[Federal Register Volume 59, Number 3 (Wednesday, January 5, 1994)]
[Notices]
[Pages 615-639]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-10105]


[[Page Unknown]]

[Federal Register: January 5, 1994]


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NUCLEAR REGULATORY COMMISSION
 

Biweekly Notice Applications and Amendments to Facility Operating 
Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from December 13, 1993, through December 22, 
1993. The last biweekly notice was published on December 22, 1993 (58 
FR 67840).

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 P-223, Phillips Building, 7920 Norfolk Avenue, 
Bethesda, 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 20555. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By February 4, 1994, 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 20555 and at the local 
public document room for the particular facility involved. If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC 20555, 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 20555, and at the local public document 
room for the particular facility involved.

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

    Date of amendment requests: December 2, 1993
    Description of amendment requests: The proposed changes would 
modify TS 3/4.6.1.2 by removing the schedular requirements for a Type A 
(overall integrated containment leakage rate) test to be performed 
specifically at 40 1B 10 month intervals and replacing these 
requirements with a requirement to perform Type A testing in accordance 
with Appendix J to 10 CFR 50.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis about the issue of no significant hazards 
consideration, which is presented below:
    (1) The proposed changes would not involve an increase in the 
probability or the consequences of an accident previously evaluated. 
The proposed change only allows flexibility in the scheduling of the 
three required Type A tests in the 10-year service period. The 
additional flexibility is needed for plants using 18-month fuel 
cycles to allow refueling outages and testing intervals to coincide. 
There is no change to the number of tests required, test 
methodology, or acceptance criteria.
    (2) The proposed changes would not create the possibility of a 
new or different type of accident from any accident previously 
evaluated. The proposed change to the test schedule only provides 
flexibility in meeting the same requirement for three tests in a 10-
year period. The testing type and bases have not changed. Therefore, 
operation of the units with this more flexible test schedule will 
not result in an accident previously not analyzed in the Updated 
Final Safety Analysis Report (UFSAR). The proposed changes do not 
impact the design bases of the containment and do not modify the 
response of the containment during a design basis accident.
    (3) The proposed changes would not involve a reduction in the 
margin of safety. The proposed changes to the schedule only provides 
flexibility in meeting the same requirement for three tests in a 10-
year period. These proposed changes do not affect or change any 
limiting conditions for operation (LCO), or any other surveillance 
requirements in the TS, and the basis for the surveillance 
requirement remains unchanged. The testing method, acceptance 
criteria, and bases are not changed. The TS continue to require 
testing that is consistent with the requirements of Appendix J to 10 
CFR 50.
    The NRC staff has reviewed the licensees' 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 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004
    Attorney for licensees: 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: Theodore R. Quay

Gulf States Utilities Company, Docket No. 50-458, River Bend 
Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: November 18, 1993
    Description of amendment request: The proposed amendment would 
permit extending the time to perform leak rate testing of certain 
containment isolation valves so that the testing can be performed 
during the refueling outage scheduled to start April 16, 1994, rather 
than requiring an earlier shutdown solely to perform the testing. The 
proposed amendment would revise Surveillance Requirements 4.6.1.3d and 
4.6.1.3f to allow a one-time extension of the surveillance intervals 
for leak rate testing of containment isolation valves. In addition, the 
proposed amendment would revise Surveillance Requirements 4.4.3.2.2a 
and 4.4.3.2.2b, replacing the requirement to leak test the reactor 
coolant pressure isolation valves every 18 months or prior to returning 
a valve to service, with a requirement to leak test the valves in 
accordance with the Inservice Testing Program. This would allow the 
testing to be performed during the fifth refueling outage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. 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:
    1. The proposed changes would not significantly increase the 
probability or consequences of a previously evaluated accident.
    One of the proposed technical specification (TS) changes requests a 
one-time only extension of the surveillance intervals for the TS 
Surveillance Requirements of TS 4.6.1.3f, leak rate testing of valves 
sealed by the main steam positive leakage control system (MS-PLCS) and 
the penetration valve leakage control system (PVLCS). The revision 
would permit eleven containment isolation valves to be tested a maximum 
of 46 days later than required by current technical specifications.
    To permit the one-time extension of the surveillance interval for 
leak rate tests of containment isolation valves, TS 4.6.1.3d must also 
be revised to permit the interval for Type C leak rate tests to exceed 
24 months. This change is consistent with an associated exemption 
request. The exemption request and this revision would permit 20 valves 
to be tested a maximum of 35 days later than required by the current 
technical specifications.
    The proposed amendment would also revise Surveillance Requirements 
4.4.3.2.2a and 4.4.3.2.2b, replacing the requirement to leak test the 
reactor coolant pressure isolation valves every 18 months or prior to 
returning a valve to service, with a requirement to leak test the 
valves in accordance with the Inservice Testing Program. This change 
would require that the pressure isolation valves be tested in 
accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 
resulting in the valves being tested at least every refueling outage, 
rather than specifying an 18 month cycle. The revision would permit 
five valves to be tested a maximum of 65 days later than allowed under 
the current technical specification.
    Based on the short duration of the requested extensions, the 
extensions will not significantly increase the probability or 
consequences of a previously evaluated accident.
    2. The proposed changes would not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed TS changes permit extension of the surveillance 
intervals for leak rate testing of containment isolation valves and 
reactor coolant system pressure isolation valves. In that the requested 
extension durations are small as compared to the overall interval 
allowed by TS, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    3. The proposed changes would not involve a significant reduction 
in the margin of safety.
    The proposed TS changes permit extension of the surveillance 
intervals for leak rate testing of containment isolation valves and 
reactor coolant system pressure isolation valves. In that the requested 
extension durations are small as compared to the overall interval 
allowed by TS, the proposed changes do not involve a significant 
reduction in the margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005
    NRC Project Director: Suzanne C. Black

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: December 8, 1993
    Description of amendments request: The frequency for Channel 
Calibration would be revised from Q (quarterly) to R (refuel) for 
Technical Specification Table 4.3.2.1, Item 4.a.4, High Pressure Core 
Injection Steam Line Tunnel Temperature-High.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The proposed change corrects Technical 
Specification pages issued for Amendment 166 for Brunswick Unit 1 
and Amendment 197 for Brunswick Unit 2, regarding NUMAC Steam Leak 
Detection Equipment. Specifically, on page 3/4 3-29 for each unit, 
the Channel Calibration frequency of Item 4.a.4, HPCI [High Pressure 
Core Injection] Steam Line Tunnel Temperature - High, was 
inadvertently left as quarterly (Q) rather than being revised to 
refuel (R). The text of CP&L's September 14, 1992 license amendment 
request and the NRC's safety evaluation for Amendments 166 and 197, 
dated October 14, 1993, addressed the frequency change from 
quarterly to refuel for this item. Therefore, the proposed change is 
purely administrative in nature and can not involve a significant 
increase in the probability of consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. As stated above, the NRC's safety evaluation for 
Amendments 166 and 197, dated October 14, 1993, addressed the 
frequency change from quarterly to refuel for Item 4.a.4 of Table 
4.3.2-1, HPCI Steam Line Tunnel Temperature - High. Therefore, the 
proposed change is purely administrative in nature and can not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed license amendment does not involve a significant 
reduction in the margin of safety. The proposed change corrects 
Technical Specification pages issued for Amendment 166 for Brunswick 
Unit 1 and Amendment 197 for Brunswick Unit 2, regarding NUMAC Steam 
Leak Detection Equipment. The NRC's safety evaluation for Amendments 
166 and 197, dated October 14, 1993, addressed the change of Channel 
Calibration frequency of Item 4.a.4, HPCI Steam Line Tunnel 
Temperature - High from quarterly (Q) to refuel (R). Therefore, the 
proposed change is purely administrative and can not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of 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: S. Singh Bajwa

Dairyland Power Cooperative, Docket No. 50-409, La Crosse Boiling 
Water Reactor (LACBWR), Vernon County, Wisconsin

    Date of application for amendment: November 5, 1993 (Reference LAC-
13320)
    Brief description of amendment: This proposed change would modify 
the Technical Specifications incorporated in Facility Operating License 
No. DPR-45 in accordance with the requirements of the revised 10 CFR 
Part 20 which becomes mandatory January 1, 1994 (56 FR 23360). In 
addition, this proposed change would correct several editorial 
oversights from previous amendments.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
the results of its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the information provided by 
the licensee and found that the licensee did not provide specific 
information as to how it determined that the three standards of 
50.92(c) were satisfied. The NRC staff performed its own evaluation of 
the proposed change to determine if the three standards of 50.92(c) 
were satisfied. The NRC staff's no significant hazards consideration 
evaluation is presented below:
    1. Will operation of the facility according to this proposed change 
involve a significant increase in the probability or consequences of an 
accident previously evaluated?
    The proposed change is to bring the LACBWR Technical Specifications 
into conformance with the revised 10 CFR Part 20 and to correct several 
editorial oversights previously evaluated. The proposed change has no 
affect on any plant operating parameters. Consequently, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Will operation of the facility according to this proposed change 
create the possibility of a new or different kind of accident from any 
previously evaluated?
    The proposed change is to bring the LACBWR Technical Specifications 
into conformance with the revised 10 CFR Part 20 and to correct several 
editorial oversights previously evaluated. The proposed change is 
administrative in nature. Further, the proposed change does not result 
in any physical alteration to any plant system, and does not result in 
any change in the method by which any safety-related system performs 
its function. Consequently, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Will operation of the facility according to this proposed change 
involve a significant reduction in a margin of safety?
    The margin of safety is the difference between the value of a 
critical design, operating, or post accident parameter, and the value 
of the parameter which would produce unacceptable results. The proposed 
change does not affect any hardware, has no effect on the current 
operating methodologies or actions which govern plant performance, and 
does not affect any accident analysis parameter. Consequently, the 
proposed change does not involve a significant reduction in a margin of 
safety.
    The NRC staff has determined based on its own no significant 
hazards consideration evaluation 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: La Crosse Public Library, 800 
Main Street, La Crosse, Wisconsin 54601
    Attorney for licensee: Fritz Schubert, Esquire, Dairyland Power 
Cooperative, 2615 East Avenue South, La Crosse, Wisconsin 54601
    NRC Branch Chief: John H. Austin

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

    Date of amendment request: November 11, 1993
    Description of amendment request: The proposed amendments would 
consolidate the Quality Verification Department with the Nuclear 
Generation Department and realign the Nuclear Safety Review Board such 
that it reports to the Senior Vice-President of the Nuclear Generation 
Department.
    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 amendments do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.]
    The proposed revisions to consolidate the Quality Verification 
Department with the Nuclear Generation Department and realign the 
NSRB [Nuclear Safety Review Board] such that it reports to the 
Senior Nuclear Officer, change the reference from Semiannual to 
Annual, change the reference from group to division, delete titles 
of persons designated to approve modifications, clarify the 
responsibilities of the Safety Assurance Manager, and delete the 
requirement to perform an annual independent Fire Protection Audit 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated because the changes 
do not have any impact upon the design or operation of any plant 
systems or components.
    [2. The amendments do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.]
    The proposed revisions will not create the possibility of a new 
or different kind of accident from any previously evaluated because 
the changes are administrative in nature and operation of Catawba, 
McGuire, and Oconee Nuclear Stations in accordance with these TS 
[technical specifications] will not create any failure modes not 
bounded by previously evaluated accidents.
    [3. The 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: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Loren R. Plisco, Acting

