,'' the basis for TS Section 15.3.1.G, ``Operational 
Limitations,'' and TS Figure 15.2.1-2, ``Reactor Core Safety Limits, 
Point Beach Unit 2.'' The proposed changes would reduce the reactor 
coolant system raw measured total flow rate limit and reflect new 
reactor core safety limits for Unit 2.Date of individual notice in 
Federal Register: October 24, 1995 (60 FR 54527)
    Expiration date of individual notice notice: November 8, 1995
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, 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.

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 6, 1995
    Brief description of amendments: The amendments extend the nominal 
surveillance interval requirements of selected safety systems 
instruments form 18 months to a refueling interval of 24 months.
    Date of issuance: October 19, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days for Unit 2 and prior to restart of the spring 1996 refueling 
outage for Unit 1.
    Amendment Nos.: 208 and 186
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35061) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated October 19, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Calvert County Library, 
Prince Frederick, Maryland 20678

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, 
Will County, Illinois Docket Nos. 50-10, 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 1, 2 and 3, Grundy County, Illinois 
Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 
2, LaSalle County, Illinois Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois Docket Nos. 50-295 and 50-304, Zion Nuclear Power Station, 
Units 1 and 2, Lake County, Illinois

    Date of application for amendments: April 24, 1995, as supplemented 
August 1 and September 14, 1995.
    Brief description of amendments: The amendments would relocate the 
requirements for the ``Review, Investigative and Audit Functions'' and 
frequencies of the quality assurance (QA) program from the 
administrative controls section of the TS to the appropriate sections 
of the licensee's Quality Assurance Topical Report (QATR), CE-1-A, 
Revision 65. In addition, the proposed TS changes include title changes 
to reflect the reorganization of the licensee's Nuclear Operations 
Division and miscellaneous administrative and editorial changes.
    Date of issuance: October 20, 1995
    Effective date: October 20, 1995
    Amendment Nos.:  75, 75, 67, 67, 38, 141, 135, 107, 93, 163, 159, 
171, and 158
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72, NPF-77, 
DPR-2, DPR-19, DPR-25, NPF-11 NPF-18, DPR-29, DPR-30, DPR-39 and DPR-
48: The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 30, 1995 (60 FR 
45175) and September 20, 1995 (60 FR 48726). The August 1 and September 
14, 1995, letters provided clarifying information that did not change 
the initial proposed no significant hazards consideration determination 
or expand the scope of the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 20, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481; for Dresden, Morris Area Public Library 
District, 604 

[[Page 56375]]
Liberty Street, Morris, Illinois 60450; for LaSalle, Jacobs Memorial 
Library, Illinois Valley Community College, Oglesby, Illinois 61348; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021; and for Zion, Waukegan Public Library, 128 N. County 
Street, Waukegan, Illinois 60085

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

    Date of application for amendments: June 30, 1995
    Brief description of amendments: The amendments modify the 
surveillance requirements for the emergency diesel generators.
    Date of issuance: October 16, 1995
    Effective date: October 16, 1995
    Amendment Nos.: 170 and 157
    Facility Operating License Nos. DPR-39 and DPR-48: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 13, 1995 (60 
FR 47615) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 16, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, 
Unit No. 2, Pope County, Arkansas

    Date of amendment request: April 4, 1995
    Brief description of amendment: The amendment deletes requirements 
associated with part length control element assemblies.
    Date of issuance: October 12, 1995
    Effective date: October 12, 1995
    Amendment No.: 169
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37090) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 12, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

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

    Date of application for amendment: March 17, 1995
    Brief description of amendment: The amendment deletes requirements 
associated with surveillance to verify position stops for High Pressure 
Safety Injection Emergency Core Cooling System throttle valves.
    Date of issuance: October 18, 1995
    Effective date: October 18, 1995
    Amendment No.: 170
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37089) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 18, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, 
Unit No. 2, Pope County, Arkansas

    Date of amendment request: April 4, 1995, as supplemented by letter 
dated October 12, 1995
    Brief description of amendment: The amendment revises the 
containment cooling response time to reduce the likelihood of a water 
hammer event in service water piping.
    Date of issuance: October 26, 1995
    Effective date: October 26, 1995
    Amendment No.: 171
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37090) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 26, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

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

    Date of application for amendments: February 28, 1994
    Brief description of amendments: The amendments delete the minimum 
frequency criteria prescribed for quality assurance audits from 
Administrative Controls sections 6.5.2.8 and 6.8.4 of the Technical 
Specifications (TS). Audit periodicity will thereby be controlled by 
the program described in the Florida Power and Light Company (FPL) 
Topical Quality Assurance Report.