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

    Date of amendment request: November 11, 1993
    Description of amendment request: The proposed amendments would 
consolidate the Quality Verification Department with the Nuclear 
Generation Department and realign the Nuclear Safety Review Board such 
that it reports to the Senior Vice-President of the Nuclear Generation 
Department.
    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 amendments do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.]
    The proposed revisions to consolidate the Quality Verification 
Department with the Nuclear Generation Department and realign the 
NSRB [Nuclear Safety Review Board] such that it reports to the 
Senior Nuclear Officer, change the reference from Semiannual to 
Annual, change the reference from group to division, delete titles 
of persons designated to approve modifications, clarify the 
responsibilities of the Safety Assurance Manager, and delete the 
requirement to perform an annual independent Fire Protection Audit 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated because the changes 
do not have any impact upon the design or operation of any plant 
systems or components.
    [2. The amendments do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.]
    The proposed revisions will not create the possibility of a new 
or different kind of accident from any previously evaluated because 
the changes are administrative in nature and operation of Catawba, 
McGuire, and Oconee Nuclear Stations in accordance with these TS 
[technical specifications] will not create any failure modes not 
bounded by previously evaluated accidents.
    [3. The amendments do not involve a significant reduction in a 
margin of safety.]
    The proposed revisions will not involve a reduction in a margin 
of safety because they are administrative in nature.
    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: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Loren R. Plisco, Acting

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

    Date of amendment request: November 11, 1993
    Description of amendment request: The proposed amendments would 
consolidate the Quality Verification Department with the Nuclear 
Generation Department and realign the Nuclear Safety Review Board such 
that it reports to the Senior Vice-President of the Nuclear Generation 
Department. In addition, the requirement to conduct an annual 
independent Fire Protection Audit is deleted.
    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 amendments do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.]
    The proposed revisions to consolidate the Quality Verification 
Department with the Nuclear Generation Department and realign the 
NSRB [Nuclear Safety Review Board] such that it reports to the 
Senior Nuclear Officer, change the reference from Semiannual to 
Annual, change the reference from group to division, delete titles 
of persons designated to approve modifications, clarify the 
responsibilities of the Safety Assurance Manager, and delete the 
requirement to perform an annual independent Fire Protection Audit 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated because the changes 
do not have any impact upon the design or operation of any plant 
systems or components.
    [2. The amendments do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.]
    The proposed revisions will not create the possibility of a new 
or different kind of accident from any previously evaluated because 
the changes are administrative in nature and operation of Catawba, 
McGuire, and Oconee Nuclear Stations in accordance with these TS 
[technical specifications] will not create any failure modes not 
bounded by previously evaluated accidents.
    [3. The amendments do not involve a significant reduction in a 
margin of safety.]
    The proposed revisions will not involve a reduction in a margin 
of safety because they are administrative in nature.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036
    NRC Project Director: Loren R. Plisco, Acting

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

    Date of amendment request: November 16, 1993
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to change the periodic test 
schedule for containment Type A integrated leak rate tests (ILRTs) from 
a set of three tests performed at approximately equal intervals during 
each 10-year period, as specified in 10 CFR Part 50, Appendix J, 
Section III.D, to one Type A test performed at 10-year intervals. The 
change is being reviewed in conjunction with a proposed exemption to 
Appendix J, as requested by the licensee in a letter dated November 16, 
1993.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The Waterford 3 Type A test history provides substantial 
justification for the proposed test schedule. Three type A tests 
have been performed over an eight (8) year period with successful 
results. The tests indicate that Waterford 3 has a low leakage 
containment and that the leakage has never exceeded 24.6% of 
La. [La is the maximum allowed leakage rate of air from 
containment where containment is pressurized to Pa; for 
Waterford 3 Pa is 44 psig. La for Waterford is 0.50 
percent by weight of the containment air per 24 hours at Pa.]
    There are no structural mechanisms which would adversely affect 
the structural capability of the containment and that would be a 
factor in extending the Type A test schedule to ten years. A risk 
impact assessment was performed, and a determination was made that 
there is no risk impact as a result of changing the Type A test 
schedule. Therefore, the proposed change will not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.
    There are no design changes being made that would create a new 
type of accident or malfunction. The proposed change will not alter 
the plant or the manner in which it is operated. The change proposes 
a change to the schedule for performing the periodic Type A test. 
The purpose of the test is to provide periodic verification by test 
of the leaktight integrity of the primary reactor containment, and 
systems and components which penetrate containment. The tests assure 
that leakage through containment and systems and components 
penetrating containment will not exceed the allowable leakage rate 
values associated with conditions resulting from an accident. The 
change in schedule for performing the Type A test will not adversely 
affect the containment integrity in the event of an accident. 
Therefore, the proposed change will not create the possibility of a 
new or different type of accident from any accident previously 
evaluated.
    The proposed change is a change to the schedule for performing 
the periodic Type A tests and does not reduce the margin of safety 
assumed in accident analysis for release of radioactive materials 
from the containment atmosphere into the environment or any margin 
of safety preserved by the Technical Specifications. The 
methodology, acceptance criteria, and the technical specification 
leakage limits for the performance of the Type A tests will not 
change, and the Type A tests will be performed in accordance with 
10CFR 50, Appendix J, and the Waterford 3 licensing basis. 
Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: November 16, 1993
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to provide acceptable 
conditions for operation when (1) the core operating limits supervisory 
system (COLSS) is in service and neither control element assembly 
calculator (CEAC) is operable and (2) the COLSS is out of service and 
either or both CEACs are operable.
    This proposed TS change modifies the departure from nucleate 
boiling ratio (DNBR) margin, Limiting Condition for Operation (LCO) 
3.2.4b and c, which limits the core power distribution to the initial 
value assumed in the accident analyses. Operation within this LCO 
either limits or prevents potential fuel cladding failures in the event 
of a postulated accident and limits damage to the fuel cladding during 
an accident by ensuring that the plant is operating within acceptable 
conditions at the onset of a transient. The limiting safety system 
settings and this LCO are based on the accident analysis, so that 
specified acceptable fuel design limits (SAFDLs) are not exceeded as a 
result of anticipated operational occurrences (AOOs) and the limits of 
acceptable consequences are not exceeded for other postulated 
accidents.
    The COLSS and core protection calculators (CPCs) monitor the core 
power distribution on line and are capable of verifying that the linear 
heat rate (LHR) and DNBR do not exceed their limits. The COLSS performs 
this function by continuously monitoring the core power distribution 
and calculating core power operating limits corresponding to the 
allowable peak LHR and DNBR. The CPCs perform this function by 
continuously calculating an actual value of DNBR and LHR for comparison 
with the respective trip setpoints. CEACs monitor CEA position. Should 
a CEA deviate from its subgroup position, the CEACs will transmit an 
appropriate ``penalty'' factor to the CPCs.
    The COLSS is normally used to monitor DNBR margin. When at least 
one CEAC is operable, TS 3.2.4a provides enough margin to DNB to 
accommodate the limiting AOO without failing the fuel. When neither 
CEAC is operable, the CPCs lack the CEA position information necessary 
to ensure a reactor trip when necessary. In this case TS 3.2.4b 
requires the COLSS calculated core power to be reduced to ensure that 
the limiting AOO will not result in fuel failure. Currently, TS 3.2.4b 
requires that the COLSS calculated power be maintained at 13% below the 
COLSS calculated power operating limit to compensate for the potential 
error in the CPC DNBR calculation. The proposed revision would increase 
this required adjustment to 16%, which is more restrictive than the 
present value.
    In instances when the COLSS is out of service, but either or both 
CEACs are operable, TS 3.2.4c states that the DNBR operating margin 
shall be maintained by comparing the DNBR indicated on any operable CPC 
channel with the allowable value from TS Figure 3.2-2. Whenever the 
COLSS is out of service, the CPCs are used to perform the same 
monitoring function. However, the extra conservatisms built into the 
CPCs for transient protection are not all required when the CPCs are 
being used for monitoring. In order not to affect the CPC transient 
protection, these conservatisms are not taken from the CPC, but are 
credited in the COLSS out-of-service limits in Figure 3.2-2. A 
reevaluation of the limiting AOOs has verified that, by maintaining the 
margin in the proposed Figure 3.2-2, sufficient margin exists to ensure 
that the limiting Cycle 7 AOO will not result in fuel failure.
    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:
    For the case when neither CEAC is operable but COLSS is in 
service, the CPCs assume a preset CEA configuration and can not 
obtain the required CEA position information to ensure the SAFDL on 
DNBR will not be violated during an AOO. Thus, as a result of 
limiting AOOs for Cycle 7, Specification 3.2.4b requires that core 
power be reduced to a value 16% less than the current COLSS 
calculated power operating limit. This ensures the limiting AOO will 
not result in a violation of SAFDLs. The proposed revision to Figure 
3.2-2 accounts for the situation when COLSS is out of service but at 
least one CEAC is operable. In this case, the Cycle 7 safety 
analysis has shown that by maintaining the CPC calculated DNBR above 
the value shown in the figure, the limiting AOO will not result in a 
violation of the SAFDLs. Therefore, the proposed change will not 
significantly increase the probability or consequences of any 
accident previously evaluated.
    The proposed changes are primarily a result of changes in Cycle 
7 core parameters. These changes do not involve any change to any 
equipment or manner in which the plant will be operated. These 
changes further restrict the plant operation when either COLSS or 
both CEACs are out of service. Therefore, the proposed change will 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The intent of this Specification is to ensure that there is 
always sufficient margin to DNB such that the CPCs can mitigate the 
consequences of the most limiting AOO prior to a violation of the 
SAFDLs. Generally, this margin is continuously monitored by COLSS; 
however, if COLSS is out of service, but at least one CEAC is 
operable, the limitation on CPC calculated DNBR (as a function of 
ASI [axial shape index]) shown in Figure 3.2-2 represents a 
conservative envelope of operating conditions consistent with the 
Cycle 7 safety analysis assumptions. This band of operating 
conditions has been analytically demonstrated to maintain an 
acceptable minimum DNBR through all AOOs. On the other hand, for the 
case when COLSS is in service, but neither CEAC is operable, the 
proposed change will ensure that the limiting AOO will not result in 
a violation of SAFDLs. Therefore, the proposed changes 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: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: December 2, 1993
    Description of amendment request: The purpose of the request is to 
change the plant Technical Specifications (TS) to remove the limiting 
conditions for operation and surveillance requirements for the chlorine 
detection system. TMI-1 removed the gases Chlorination System for the 
Circulating Water and River Water Systems. This modification eliminated 
the need for a Chlorine Dectection System (CDS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated. The TS requirements assured the operability of 
the CDS in the event of an on-site chlorine release from a one ton 
cylinder. These TS requirements reduced the probability and the 
consequences of a radiological accident which may result from an 
incapacitation of control room operatorsafter entry of chlorine into 
the control room. With the removal and the restriction on delivery 
of one ton chlorine cylinders, this postulated event is no longer 
credible, and there is a decrease in the probability of a 
radiological accident.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated. The TS 
requirements associated with the CDS were for the on-site release of 
chlorine from a one ton cylinder. These cylinders are removed and 
prohibited from the TMI-1 site. These actions preclude a significant 
on-site release of chlorine which could affect the control room 
operators.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. The purpose of the TS requirements was to maintain 
operability of the CDS in the event of on-site release from a one 
ton chlorine cylinder. Since chlorine cylinders greater than 150 
pounds are prohibited on-site, the TS requirements for chlorine 
detection are no longer required, and their removal will not reduce 
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 Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: November 23, 1993
    Description of amendment request: The licensee proposes to modify 
the South Texas Project, Units 1 and 2, Technical Specification 3/
4.8.1.1, ``A.C. Sources,'' to modify the action statements and 
surveillance requirements for testing of the standby diesel generator. 
This amendment would incorporate the recommendations of NRC Generic 
Letter (GL) 93-05, ``Line-Item Technical Specifications Improvements To 
Reduce Surveillance Requirements For Testing During Power Operation,'' 
dated September 27, 1993.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change seeks to eliminate the unnecessary testing 
of an operable Standby Diesel Generator (SDG). Technical 
Specification (TS) 3.8.1.1 Actions a. and e. require all operable 
SDGs be started as a demonstration of operability whenever one or 
more of the offsite AC [alternating current] power sources is 
declared inoperable. The inoperability of an offsite AC power source 
has no effect on the reliability of a SDG. Deleting this requirement 
does not affect the design or performance characteristics of the 
SDGs. Therefore, the SDGs maintain their ability to perform their 
design function.
    TS 3.8.1.1 Actions b. and c. require all remaining operable SDGs 
be started as a demonstration of operability whenever one of the SDG 
is declared inoperable except for preplanned preventive maintenance 
or testing. The proposed amendment would expand the testing 
exclusion to include an inoperable support system and an 
independently testable component in addition to preplanned 
preventive maintenance and testing. The proposed amendment would 
also eliminate the testing requirement of the remaining operable 
SDGs, when a SDG is declared inoperable, unless there is cause to 
believe a potential common mode failure exists for the remaining 
SDGs. The normal TS surveillance testing schedule assures that 
operable SDG(s) are capable of performing their intended safety 
functions. A failure of one SDG does not reduce the reliability of 
another, otherwise operable SDG. Deleting this requirement does not 
affect the design or performance characteristics of the SDGs, once a 
common mode failure has been dismissed. Therefore, the SDGs maintain 
their ability to perform their design function.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The elimination of these unnecessary tests does not affect the 
design bases of the SDGs, or any of the accident evaluations 
involving the SDGs. The SDGs are designed to provide electrical 
power to the equipment important for safety during all modes and 
plant conditions following a loss of offsite power. The test 
schedule established in accordance with GL 84-15 [``Proposed Staff 
Actions To Improve and Maintain Diesel Generator Reliability''] 
assures that operable SDGs are capable of performing their intended 
safety function. Therefore, this change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    Since the proposed change does not affect the design bases, 
accident analysis, reliability or capability of the SDGs to perform 
their intended safety function, this change does not involve any 
reduction in a margin to safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton Texas 77488
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, NW, Washington, DC 20036
    NRC Project Director: Suzanne C. Black

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit No. 1, Berrien County, Michigan

    Date of amendment request: December 15, 1993. This submittal 
supersedes a previous submittal dated March 10, 1993.
    Description of amendment request: The proposed amendment would 
implement interim tube plugging criteria for the tube support plate 
elevation outer diameter stress corrosion cracking for cycle 14.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Operation of Donald C. Cook Nuclear Plant Unit 1 in 
accordance with the proposed license amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Testing of model boiler specimens for free span tubing (no TSP 
restraint) at room temperature conditions shows burst pressures in 
excess of 5000 psi for indications of ODSCC with voltage 
measurements as high as 19 volts. Burst testing performed on pulled 
tubes from Cook Nuclear Plant Unit 1 with up to a 2.02 volt 
indication shows measured burst in excess of 10,000 psi at room 
temperature. Correcting for the effects of temperature on material 
properties and minimum strength levels (as the burst testing was 
done at room temperature), tube burst capability significantly 
exceeds the RG 1.121 criterion requiring the maintenance of a margin 
of 3 times normal operating pressure differential on tube burst. The 
3 times normal operating pressure differential for the Cook Nuclear 
Plant Unit 1 steam generators corresponds to 4275 psi. Based on the 
existing data base, this criterion is satisfied with 7/8'' diameter 
tubing with bobbin coil indications with signal amplitudes less than 
4.9 volts, regardless of the indicated depth measurement. A 1.0 volt 
plugging criteria compares favorably with the structural limit 
considering the previously calculated growth rates for ODSCC within 
the Cook Nuclear Plant Unit 1 steam generators. Considering a 
voltage increase of 0.4 volts, and adding a 20% NDE uncertainty of 
0.20 volts (90% Cumulative Probability) to the IPC of 1.0 volts 
results in an EOC voltage of approximately 1.6 volts for Cycle 14 
operation. A 3.3 volt safety margins implied (4.9 structural limit - 
1.6 volt EOC - 3.3 volt margin). This EOC voltage compares favorably 
with the Structural Limit of 4.9 volts.
    For the voltage/burst correlation, the EOC structural limit is 
supported by a voltage of 4.9 volts. A 3.1 volt BOC repair limit 
confirms the structural limit when 40% growth and 20% uncertainty 
are applied to the repair limit. This repair limit will be applied 
for Cycle 14 IPC implementation to repair bobbin indication greater 
than 3.1 volts independent of RPC confirmation of the indication.
    The conservatism of this repair limit is shown by the EOC 12 
(Summer 1992) eddy current data. The overall average voltage growth 
was determined to be only 2.2%, with a 12% average voltage growth 
for indications less than 0.75 volt BOC and a 1% average voltage 
growth for indication >0.75 volt at the BOC. In addition, the Cycle 
12 maximum observed voltage increase was found to be 0.49 volts, and 
occurred in a tube initially <1.0 BOC. In accordance with the 
technical specification requirements, the applicability of Cycle 13 
growth rates for Cycle 14 operation will be confirmed prior to 
return to power of Cook Nuclear Plant Unit 1. Similar large 
structural margins are anticipated.
    As stated previously, TSP proximity to the tubes will prevent 
tube burst during all plant conditions. Test data indicates that 
tube burst cannot occur within the TSP, even for tubes which have 
100% through-wall EDM notches, 0.75 inch long, provided that the TSP 
is adjacent to the notched area. Therefore, a more realistic 
assessment of tube operability should be performed against the RG 
1.121 loading requirements during accidents SLB conditions, since 
the TSP has the potential to deflect during blowdown following a 
main SLB, thereby uncovering the intersection. At the ASME Code 
recommended faulted condition loading of 3657 psi (2560 psi/0.7) 
structural integrity is provided for bobbin voltage indications of a 
minimum of 9.6 volts. The repair limit based on the structural 
limited conservative SLB conditions would be 6.0 volts (compared to 
a 3.1 volt repair limit for a structural limit based on the 
3[delta]P burst capability voltage).
    Only three indications of ODSCC have been reported to have 
operating leakage, and all three have been in European plants. No 
field leakage has been reported at other plants from tubes with 
indications of a voltage level of under 7.7 volts (from 3/4'' 
tubing). Relative to the expected leakage during accident condition 
loadings, it has been previously established that a postulated main 
SLB outside of containment but upstream of the MSIV represents the 
most limiting radiological condition relative to the IPC. In support 
of implementation of the IPC, it will be determined whether the 
distribution of cracking indications at the TSP intersections at the 
EOC 14 are projected to be such that primary to secondary leakage 
would result in site boundary doses within a small fraction of the 
10 CFR 100 guidelines. The SLB leakage rate calculation methodology 
prescribed in Section 3.3 of draft NUREG-1477 will be used to 
calculate EOC 14 leakage. Due to the relatively low voltage growth 
rates at Cook Nuclear Plant Unit 1 and the relatively small number 
of indications affected by the IPC, SLB leakage prediction per draft 
NUREG-1477 is expected to be less than the acceptance limit of 1.0 
gpm in the faulted loop and far below the conservatively calculated 
SRP based allowable value of 120 gpm in the faulted loop. The NRC 
leakage rate calculation methodology applies a 98% confidence limit 
on leakage that is independent of voltage. This method for 
calculating SLB leakage is conservative as it assumes no 
correlations exists between SLB leakage and bobbin probe voltage. 
Tube pull results from Cook Nuclear Plant Unit 1 indicate that tube 
wall degradation of greater than 40% through-wall was detectable 
either by the bobbin or RPC probe. The tube with maximum through-
wall penetration of 56% (42% average) had a voltage of 2.02 volts. 
This indication also was the largest recorded bobbin voltage from 
the EOC 12 eddy current data. All burst tested tube intersections 
had degradation depths of 40% to 56 % (actual) deep and all were 
detected by both probes, with all bobbin voltage grater than or 
equal to 1.0. Since the criteria requires the plugging of >1.0 volt 
bobbin indications with confirmed RPC calls, using the Cook Nuclear 
Plant Unit 1 pulled tube destructive examination results, it is 
reasonable that no indications of degradation greater than 40% to 
56% deep with an ability to influence tube burst capability were 
left in service. Since the majority of the EOC 14 indications at 
Cook Nuclear Unit 1 are expected to be below this level, the 
inclusion of all IPC intersections into the leakage calculation is 
exceptionally conservative.
    Therefore, as re-implementation of the 1.0 volt IPC during Cycle 
14 does not adversely affect steam generator tube integrity and 
results in acceptable dose consequences, the proposed amendment does 
not result in any increase in the probability or consequences of an 
accident previously evaluated within the Cook Nuclear Plant Unit 1 
FSAR.
    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 proposed steam generator tube IPC does not 
introduce any significant changes to the plant design basis. Use of 
the criteria does not provide a mechanism which could result in a 
tube rupture outside of the region of the TSP elevations; no ODSCC 
is occurring outside the thickness of the TSPs. Neither a single or 
multiple tube rupture event would be expected in a steam generator 
in which the plugging criteria has been applied (during all plant 
conditions).
    Specifically, Cook Nuclear Plant will continue to implement a 
maximum leakage rate limit of 150 gpd (0.1 gpm) per steam generator 
to help preclude the potential for excessive leakage during all 
plant conditions. The Cycle 14 Technical Specification limits on 
primary to secondary leakage at operating conditions is a maximum of 
0.4 gpm (600 gpd) for all steam generators, or, a maximum of 150 gpd 
for any one steam generator. The RG 1.121 criterion for establishing 
operational leakage rate limits that require plant shutdown are 
based upon leaks-before-break consideration to detect a free span 
crack before potential tube rupture. The 150 gpd limit should 
provide for leakage detection and plant shutdown in the event of the 
occurrence of an unexpected single crack resulting in leakage that 
is associated with the longest permissible crack length. RG 1.121 
acceptance criteria for establishing operating leakage limits are 
based on leak-before-break considerations such that plant shutdown 
is initiated if the leakage associated with the longest permissible 
crack is exceeded. The longest permissible crack is the length that 
provides a safety factor of 3 against bursting at normal operating 
pressure differential. A voltage amplitude of 4.9 volts for typical 
ODSCC corresponds to meeting this tube burst requirement at a lower 
95% prediction limit on the burst correlation coupled with 95/95 LTL 
material properties. Alternate crack morphologies can correspond to 
4.9 volts so that a unique crack length is not defied by the burst 
pressure versus voltage correlation. Consequently, typical burst 
pressure versus through-wall crack length correlations are used 
below to define the ``longest permissible crack'' for evaluating 
operating leakage limits.
    At current plant conditions, the single through-wall crack 
lengths that result in tube burst at 3 times normal operating 
pressure differential and SLB conditions are 0.44 inch and 0.84 
inch, respectively. A leak rate of 150 gpd will provide for 
detection of 0.42 inch long cracks at nominal leak rates and 0.61 
inch long cracks at the lower 95% confidence level leak rates. Since 
tube burst is precluded during normal operation due to the proximity 
of the TSP to the tube and the potential for the crevice to become 
uncovered during SLB conditions, the leakage from the maximum 
permissible crack must preclude tube burst at SLB conditions. Thus, 
the 150 gpd limit provides for plant shutdown prior to reaching 
critical crack lengths for SLB conditions.
    3. The proposed license amendment does not involve a significant 
reduction in margin of safety.
    The use of the voltage based bobbin probe interim TSP elevation 
plugging criteria at Cook Nuclear Plant Unit 1 is demonstrated to 
maintain steam generator tube integrity commensurate with the 
criteria of Regulatory Guide 1.121. RG 1.121 describes a method 
acceptable to the NRC staff for meeting GDCs 14, 15, 31, and 32 by 
reducing the probability or the consequences of steam generator tube 
rupture. This is accomplished by determining the limiting conditions 
of degradation of steam generator tubing, as established by 
inservice inspection, for which tubes with unacceptable cracking 
should be removed from service. Upon implementation of the criteria, 
even under the worst case conditions, the occurrence of ODSCC at the 
TSP elevations is not expected to lead to a steam generator tube 
rupture event during normal or faulted plant conditions. The EOC 14 
distribution of crack indications at the TSP elevations will be 
confirmed to result in acceptable primary to secondary leakage 
during all plant conditions and that radiological consequences are 
not adversely impacted.
    In addressing the combined effects of LOCA + SSE on the steam 
generator component (as required by GDC 2), it has been determined 
that tube collapse may occur in the steam generators at some plants. 
This is the case as the TSPs may become deformed as a result of 
lateral loads at the wedge supports at the periphery of the plant 
due to the combined effects of the LOCA rarefaction wave and SSE 
loadings. Then, the resulting pressure differential on the deformed 
tubes may cause some of the tubes to collapse.
    There are two issues associated with steam generator tube 
collapse. First, the collapse of steam generator tubing reduces the 
RCS flow area through the tubes. The reduction in flow are increases 
the resistance to flow of steam from the core during a LOCA which, 
in turn, may potentially increase peak clad temperature (PCT). 
Second, there is a potential that partial through-wall cracks in 
tubes could progress to through-wall cracks during tube deformation 
or collapse.
    Consequently, since the leak-before-break methodology is 
applicable to the Cook Nuclear Plant Unit 1 reactor coolant loop 
piping, the probability of breaks in the primary loop piping is 
sufficiently low that they need not be considered in the structural 
design of the plant. The limiting LOCA event becomes either the 
accumulator line brake or the pressurizer surge line break. LOCA 
loads for the primary pipe breaks were used to bound the Cook 
Nuclear Plant Unit 1 smaller breaks. The results of the analysis 
using the larger break inputs show that the LOCA loads were found to 
be of insufficient magnitude to result in steam generator tube 
collapse or significant deformation.
    Addressing RG 1.83 consideration, implementation of the bobbin 
probe voltage based interim tube plugging criteria of 1.0 volt is 
supplemented by the following: enhanced eddy current inspection 
guidelines to proved consistency in voltage normalization, a 100% 
eddy current inspection sample size at the TSP elevations, and RPC 
inspection requirements as outlined in the technical specifications 
and Appendix A ``NDE Data Acquisition and Analysis Guidelines'' 
(Attachment 6).
    As noted previously, implementation of the TSP elevation 
plugging criteria will decrease the number of tubes which must be 
repaired. The installation of steam generator tube plugs reduce the 
RCS flow margin. Thus, implementation of the alternate plugging 
criteria will maintain the margin of flow that would otherwise be 
reduced in the event of increased tube plugging.
    Based on the above, it is concluded that the proposed license 
amendment request does not result in a significant reduction in 
margin with respect to plant safety as defined in the Final Safety 
Analysis Report or any of the plant Technical Specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Leaks Preston Palenske 
Memorial Library, 500 Market Street, St. Joseph, Michigan 49085
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: A Randolph Blough, Acting