    Date of issuance: October 25, 1995
    Effective date: October 25, 1995
    Amendment Nos.: 140 and 80
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17599) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 25, 1995. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, 
Turkey Point Plant Units 3 and 4, Dade County, Florida

    Date of application for amendments: July 26, 1995
    Brief description of amendments: These amendments consist of 
administrative corrections and clarifications.
    Date of issuance: October 17, 1995
    Effective date: October 17, 1995
    Amendment Nos. 177 and 171Facility Operating Licenses Nos. DPR-31 
and DPR-41: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 13, 1995 (60 
FR 47619) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 17, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, 
Turkey Point Plant Units 3 and 4, Dade County, Florida

    Date of application for amendments: July 26, 1995
    Brief description of amendments: These amendments consist of 
administrative corrections and clarifications.
    Date of issuance: October 17, 1995
    Effective date: October 17, 1995
    Amendment Nos.: 178 and 172Facility Operating Licenses Nos. DPR-31 
and DPR-41: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 13, 1995 (60 
FR 47619) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 17, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199

[[Page 56376]]


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

    Date of application for amendment: February 13, 1995, as 
supplemented April 21, 1995, and August 7, 1995.
    Brief description of amendment: The proposed amendment deletes the 
audit requirements from the Duane Arnold Energy Center Technical 
Specifications (TS) and adds them to the Quality Assurance Program.
    Date of issuance: October 17, 1995
    Effective date: October 17, 1995
    Amendment No.: 213
    Facility Operating License No. DPR-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16190) The additional information contained in the supplemental letters 
dated April 21, 1995, and August 7, 1995, was clarifying in nature and 
did not change the NRC staff's initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 17, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401

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

    Date of application for amendments: March 31, 1995
    Brief description of amendments: The amendments revise Technical 
Specification (TS) surveillance requirements for safety-related pump 
testing to eliminate recirculation alignments. In addition, specific 
test parameters, discharge pressures, and flows associated with these 
pumps are removed from the TS and will be controlled by the Inservice 
Testing Program.
    Date of issuance: October 17, 1995
    Effective date: October 17, 1995, with full implementation within 
45 days
    Amendment Nos.: 203 and 188
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 21, 1995 (60 FR 
32368) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 17, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085

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

    Date of application for amendment: February 1, 1995
    Brief description of amendment: The amendment revises Technical 
Specification 3.6.13 and associated Bases to permit the controls and 
instruments from both Remote Shutdown Panels to be considered when 
assuring that one complete set of controls and instruments is operable. 
The changes also allow 30 days to restore an inoperable function to 
operable status, remove MODE 3 (hot shutdown) from the existing 
requirement for operability, and revise the LIMITING CONDITION FOR 
OPERATION ACTION to require achieving hot shutdown in 12 hours instead 
of cold shutdown in 36 hours. An additional change permits the operator 
30 days to establish an alternate method of monitoring a parameter (and 
90 days to restore the function) when the function is inoperable.
    Date of issuance: October 16, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 155
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11135) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 16, 1995. No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126

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

    Date of application for amendment: March 29, 1995
    Brief description of amendment: The amendment modifies the current 
Technical Specifications that have cycle-specific parameter limits in 
the Core Operating Limits Report to include an additional cycle-
specific parameter and its supporting methodologies.
    Date of issuance: October 18, 1995
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 120
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24912) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 18, 1995.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

PECO Energy Company, Public Service Electric and Gas Company, 
Delmarva Power and Light Company, and Atlantic City Electric 
Company, Docket No. 50-278, Peach Bottom Atomic Power Station, Unit 
No. 3, York County, Pennsylvania

    Date of application for amendment: September 1, 1995
    Brief description of amendment: The amendment deleted License 
Condition 2.C.(5) which restricts power levels to no less than seventy 
percent in the coastdown condition.
    Date of issuance: October 17, 1995
    Effective date: As of date of issuance
    Amendment No.: 215
    Facility Operating License No. (DPR-56): This amendment revised the 
Facility Operating License. Public comments requested as to proposed no 
significant hazards consideration: Yes. (60 FR 48530). That notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by October 18, 1995, but indicated that if the Commission makes 
a final no significant hazards consideration determination any such 
hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and final no significant hazards consideration 
determination are contained in a Safety Evaluation dated October 17, 
1995.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. Vice 
President and General Counsel, PECO Energy Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 


[[Page 56377]]
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105

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

    Date of application for amendments: May 19, 1995
    Brief description of amendments: The amendments revise the 
Technical Specifications Table 3.3.3-3, ``Emergency Core Cooling System 
Response Times'' to reflect the value of 60 seconds for the High 
Pressure Coolant Injection system response time instead of 30 seconds 
as previously specified.
    Date of issuance: October 16, 1995
    Effective date: For both units, as of the date of issuance and to 
be implemented within 30 days.
    Amendment Nos.: 102 and 66
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35084) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 16, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464

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

    Date of application for amendment: July 21, 1995
    Brief description of amendment: The amendment revises TS Section 
6.0 (Administrative Controls) to replace the title-specific list of 
members on the Plant Operating Review Committee with a more general 
statement of membership requirements, and expands the scope of 
disciplines represented on the committee to include Nuclear Licensing 
and Quality Assurance. The amendment also changes the following 
management position titles: ``First Executive Vice President and Chief 
Nuclear Officer'' to ``Chief Nuclear Officer'', ``Resident Manager'' to 
``Site Executive Officer'', ``Shift Supervisor'' to ``Shift Manager'', 
and ``Assistant Shift Supervisor'' to ``Control Room Supervisor.'' 
These changes in title do not affect the reporting relationships, 
authority, or responsibilities of these positions. Finally, the 
amendment also makes editorial corrections to the TSs.
    Date of issuance: October 13, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 228
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 13, 1995 (60 
FR 47624) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 13, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126