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

    Date of amendment requests: November 15, 1993
    Description of amendment requests: The proposed amendments delete 
certain Limiting Conditions for Operation, Actions, and Surveillance 
Requirements for Reactor Coolant System Pressure Isolation Valves in 
the Technical Specifications. The Technical Specifications for these 
Reactor Coolant System Pressure Isolation Valves were added by Order 
dated April 20, 1981. This Order was prompted by concerns for an 
interfacing system loss-of-coolant accident as identified in the 
Reactor Safety Study (WASH-1400). The proposed Technical Specification 
change, by inference, also requests rescission of the April 20, 1981 
Order.
    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:
    Per 10 CFR 50.92, a proposed amendment to an operating license 
will not involve a significant hazards consideration if the proposed 
amendment satisfies the following three criteria:
    1. Does not involve a significant increase in the probability or 
consequences of an accident previously analyzed,
    2. Does not create the possibility of a new or different kind of 
accident from an accident previously analyzed or evaluated, or
    3. Does not involve a significant reduction in a margin of 
safety.
    Criterion 1
    The ISLOCA is not one of the accidents previously analyzed in 
Chapter 14, Safety Analysis, of the Cook Nuclear Plant Updated Final 
Safety Analysis Report. Chapter 14 analyzes the large break LOCA in 
Section 14.3.1, and ``loss of reactor coolant from small ruptured 
pipes or from cracks in large pipes which actuates the ECCS'', or 
small break LOCA in Section 14.3.2. Therefore, deleting from the 
Technical Specifications the Reactor Coolant System pressure 
isolation valves in Table 3.4-0, will not increase the probability 
or the consequences of the large break or the small break LOCAs 
previously analyzed for the Cook Nuclear Plant.
    Criterion 2
    The Reactor Coolant System pressure isolation valves in Table 
3.4-0 of the Technical Specifications were added because WASH-1400 
identified the ISLOCA as a significant contributor to core damage 
frequency. Deletion of the subject valves from the Technical 
Specifications and reliance on the testing requirements mandated by 
the In-Service Testing Program of ASME XI does not create the 
possibility of a new or different kind of accident from the large 
break or the small break LOCAs previously analyzed for the Cook 
Nuclear Plant.
    Criterion 3
    Deleting the Reactor Coolant System pressure isolation valves 
from the testing requirements in Table 3.4-0 of the Technical 
Specifications will result in these valves only being tested on a 
refueling outage frequency as part of the ASME B&PV Code Section XI 
IST Program. This somewhat reduced testing frequency will result in 
a slight increase in the ISLOCA contribution to core damage 
frequency of 5.4%, from lower 5.00E-08/reactor year to mid 5.00E-08/
reactor year. This insignificant increase will not affect the 
overall core damage frequency of 6.26E-05/reactor year. Therefore, 
it is concluded that the proposed deletion of the Reactor Coolant 
System pressure isolation valves in Table 3.4-0 of the Technical 
Specifications, as well as the proposed deletion of the portions of 
the Technical Specifications that are affected by Table 3.4-0, will 
not result in a significant reduction in the margin of safety that 
exists at Cook Nuclear Plant to prevent an ISLOCA or to mitigate the 
consequences of an ISLOCA.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: A. Randolph Blough, Acting

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay 
Power Plant, Unit 3, Humboldt County, California

    Date of application for amendment: October 8, 1993 (Reference LAR 
93-02)
    Brief description of amendment: This Licensee Amendment Request 
(LAR) proposes to revise the Humboldt Bay Power Plant (HBPP), Unit 3, 
Technical Specifications (TS) by deleting Figure II-2 in Section II, 
``Site,'' by deleting the Restricted Area boundary line in Figure V-3, 
Section V, ``Monitoring Systems,'' by incorporating a title change into 
Section VII, ``Administrative Controls,'' and by revising Figure VII-2, 
``Plant Staff Organization.'' The proposed changes are in response to 
the revised 10 CFR Part 20 which becomes mandatory on January 1, 1994 
(56 FR 23360). The specific TS changes proposed are as follows:
    (1) Page v, Figures - delete reference to Figure II-2.
    (2) Page II-1, Section II.B, Plant Areas - change ``is shown in 
Figure II-2'' to ``shall be defined in plant procedures.''
    (3) Page II-3, Section II - delete Figure II-2.
    (4) Page V-14, Section V - delete the Restricted Area boundary line 
from Figure V-3, ``HBPP Groundwater Monitoring Systems Wells,'' to be 
consistent with item 3 above.
    (5) Page VII-5, Section VII.C.2.e, Supervisor of Maintenance - 
change the title from ``Supervisor of Maintenance'' to ``Maintenance 
Planner.''
    (6) Page VII-10, Section VII.D.1.b., Membership, List of minimum 
membership - replace ``Supervisor of Maintenance'' with ``Maintenance 
Planner.''
    (7) Page VII-31, Section VII, Figure VII-2, Plant Staff 
Organization - replace ``Maintenance Supervisor'' with ``Maintenance 
Planner.''
    (8) Page VII-31, Section VII, Figure VII-2, Plant Staff 
Organization - both the Mechanical Foreman and the Instrument/
Electrical Foreman report directly to the Plant Manager, not to the 
``Maintenance Planner,'' as previously shown.
    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:
    a. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    A change to the defined restricted area has no affect on any 
plant operating parameters. Consequently, a change to the defined 
restricted area will not affect the probability or consequences of 
an accident occurring.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    b. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed revisions to the HBPP TS are administrative in 
nature. Further, the proposed changes would not result in any 
physical alteration to any plant system, and there would not be a 
change in the method by which any safety-related system performs its 
function.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    c. Does the change involve a significant reduction in a margin 
of safety?
    The proposed revisions to the HBPP TS do not affect the margin 
of safety of any accident analysis since they do not affect the 
parameters for any accident analysis, and have no effect on the 
current operating methodologies or actions which govern plant 
performance.
    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 requests involve no significant hazards consideration.
    Local Public Document Room location: Humboldt County Library, 636 F 
Street, Eureka, California 95501
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas & 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Branch Chief: John H. Austin

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

    Date of amendment request: October 29, 1993
    Description of amendment request: The amendment would revise the 
Limerick Generating Station, Units 1 and 2, Technical Specifications to 
eliminate the Main Steam Line Radiation Monitoring System high 
radiation trip function for initiating 1) an automatic reactor scram 
and automatic closure of the Main Steam Line Isolation Valves, and 2) 
automatic closure of the Main Steam Line drain valves, and Main Steam 
and Reactor Water Sample line valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed Technical Specification (TS) changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed TS changes involve eliminating the Main Steam Line 
Radiation Monitoring (MSLRM) system high radiation trip function for 
initiating an automatic reactor scram and automatic closure of the 
Main Steam Line Isolation Valves (MSIVs), Main Steam line drain 
valves, and Main Steam and Reactor Water sample line valves. The 
proposed TS changes support installation of a plant modification to 
defeat portions of MSLRM system high radiation trip function logic 
circuitry in the Reactor Protection System (RPS) and Primary 
Containment and Reactor Vessel Isolation Control System (PCRVICS). 
Installation of this modification will not adversely impact the 
operation of the RPS or PCRVICS with respect to performing its other 
intended safety functions. The proposed TS changes will not affect 
the operation of other plant systems or equipment important to 
safety. The MSLRM system high radiation trip function for the 
Mechanical Vacuum Pump (MVP) will be retained. The safety assessment 
and justification for eliminating the MSLRM system high radiation 
trip function for initiating an automatic reactor scram and 
automatic closure of the MSIVs [are] based on General Electric's 
(GE's) Topical Report NEDO-31400A, ``Safety Evaluation for 
Eliminating the Boiling Water Reactor Main Steam Line Isolation 
Valve Closure Function and Scram Function of the Main Steam Line 
Radiation Monitor,'' and the applicability of this report to 
Limerick Generating Station (LGS), Units 1 and 2. By letter dated 
May 15, 1991, the NRC approved this topical report and indicated 
that it was acceptable for licensees to reference this report as the 
basis for requesting a TS change to eliminate the MSLRM system high 
radiation trip functions as documented in the report and associated 
NRC Safety Evaluation Report (SER).
    The safety assessment provided in NEDO-31400A can also be 
applied to eliminate the MSLRM system high radiation trip function 
for initiating the automatic closure of the Main Steam line drain 
valves although this aspect was not explicitly evaluated in NEDO-
31400A. The flow from these valves ultimately discharges to the main 
condenser as do the MSIVs and therefore, any radioactive material 
passing through these valves would be processed in the same fashion 
as that passing through the MSIVs. The effects of eliminating the 
MSLRM system high radiation trip function for initiating the closure 
of the Main Steam and Reactor Water sample line valves is [are] 
negligible. The sample lines are routed to a sample sink where inlet 
valves installed on the sample lines are normally closed. 
Additionally, downstream of the inlet valves are needle valves 
designed to control and limit sample line flow. The sample sink is 
enclosed, and air vented from its exhaust hood is passed through 
filters prior to release to the environment. There is the potential 
that a minimal amount of radioactive material could be released to 
the environment if the sample sink inlet and needle valves failed to 
properly function. This potential release has been evaluated and 
determined to a small fraction of the dose limit requirements 
specified in 10 CFR 100.
    The MSLRM system high radiation trip was intended to function in 
response to a Control Rod Drop Accident (CRDA), a Design Basis 
Accident previously evaluated. Although the CRDA assumes MSIV 
closure, no credit was taken for this in the CRDA analysis since it 
postulates that the radioactive material calculated to be released 
from the fuel is transported to the main condenser prior to the 
MSIVs completely closing. Furthermore, the probability of a fuel 
failure is independent of the operation of the MSLRM system.
    The Steam Jet Air Ejectors (SJAEs) will continue to operate to 
remove non-condensable gases from the main condenser for processing 
by the Offgas Treatment system. The Offgas Treatment system will 
continue to function as designed to reduce offgas radioactivity 
levels prior to release to the environment. Eliminating the MSLRM 
system high radiation isolation functions will improve operational 
flexibility in that the main condenser will be available to aid in 
decay heat removal. Elimination of the MSLRM system high radiation 
trip functions in conjunction with proper operation of the Offgas 
Treatment system will ensure that any radioactive material released 
to the environment is a small fraction of 10 CFR 100 limits.
    Therefore, the proposed TS changes associated with eliminating 
the MSLRM system high radiation trip function for initiating an 
automatic reactor scram and automatic closure of the MSIVs, Main 
Steam line drain valves, and Main Steam and Reactor Water sample 
line valves do not involve an increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS changes involve eliminating the MSLRM system 
high radiation trip function for initiating an automatic reactor 
scram and automatic closure of the MSIVs, Main Steam Line drain 
valves, and Main Steam and Reactor Water sample line valves. The 
proposed TS changes will not affect the operation of other plant 
systems or equipment important to safety. The associated plant 
modification simply defeats the MSLRM system high radiation trip 
function logic circuitry in the RPS and PCRVICS. The RPS and PCRVICS 
will continue to respond in performing its other design intended 
safety functions. The MSLRM system high radiation trip function for 
the MVP will be retained. The proposed TS changes do not involve any 
plant hardware changes that could introduce any new failure modes or 
effects. The MSLRM system radiation monitors will remain active to 
initiate Main Control Room (MCR) annunciation alarms. Plant 
procedures will be in place to implement the appropriate mitigative 
measures in response to a MSLRM system high radiation alarm signal.
    The SJAEs will continue to operate to remove non-condensable 
gases from the main condenser for processing by the Offgas Treatment 
system. The Offgas Treatment system will continue to function as 
designed to reduce offgas radioactivity levels prior to release to 
the environment.
    Since the Design Basis Accident analysis (i.e., CRDA) does not 
credit the MSLRM system high radiation trip function for reducing 
the radiological consequences of the postulated accident, the 
proposed TS changes have effectively been evaluated and are included 
in the existing analysis. That is, the CRDA analysis already assumes 
that the radioactive material released from the failed fuel is 
immediately transported to the main condenser prior to the MSIVs 
completely closing.
    The safety assessment and assumptions documents in GE Topical 
Report NEDO-31400A provide the basis for eliminating the MSLRM 
system high radiation trip function for initiating an automatic 
reactor scram and automatic closure of the MSIVs. The safety 
assessment provided in NEDO-31400A can also be applied to eliminate 
the MSLRM system high radiation trip function for initiating the 
closure of the Main Steam Line drain valves, since any radioactive 
material passing through these valves would be processed in the same 
fashion as that passing through the MSIVs. Eliminating the MSLRM 
system high radiation trip function for initiating the closure of 
the Main Steam and Reactor Water sample line valves will have a 
negligible impact. The sample lines are routed to a sample sink 
where inlet valves installed on the sample lines are normally 
closed. Downstream of the inlet valves are needle valves designed to 
control and limit sample line flow. The sample sink is located in 
the Reactor Enclosure and is enclosed, and air vented from its 
exhaust hood is passed through filters prior to release to the 
environment. The Reactor Enclosure ventilation duct radiation 
monitor samples air from the sample sink hood exhaust, and will 
isolate the Reactor Enclosure ventilation system if the radiation 
levels exceed the monitor's setpoint. There is the potential that a 
minimal amount of radioactive material could be released to the 
environment through this flowpath if the sample sink inlet and 
needle valves failed to properly function. This potential release 
has been evaluated and determined to a small fraction of the dose 
limit requirements specified in 10CFR100.
    Therefore, the proposed TS changes do not create the possibility 
of a new or different kind of accident previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed TS changes to eliminate the MSLRM system high 
radiation trip function for initiating an automatic reactor scram 
and automatic closure of the MSIVs, Main Steam line drains valves, 
and Main Steam and Reactor Water sample line valves do not change 
the conclusion reached in the LGS Updated Final Safety Analysis 
Report (UFSAR) that the calculated radiological consequences of the 
bounding Design Basis Accident (i.e., CRDA) will not exceed the dose 
limit requirements established by 10 CFR 100. The proposed TS 
changes will improve the overall reliability of the plant when 
compared to the existing system lineup configuration, since it will 
reduce the potential of an unnecessary plant transient occurring as 
a result of an inadvertent MSIV closure.
    A reliability assessment analysis was performed to evaluate the 
effects of eliminating the MSLRM system high radiation reactor scram 
function on reactivity control failure frequency and core damage 
frequency in GE Topical Report NEDO-31400A. This analysis indicated 
that there is a negligible increase in reactivity control frequency 
with the elimination of the MSLRM trip function. However, this 
increase is compensated for by the reduction in transient initiating 
events (i.e., inadvertent reactor scrams). This reduction in 
transient initiating events represents a reduction in core damage 
frequency and thus, results in a net improvement in safety.
    Therefore, the proposed TS changes do not involve a reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: Larry E. Nicholson, Acting