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

    Date of application for amendment: April 12, 1995.
    Brief description of amendment: The amendment extends the 
surveillance test intervals for the nuclear steam supply system to 
support 24-month operating cycles. Surveillance test interval 
extensions that are justified will be denoted as being performed 
``every 24 months'' or ``at least once per 24 months'' consistent with 
the guidance provided in Reference 1. Other surveillances currently 
performed ``once each operating cycle,'' ``at least once during each 
operating cycle,'' ``each refueling,'' or similar notation, that are 
not being extended at this time will be denoted as being performed ``at 
least once per 18 months.'' The NRC staff has determined that the 
proposed TS changes follow the guidance of Generic Letter 91-04, and 
are therefore acceptable.
    Date of issuance: October 13, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 229
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24916) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 13, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126

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

    Date of application for amendment: April 18, 1995
    Brief description of amendment: This amendment changes Technical 
Specification Table 4.3.7.1-1, ``Radiation Monitoring Instrumentation 
Surveillance Requirements,'' to increase the channel functional test 
interval from monthly to quarterly for each instrument.
    Date of issuance: October 16, 1995
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 83
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 16, 1995 (60 FR 
42607) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 16, 1995. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

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

    Date of application for amendment: May 4, 1995
    Brief description of amendment: This amendment changes Technical 
Specification (TS) 3/4.6.1.8, ``Drywell and Suppression Chamber Purge 
System,'' increasing the annual operational limit for the drywell and 
suppression chamber purge system from 120 to 500 hours.
    Date of issuance: October 16, 1995
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 84
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 16, 1995 (60 FR 
42607) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 16, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

[[Page 56378]]


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

    Date of applications for amendment: November 30, 1994 and March 30, 
1995, as supplemented by letter dated September 5, 1995.
    Brief description of amendment: The change to TS Table 3.3.1-2, 
``Reactor Protection System Response Times,'' TS Table 3.3.2-3, 
``Isolation System Instrumentation Response Time,'' TS Table 3.3.3-3, 
``Emergency Core Cooling System Response Times,'' and associated Bases, 
eliminates the requirement to perform response time testing for certain 
classes of equipment and transfers the requirements of the above-
referenced TS Tables to the Updated Final Safety Analysis Report.
    Date of issuance: October 24,1995
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment No.: 85
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16198 and August 16, 1995 (60 FR 42606) The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
October 24, 1995. No significant hazards consideration comments 
received: No
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

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

    Date of application for amendment: May 20, 1994, as supplemented on 
March 29, 1995
    Brief description of amendment: The amendment revises Technical 
Specifications to implement the NRC's Final Policy Statement on 
Technical Specification Improvements for Nuclear Power Reactor by 
relocating specifications that do not meet policy statement criteria to 
the Final Safety Analysis Report.
    Date of issuance: October 20, 1995
    Effective date: Immediately, to be implemented within 120 days.
    Amendment No.: 103
    Facility Operating License No. NPF-30. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 
45036). The March 29, 1995, letter provided supplemental information 
that did not change the initial proposed no significant hazards 
consideration determination or expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 20, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251

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

    Date of application for amendments: January 26, 1994, as 
supplemented by letters dated December 1, 1994, and June 23, 1995
    Brief description of amendments: These amendments revise Technical 
Specification (TS) Section 15.3.0, ``General Considerations.'' This 
section specifies the actions to be taken for conditions not directly 
addressed in the action statements fo the TSs. In addition, changes to 
the applicable bases (including the bases for TS 15.3.3) and editorial 
changes are also included.
    Date of issuance: October 12, 1995
    Effective date: October 12, 1995
    Amendment Nos.: 163 and 167
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 16, 1994 (59 FR 
12373) The December 1, 1994 and June 23, 1995, submittals provided 
supplemental information that did not change the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 12, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241

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

    Date of application for amendments: April 17, 1995
    Brief description of amendments: These amendments change TS 
Sections 15.6.2, ``Organization,'' and 15.6.3, ``Facility Staff 
Qualifications.'' The requirement for the Operations Manager to hold an 
NRC Senior Reactor Operator's (SRO) license has been changed to provide 
additional staffing flexibility.
    Date of issuance: October 12, 1995
    Effective date: October 12, 1995
    Amendment Nos.: 164 and 168
    Facility Operating License Nos. DPR-24 and DPR-27. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 23, 1995 (60 FR 
27346). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 12, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241
    Dated at Rockville, Maryland, this 1st day of November 1995.
    For the Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects - III/IV,Office of Nuclear 
Reactor Regulation
[Doc. 95-27543 Filed 11-7-95; 8:45 am]
BILLING CODE 7590-01-F