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

    Date of amendment request: November 30, 1993
    Description of amendment request: The amendment would extend the 
surveillance interval of the primary containment drywell-to-suppression 
chamber bypass leak test from the current 18-month interval as required 
by Technical Specification (TS) Surveillance Requirement 4.6.2.1.d to a 
40 +/- 10-month interval. This change would allow the drywell-to-
suppression chamber bypass test to coincide with the 10 CFR 50, 
Appendix J, Type A test (i.e., Containment Integrated Leakage Rate Test 
(CILRT)) interval.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The failure effects that are potentially created by the proposed 
Technical Specifications (TS) changes have been considered. The 
accident which is potentially negatively impacted by the proposed TS 
changes are any Loss of Coolant Accident (LOCA) inside primary 
containment with or without offsite power available.
    The proposed TS changes increase the surveillance interval of 
the drywell-to-suppression chamber bypass leak test required by TS 
Section 4.6.2.1.d, and will require that an additional test be 
performed on the downcomer vacuum breakers assemblies. The primary 
containment structure and associated equipment are not considered to 
be accident initiators, they act to mitigate the consequences of an 
accident. There are no physical or operational changes being made as 
a result of these proposed changes. Therefore, the probability of 
occurrence of an accident previously evaluated is not increased.
    There is a potential increased risk that an increase in the 
bypass leakage may go undetected for the duration of the proposed 
extension of the interval between the performance of the drywell-to-
suppression chamber bypass leak test. However, as discussed below, 
the increased risk is considered to be negligible due to the design 
of the diaphragm structure and past test data. Therefore, we have 
concluded that the probability of bypass leakage exceeding the 
allowed value is not increased as a result of the proposed TS 
changes.
    The proposed TS changes will extend the surveillance interval 
for the drywell-to-suppression chamber bypass leak test from 18 
months to 40 +/- 10 months. These proposed changes would allow this 
test to be performed at the same interval as the 10CFR50, Appendix 
J, Type A test (i.e., Containment Integrated Leakage Rate Test 
(CILRT)). In addition, the proposed changes will add an additional 
surveillance requirement to be performed on the vacuum breaker 
assemblies during refueling outages when the drywell-to-suppression 
chamber bypass leak test is not required to be performed. The 
proposed TS changes do not increase the consequences of an accident 
previously evaluated. This is based on the evaluation summarized 
below that demonstrates that the overall impact, if any, on the 
plant containment integrity is negligible. Furthermore, the 
performance history for the previous LGS bypass leak tests does not 
indicate any time based failures. The proposed TS changes also 
include a change to the frequency of testing, if two consecutive 
tests fail, from once every nine (9) months to once every 24 months 
in order to coincide with the 24 month refueling cycle. This change 
has no impact on the consequences of an accident based on 
maintaining the original requirement to increase the frequency of 
testing if two consecutive bypass leak tests fail, and maintaining a 
TS requirement for the NRC to review the schedule for subsequent 
tests.
    During a LOCA inside containment, potential leak paths between 
the drywell and suppression chamber airspace could result in 
excessive containment pressures, since the steam flow into the 
airspace would bypass the heat sink capabilities of the suppression 
pool. The containment pressure response to the postulated bypass 
leakage can be mitigated by manually actuating the suppression 
chamber sprays. Accordingly, since the sprays are manually actuated, 
an analysis was performed to show that the operator has sufficient 
time to initiate the sprays prior to exceeding the containment 
design pressure. This analysis is described in section 6.2.1.1.5 of 
the LGS Updated Final Safety Analysis Report (UFSAR). The analysis 
is based on a small break LOCA inside containment with a 
differential pressure between the drywell-to-suppression chamber 
equal to the static pressure due to downcomer submergence. The 
analysis concludes that the containment design pressure of 55 psig 
will be reached in over 30 minutes from the onset of a small break 
LOCA assuming a drywell-to-suppression chamber bypass flow area 
(i.e., A/square root of k) equal to 7.20 in2 without operator 
intervention.
    TS Limiting Condition for Operation 3.6.2.1.b conservatively 
specifies a maximum allowable bypass area of 10 % of the design 
value of 7.20 in2. This TS limit provides an additional safety 
factor of 10 above the conservatism taken in the steam bypass 
analysis (i.e., 0.720 in2). The drywell-to-suppression chamber 
bypass leak test required by TS Surveillance Requirement 4.6.2.1.d 
verifies that the actual bypass flow area is less than or equal to 
the TS limit of 0.720 in2. The bypass leakage test ensures that 
degradation in the measured bypass area is identified and corrected 
to ensure containment integrity during LOCA events.
    The potential bypass leakage paths can be divided into two 
categories as described below.
    1) Leakage pathways other than those associated with the 
drywell-to-suppression chamber vacuum breaker assemblies such as 
diaphragm floor penetrations (i.e., downcomer and Safety/Relief 
Valve (SRV) discharge line penetrations), cracks in the diaphragm 
floor and/or liner plate, and cracks in the downcomers and SRV 
discharge lines that pass through the suppression chamber airspace.
    2) The four sets of drywell-to-suppression chamber vacuum 
breaker assemblies.
    All other potential bypass leakage pathways have at least two 
isolation valves in the potential leakage path. These valves are 
high quality leak-tight containment isolation valves that are 
normally closed and receive an isolation signal to close. All Air 
Operated Valves (AOVs) in these paths fail closed.
    Several plant design features and the bypass leak test data 
measured to date confirm that the leakage from other than the vacuum 
breaker assemblies is negligible and indicates that this leakage 
will continue to be negligible for the proposed increased duration 
between tests. All pressure boundary penetrations between the 
drywell and the suppression chamber are welded except the vacuum 
breaker valves and the blind flanges closing 10 spare nozzles in the 
downcomers. All pressure boundary penetrations between the drywell-
to-suppression chamber have been fabricated, erected, and inspected 
in accordance with the American Society of Mechanical Engineers 
(ASME) Code, Section III, Subsection NC, 1971 Edition, with the 
exception of the tees supporting the vacuum breakers.
    The downcomer and SRV discharge lines penetrate through the 
diaphragm slab and terminate in the suppression pool. A steel ring 
plate is welded to the outside of the downcomers. The downcomer/ring 
plate assemblies are embedded in the diaphragm slab with the top 
surface of the ring plate flush with the drywell side of the 
diaphragm slab. All connections are welded to form a continuous 
steel membrane between the liner plate and downcomer penetrations. 
The SRV discharge lines are routed through welded flued heads at the 
diaphragm floor. The flued head design and construction are similar 
to the downcomer penetrations and also provide a continuous steel 
barrier. The downcomer and SRV discharge lines are designed and 
constructed to safety-related requirements. In addition, they are 
designed for all postulated loading conditions, including seismic, 
hydrodynamic, pressure, and temperature loads. The conservative 
design requirements ensure that the SRV discharge and the downcomer 
lines will not contribute to bypass leakage.
    The diaphragm floor is a reinforced concrete slab approximately 
3.5 feet thick. The drywell side surface of the diaphragm slab is 
capped with a 1/4 inch thick carbon steel liner plate. The liner 
plate and diaphragm slab provide a barrier against the potential for 
bypass leakage through the diaphragm floor. The structural integrity 
of the diaphragm floor and penetrations was demonstrated during the 
pre-operational test program. The drywell was pressurized to a 
drywell-to-suppression chamber differential pressure of above 30 
psid, which envelopes the maximum drywell-to-suppression chamber 
differential pressure postulated to occur during LOCA conditions.
    There have been six Unit 1 and three Unit 2 bypass leak tests 
performed in accordance with TS Surveillance Requirement 4.6.2.1.d. 
These tests were conducted at a drywell-to-suppression chamber 
differential pressure of at least 4.0 psid. The measured leakage 
area includes leakage from both the vacuum breakers and sources 
other than vacuum breakers.
    In all cases, the measured leakage is significantly less than 
the TS and design values. The maximum measured leakage areas are 
0.0400 in2 and 0.0111 in2 for Unit 1 and Unit 2, 
respectively; or 5.56% and 1.55 %, respectively, of the TS limit. 
The average values are 0.0180 in2 for Unit 1 and 0.0107 
in2 for Unit 2; or 2.5% and 1.49%, respectively, of the TS 
limit of 0.720 in2. The minimum measured leakage areas are 0.0 
in2 and 0.0100 in2 for Unit 1 and Unit 2, respectively, or 
0% and 1.3 %, respectively, of the TS limit. Clearly, the test data 
confirm that the bypass leakage measured to date at LGS has been 
negligible.
    In addition, we have obtained bypass leakage data from the 
Pennsylvania Power and Light Company, Susquehanna Steam Electric 
Station (SSES), Units 1 and 2, which also has Mark II containments 
with the Anderson Greenwood vacuum breakers (i.e., the same 
manufacturer as the vacuum breakers installed in the LGS, Unit 1 and 
Unit 2 containments) and therefore the data is applicable to LGS. 
The maximum bypass leakage area for the SSES Unit 1 containment was 
0.037 in2, and 0.009 in2 for the SSES Unit 2 containment, 
or 4.81% and 1.17%, respectively, of the SSES TS limit. Approval for 
a similar TS change for SSES, Units 1 and 2 was issued by the NRC by 
letter dated August 11, 1993.
    The remaining and most likely source of potential bypass leakage 
is the four sets of drywell-to-suppression chamber vacuum breakers. 
Each set consists of two vacuum breakers in series, flange mounted 
to a tee off the downcomers in the suppression chamber airspace. The 
drywell-to-suppression chamber bypass leak test is currently 
required by TS Surveillance requirement 4.6.2.1.d to be completed 
during each refueling outage and the results are used to verify that 
the total bypass area, including that due to the vacuum breakers, 
meets the TS limit. If maintenance has been performed on the vacuum 
breakers, this test also serves as a post-maintenance vacuum 
breakers leakage area test.
    The proposed TS changes decrease the frequency of the drywell-
to-suppression chamber bypass leak test. The drywell-to-suppression 
chamber bypass leak test data obtained following vacuum breakers 
maintenance cannot be utilized to determine vacuum breakers leakage 
reliability over the duration of the proposed test interval 
extension. To address this concern and collect additional vacuum 
breakers leakage data, the proposed TS changes include an additional 
requirement to perform a vacuum breaker leakage test as described 
below.
    The leakage test will be conducted on each set of vacuum 
breakers (i.e., four vacuum breakers sets per unit) during each 
refueling outage when the drywell-to-suppression chamber bypass leak 
test would not be required to be performed. If maintenance is 
performed on the vacuum breaker assemblies, this additional test 
will be performed post-maintenance to verify that the leakage is 
acceptable. This test will be conducted at a drywell-to-suppression 
chamber differential pressure of 4.0 psid (i.e., the same as 
differential pressure required for the drywell-to-suppression 
chamber bypass leak test) by either pressurizing the drywell side of 
the vacuum breakers or inducing a vacuum on the suppression chamber 
side of the vacuum breakers. The acceptance criteria for the vacuum 
breaker leakage tests will be as follows. The total vacuum breaker 
leakage areas for all four sets of vacuum breakers will be less than 
or equal to 24% of the TS limit (i.e., 0.24 x 0.720 in2 = 0.173 
in2). This proposed acceptable vacuum breaker leakage area 
provides a 76% margin to the TS limit to account for the leakage 
paths other than the vacuum breakers. As described above, previous 
bypass leakage testing measured a maximum bypass leakage area of 
5.56% of the TS limit. The 76% margin is sufficiently large to 
accommodate the other expected leakage sources. In addition, each 
set of vacuum breakers will be limited to a leakage area twice the 
assumed leakage from a single vacuum breaker set, assuming the 
leakage area is evenly distributed among the four sets of vacuum 
breakers (i.e., four sets equate to 24% of the TS Limit where each 
set is 6% and twice this total is 12% of the TS Limit). This allows 
a leakage of less than or equal to 0.0865 in2 (i.e., (0.173 
in2 divided by 4 sets of vacuum breakers) x (a factor of 2 
times the acceptable total) = 0.0865 in2) for an individual set 
of vacuum breakers. This criterion is stipulated to identify 
individual sets of vacuum breakers with higher leakage area.
    The drywell-to-suppression chamber bypass leak test data 
obtained during previous testing at LGS demonstrates conformance by 
a large margin compared to the TS and design leakage requirements. 
The test data indicates that there is negligible risk that the 
bypass leakage will change adversely in future years. Furthermore, 
the proposed test frequency is judged to be acceptable based on the 
risk of the leakage sources other than the vacuum breakers being 
essentially equivalent to that of the rest of the primary 
containment structure, which is leak tested (i.e., CILRT) every 40 
+/- 10 months as required by TS Surveillance Requirement 4.6.1.2.a. 
A bypass leak test will be developed and conducted to verify 
acceptable vacuum breaker bypass leakage areas for those outages 
when the bypass leak test will not be required to be performed. The 
proposed vacuum breaker leakage test with stringent acceptance 
criteria, combined with other negligible leakage areas, provide an 
acceptable level of assurance that the bypass leakage can be 
measured and an adverse condition can be detected and corrected such 
that the existing level of confidence that the primary containment 
will function as required during a LOCA is maintained.
    Therefore, the proposed TS changes will not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS changes involve the drywell-to-suppression 
chamber bypass leak test frequency. There are no physical or 
operational changes as a result of these proposed changes. These 
proposed changes include the requirement to perform an additional 
surveillance test on the vacuum breaker assemblies, applying a 
differential pressure of 4.0 psid which is the same differential 
pressure as currently required by TS for the drywell-to-suppression 
chamber bypass leak test. This required test will ensure that 
acceptable vacuum breaker leakage is maintained during those 
intervals when the drywell-to-suppression chamber bypass leak test 
is not required to be performed. Furthermore, the affected structure 
(i.e., primary containment) acts as an accident mitigator and not as 
an accident initiator. Accordingly, the possibility of a different 
type of malfunction of equipment or the possibility of an accident 
of a different type is not introduced.
    Therefore, the proposed TS changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The drywell-to-suppression chamber bypass leak test data 
obtained during previous testing at LGS demonstrates conformance by 
a large margin to the TS and design leakage requirements. The test 
data indicate that there is negligible risk that the bypass leakage 
will change adversely in future years. Furthermore, the proposed 
test frequency is judged to be acceptable based on the risk of 
sources of leakage other than the vacuum breakers being essentially 
equivalent to that of the rest of the primary containment structure, 
which is tested every 40 +/- 10 months. A bypass leak test will be 
developed and conducted to verify acceptable vacuum breaker bypass 
leakage areas for those outages when the bypass leak test will not 
be required to be performed. The proposed vacuum breaker leakage 
test with stringent acceptance criteria, combined with the other 
negligible potential leakage areas, provide an acceptable level of 
assurance that the bypass leakage can be measured and an adverse 
condition can be detected and corrected such that the existing 
levels of confidence that the primary containment will function as 
required during a LOCA is maintained.
    Therefore, the consequences of an accident are not impacted by 
this change and containment integrity during a LOCA will be 
maintained.
    Therefore, the proposed TS changes do not involve a reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: Larry E. Nicholson, Acting

Philadelphia Electric Company, Public Service Electric and Gas 
Company, Delmarva Power and Light Company, and Atlantic City 
Electric Company, Dockets Nos. 50-277 and 50-278, Peach Bottom 
Atomic Power Station, Units Nos. 2 and 3, York County, Pennsylvania

    Date of application for amendments: November 17, 1993
    Description of amendment request: The proposed Technical 
Specification (TS) changes to Surveillance Requirements would eliminate 
unnecessary emergency diesel generator (EDG) testing when a diesel 
generator or an offsite power source becomes inoperable. The proposed 
change would reduce the stresses on the diesel generators caused by 
unnecessary testing.
    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 because implementation of the proposed TS change, which 
would delete the requirement to demonstrate the operability of an 
otherwise operable EDG once the potential for a common cause failure 
has been dismissed, does not affect the design or performance 
characteristics of an EDG. Similarly, deleting the requirement to 
demonstrate the operability of EDGs when an offsite power source is 
inoperable does not affect the design or performance characteristics 
of an EDG. Therefore, the EDGs will maintain their ability to 
perform their design function. The EDGs are not assumed to be an 
initiator of any analyzed event. The role of the EDGs is the 
mitigation of accident consequences. Therefore, this proposed TS 
change does not increase the probability of an accident previously 
evaluated.
    The consequences of an accident previously evaluated could be 
affected by the proposed TS change. As described above, 
implementation of the proposed change will result in the EDGs 
maintaining their ability to perform their design function. 
Excessive testing of EDGs can cause reduced reliability. Precluding 
unnecessary testing of operable EDGs will improve EDG reliability 
and thereby have an overall positive affect on plant safety. 
Therefore, this proposed TS change does not increase the 
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 
because implementation of the proposed TS change will not involve 
physical changes to plant systems, structures, or components (SSC). 
The design or performance characteristics of the EDG will not be 
affected by the proposed change. The proposed change does not 
introduce any new modes of plant operation or make any changes to 
system setpoints which would initiate a new or different kind of 
accident. Therefore, the possibility of a new or different kind 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 because the proposed TS change does not affect 
the design or performance of any EDG. The change will increase EDG 
reliability by reducing the stresses on the EDG from unnecessary 
testing. This will result in an overall increase in plant safety. 
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: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: Charles L. Miller

Philadelphia Electric Company, Public Service Electric and Gas 
Company, Delmarva Power and Light Company, and Atlantic City 
Electric Company, Dockets Nos. 50-277 and 50-278, Peach Bottom 
Atomic Power Station, Units Nos. 2 and 3, York County, Pennsylvania

    Date of application for amendments: November 19, 1993
    Description of amendment request: The proposed change would 
eliminate the listing of specific position titles for the Plant 
Operations Review Committee (PORC) composition in favor of allowing the 
Plant Manager to appoint PORC members. This would eliminate the need to 
change the Technical Specifications (TSs) in the future whenever a 
position title is changed.
    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 because the proposed TS change is administrative in 
nature. The PORC member titles will be removed from the TS to 
facilitate not requiring that a TS change be submitted for NRC 
approval when position titles change. PORC member qualifications 
will continue to be consistent with those required for the Facility 
Staff and meet or exceed Sections 4.2, 4.4, or 4.6 of ANSI N18.1-
1971. The proposed change ensures that PORC will continue to be 
comprised of personnel involved in daily plant activities who are 
experienced individuals with varied expertise. By maintaining the 
qualification requirements for members of PORC who represent various 
areas of expertise, PORC will continue to fulfill its requirements 
specified in TS Section 6.5.1.6. The proposed change does not 
involve any physical changes to plant systems, structures, or 
components (SSC), or the manner in which these SSC are operated, 
maintained, modified, tested, or inspected. Therefore, the proposed 
TS 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 
because implementation of the proposed TS change will not involve 
physical changes to plant SSC or the manner in which these SSC are 
operated, maintained, modified, tested or inspected. The proposed 
change does not introduce any new modes of plant operation or make 
any changes to system setpoints which would initiate a new or 
different kind of accident. Therefore, the possibility of a new or 
different kind 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 because the proposed TS change is 
administrative in nature by providing internal flexibility in 
changing organizational titles and does not reduce the PORC function 
or responsibilities. PORC will continue to be filled by 
appropriately qualified personnel who have a variety of expertise. 
The change does not affect the plant material condition, operation, 
or accident analyses. 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: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: Charles L. Miller

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: November 24, 1993
    Description of amendments request: The proposed changes to the 
Technical Specification will relocate the reactor trip system and 
engineered safety feature actuation system response time limits from 
the TS to the Final Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed changes are administrative in nature and do not involve 
any change to the configuration or method of operation of any plant 
equipment used to mitigate the consequences of an accident. Also, 
the proposed changes do not alter the conditions or assumptions in 
any of the FSAR accident analyses. Since the FSAR accident analyses 
remain bounding, the radiological consequences previously evaluated 
are not adversely affected by the proposed changes. Therefore, the 
proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed changes are administrative in nature and do 
not involve any change to the configuration or method of operation 
of any plant equipment used to mitigate the consequences of an 
accident.Accordingly, no new failure modes have been defined for any 
plant system or component important to safety nor has any new 
limiting failure been identified as a result of the proposed 
changes. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety. The proposed changes are administrative in 
nature and will continue to ensure that the response times for the 
RTS and ESFAS instrumentation do not exceed the limits assumed in 
the accident analyses. As a result of the proposed changes, response 
time limits for the RTS and ESFAS will be administratively 
controlled in accordance with the provisions of 10 CFR 50.59, thus 
eliminating an unnecessary burden of governmental regulation without 
reducing protection for public health and safety. Therefore, the 
proposed changes do not involve a significant reduction in a margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Attorney for licensee: James H. Miller, III, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: S. Singh Bajwa

Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260 and 
50-296, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, 
Limestone County, Alabama

    Date of amendment request: September 30, 1993 (TS 337)
    Description of amendment request: The proposed amendments provide 
an administrative vehicle for modifying a condition of the facility 
operating license for each of the BFN units. The condition requires the 
licensee to implement and maintain in effect all provisions of the 
``Fire Protection Program (FPP)'' and lists the U.S. Nuclear Regulatory 
Commission (NRC) staff safety evaluations (SE) approving the FPP. If 
the staff approves of a revision currently under review to an element 
of the FPP, the ``Appendix R Safe Shutdown Program (SSP)'', the 
proposed amendments would add the staff SE documenting approval of the 
revised SSP to the above listing of SEs in each facility operating 
license. The current SSP is directed toward the safe shutdown of only 
one operating plant (Unit 2). The revised SSP would be directed toward 
the safe shutdown of two operating plants (Units 2 and 3).
    Additionally, the proposed amendments add the definition of the SSP 
to Section 1.0 of the Unit 3 Technical Specifications (TS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This proposed change is administrative in nature. The proposed 
change is being made to revise the license condition to reflect a 
combined Unit 2 and 3 Appendix R Safe Shutdown Program following NRC 
approval. Compliance with the applicable Appendix R requirements is 
ensured through implementation of the Fire Protection Program and 
the Appendix R Safe Shutdown Program. The change does not affect any 
design bases accident or the ability of any safe shutdown equipment 
to perform its function. Also, there are no physical modifications 
required to implement this TS change. Therefore, these proposed 
administrative changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change is administrative in nature. The proposed 
change is being made to revise the license condition to reflect a 
combined Unit 2 and 3 Safe Shutdown Program following NRC approval. 
Compliance with the applicable Appendix R requirements is ensured 
through implementation of the Fire Protection Program and Appendix R 
Safe Shutdown Program. This change does not affect any design basis 
accident or the ability of any safe shutdown equipment to perform 
its function. Also, there are no physical modifications required to 
implement this TS change. Therefore, these proposed administrative 
changes do not create the possibility of a new or different kind of 
accident from an accident previously evaluated.
    3. This change does not involve a significant reduction in the 
margin of safety.
    The proposed changes are administrative in nature. Compliance 
with the applicable Appendix R requirements is ensured through the 
implementation of the Fire Protection Program and Appendix R Safe 
Shutdown Program. The proposed change does not affect any design 
basis accident and does not reduce or adversely affect the 
capability to achieve and maintain safe shutdown in the event of a 
fire. Furthermore, no reductions to the requirements for equipment 
operability, surveillance requirements or setpoints are being made 
which could result in reduction in the margin of safety. Therefore, 
these proposed administrative changes will not result in a reduction 
in the margin of safety.
    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Mr. Frederick J. Hebdon

Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260 and 
50-296, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, 
Limestone County, Alabama

    Date of amendment request: October 12, 1993 (TS 320)
    Description of amendment request: The proposed amendment would 
delete reference in the BFN Unit 3 Technical Specifications (TS) to the 
Reactor Water Cleanup (RWCU) system floor drain high temperature 
switches and the RWCU system space high temperature switches. The 
piping configuration for the Unit 3 RWCU system has been modified, and 
the licensee contends that its revised High Energy Line Break (HELB) 
analysis has demonstrated that these switches are no longer required. 
Instead, to initiate RWCU system isolation, the HELB analysis has 
indicated the need for temperature switches in the main steam vault, 
the heat exchanger room, and the RWCU pipe trench. The proposed 
amendment therefore would add temperature switches to the Unit 3 TS for 
these areas and modify the TS Bases section accordingly. The proposed 
amendment also adds clarifying remarks to Tables 3.2.A and 3.2.B of the 
TS for each of the BFN units. The proposed remarks list the actuation 
signals for the various primary containment valve group isolations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    An analysis of HELBs in the Unit 3 reactor building identified 
certain RWCU pipe breaks which could not be automatically detected 
and isolated in a reasonable time frame. To resolve this issue, a 
design change is being performed to remove from service the existing 
non-environmentally qualified temperature switches used to detect 
RWCU line breaks and replace them with environmentally qualified 
RTDs [resistance temperature detectors] and IEEE [Institute of 
Electrical and Electronics Engineers] Class 1E qualified ATUs 
[analog trip units] located to detect and isolate the critical RWCU 
pipe breaks. This TS amendment adds the new ATUs [sic] function to 
Tables 3.2.A and 4.2.A. Note 14 is deleted from Table 3.2.A since it 
only applies to the temperature switches being removed from the 
table.
    The safety function of the RTD/ATU temperature loops is to 
provide an isolation signal to close the RWCU suction line isolation 
valves (FCV-69-001 and FCV-69-002) and RWCU return line valve (FCV-
69-012) on a high area temperature. This ensures RWCU pipe breaks 
are isolated. No other RWCU safety functions are affected by the 
change.
    The new RTD/ATU temperature loops were chosen to decrease the 
time required to initiate closure of the RWCU valves. This improves 
the detection/isolation of RWCU breaks and helps to limit the 
reactor coolant lost, helps ensure core cooling, and helps ensure 
that environmental conditions inside the reactor building are 
maintained within the required limits.
    Components added by this change are qualified for the 
environment in which they will operate. This ensures that the system 
will perform its function in a post accident environment. No 
additional paths for the release of radiation or contamination are 
created. The failure modes of the RTDs and ATUs are such that any 
single failure will result in a gross failure alarm and/or a channel 
trip. Because of the redundancy, separations, and logic designed 
into the system, a single failure of any part of the system will not 
prevent isolation of the primary containment isolation valves and 
spurious operation is minimized. The RTDs will be located and the 
instrument setpoints will be set to preclude spurious trips due to 
ambient temperatures including localized hot areas while assuring a 
timely trip due to a pipe break. Therefore, the proposed amendment 
does not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    This change is being made to improve the RWCU leak detection/
isolation function of the RWCU Primary Containment Isolation System 
(PCIS). The PCIS will perform its intended safety function in the 
same manner as the previous installation. There is no affect [sic] 
on the function or operation of any other plant system.
    Failure of the RTD/ATU temperature loops would be no different 
than failure of existing temperature switches. Since environmental 
qualification requirements, divisional separation, single failure 
requirements and one-out-of-two taken twice logic requirements are 
maintained, the possibility of a RWCU isolation failure on a RWCU 
line break or of a spurious isolation is no more likely after the 
change than before.
    In the existing design, logic relays are powered from RPS Bus A 
or B. The new design also uses RPS Bus A or B to feed the ATUs. 
Therefore, the consequence of a power failure is unchanged from the 
present design. The seismic qualification and proper circuit 
coordination of the installation is maintained. The system functions 
and operates in the same manner as previously evaluated in the 
Safety Analysis Report. No new system interactions other than 
additional RTDs located in the main steam valve vault to input into 
the PCIS logic for isolation of the RWCU have been introduced by 
this activity. Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The margin of safety will be enhanced by installing instruments 
that provide quicker response to a temperature rise indicative of a 
pipe break. Calculations have been performed to determine the 
analytical limits for the RTD/ATU temperature loops in each of the 
monitored areas and to determine the setpoints for the ATUs in each 
area. The setpoints are set above the maximum expected room 
temperatures to avoid spurious actuations due to ambient conditions 
and below the analytical limits to ensure timely detection of a pipe 
break. This type of design utilizing ATUs has been analyzed by the 
NRC [U.S. Nuclear Regulatory Commission staff] (NEDO-21617, Analog 
Transmitter/Trip Unit System for Engineered Safeguard Sensor Trip 
Input) and has been found to be generically acceptable at BWR 
facilities. Therefore, the proposed amendment does not involve a 
significant reduction in any margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Mr. Frederick J. Hebdon

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

    Date of amendment request: December 10, 1993
    Description of amendment request: The proposed change would change 
the Technical Specifications (TS) for the North Anna Power Station, 
Units No. 1 and No. 2 (NA-1&2).
    Specifically, the proposed changes would modify the surveillance 
frequency of the Auxiliary Feedwater System pumps from monthly to 
quarterly in accordance with the guidance provided in Generic Letter 
93-05, ``Line-Item Technical Specifications Improvements to Reduce 
Surveillance Requirements for Testing During Power Operation,'' dated 
September 27, 1993.
    The NRC has completed a comprehensive examination of surveillance 
requirements in TS that require testing at power. The evaluation is 
documented in NUREG-1366, ``Improvements to Technical Specification 
Surveillance Requirements,'' dated December 1992. The NRC staff found, 
that while the majority of testing at power is important, safety can be 
improved, equipment degradation decreased, and an unnecessary burden on 
personnel resources eliminated by reducing the amount of testing at 
power that is required by TS. Based on the results of the evaluations 
documented in NUREG 1366, the NRC issued Generic Letter 93-05.
    The Auxiliary Feedwater System supplies water to the steam 
generators to remove decay heat from the Reactor Coolant System. To 
ensure operability of the Auxiliary Feedwater System, the pumps are 
currently tested on a monthly basis as required by the TS. Consistent 
with Generic Letter 93-05, Item 9.1 and NUREG-1366, the licensee is 
requesting a change to the surveillance testing frequency for the 
Auxiliary Feedwater Pumps from monthly to quarterly on a staggered test 
basis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Specifically, operation of North Anna Power Station in 
accordance with the proposed Technical Specifications changes will 
not:
    1. Involve a significant increase in the probability of 
occurrence or consequences of an accident previously evaluated.
    Changing the surveillance test frequencies of the Auxiliary 
Feedwater System pumps does not significantly affect the probability 
of occurrence or consequences of any previously evaluated accidents. 
Quarterly testing of the pumps on a staggered basis will continue to 
assure that the Auxiliary Feedwater System will be capable of 
performing its intended functions. Therefore, the change in 
frequency of testing the Auxiliary Feedwater System pumps does not 
affect the probability or consequences of any previously analyzed 
accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Changing the surveillance test frequency of the Auxiliary 
Feedwater System pumps does not involve any physical modification of 
the plant or result in a change in a method of operation. Quarterly 
testing of the Auxiliary Feedwater System pumps on a staggered basis 
will continue to assure that the Auxiliary Feedwater System will be 
capable of performing its intended function. Therefore, a new or 
different type of accident is not made possible.
    3. Involve a significant reduction in a margin of safety.
    Changing the surveillance test frequency of the Auxiliary 
Feedwater System pumps does not affect any safety limits or limiting 
safety system settings. System operating parameters are unaffected. 
The availability of equipment required to mitigate or assess the 
consequence of an accident is not reduced. Quarterly testing of the 
Auxiliary Feedwater System pumps on a staggered basis will continue 
to assure that the Auxiliary Feedwater System will be capable of 
performing its intended functions. Safety margins are, therefore, 
not decreased.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: Herbert N. Berkow

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 20555, and at the local public document 
rooms for the particular facilities involved.

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

    Date of application for amendment: October 5, 1993
    Brief description of amendment: The proposed change to the 
Technical Specifications would revise the wording of liquid release 
rate limit and its associated bases, and relocate the old 10 CFR 20.106 
requirements to the new 10 CFR 20.1302 to be consistent with the 
revised terminology of 10 CFR Part 20. The new wording will retain the 
same overall level of effluent control required to meet the design 
objectives of Appendix I to 10 CFR Part 50.
    Date of issuance: December 14, 1993
    Effective date: December 14, 1993
    Amendment No.: 40
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59746) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 14, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

    Date of application for amendments: August 5, 1992
    Brief description of amendments: The amendment revises the 
Braidwood Station, Units 1 and 2, Technical Specifications (TS) 
regarding Engineered Safety Features Actuation System (ESFAS) 
instrumentation. The ESFAS, Functional Units, Analog Channel 
Operational Test interval is changed from monthly to quarterly. 
Eighteen changes to the Reactor Trip System (RTS) are also included in 
this TS change.
    Date of issuance: December 16, 1993
    Effective date: December 16, 1993
    Amendment Nos.: 44 and 44
    Facility Operating License Nos. NPF-72 and NPF-77. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 28, 1992 (57 FR 
48816) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 16, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Wilmington Township Public 
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.

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

    Date of application for amendments: October 29, 1993
    Brief description of amendments: The amendments revise Table 3.6.3-
1, ``Primary Containment Isolation Valves,'' of the LaSalle Technical 
Specifications for Units 1 and 2 by adding a new category of valves to 
these tables. There are a total of eight new valves added in each 
table, consisting of two check valves in each of four backfill lines. 
The backfill lines were added in response to NRC Bulletin 93-03, 
``Resolution of Issues Related to Reactor Vessel Water Level 
Instrumentation in BWRs,'' dated May 28, 1993.
    Date of issuance: December 10, 1993
    Effective date: December 10, 1993
    Amendment Nos.: 92 and 76
    Facility Operating License Nos. NPF-11 and NPF-18. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1993 (58 FR 
59493). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 10, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Public Library of Illinois 
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348.

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

    Date of application for amendment: November 25, 1992, as 
supplemented by letter dated February 5, 1993.
    Brief description of amendment: The amendment revises surveillance 
intervals for Process Radiation Monitors, Area Radiation Monitors, the 
Main Steam Line Radiation Monitors, the Auxiliary Feedwater System 
Initiating Logic, the Main Steam Safety Valves Setpoints, and the Toxic 
Gas Detection System Monitors to accommodate a 24-month refueling 
cycle. These revisions are being made in accordance with the guidance 
provided by Generic Letter 91-04, ``Changes in Technical Specification 
Surveillance Intervals to Accommodate a 24-Month Fuel Cycle.''
    Date of issuance: December 16, 1993
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 166
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 25, 1993 (58 FR 
16219) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 16, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendments: October 5, 1993, as 
supplemented November 15 and 22, 1993
    Brief description of amendments: The amendments revise the 
Technical Specifications to reflect the appropriate operability 
requirements for cold leg accumulator water volume and surveillance 
requirements values for the centrifugal changing pumps, safety 
injection pumps, and residual heat removal pumps to prevent possible 
runout conditions during a loss of coolant accident event.
    Date of issuance: December 15, 1993
    Effective date: December 15, 1993
    Amendment Nos.: 110 and 104
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 27, 1993 (58 FR 
57848) The November 15 and 22, 1993, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 15, 1993. No significant hazards 
consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

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

    Date of application for amendments: October 5 and 14, 1993, as 
supplemented November 15 and December 14, 1993
    Brief description of amendments: The amendments revise the 
Technical Specifications to allow the implementation of interim steam 
generator tube plugging criteria for the tube support plate elevations.
    Date of issuance: December 16, 1993
    Effective date: December 16, 1993
    Amendment Nos.: 111 and 105
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 27, 1993 (58 FR 
57849) The November 15 and December 14, 1993, letters provided 
clarifying information and revisions to the coolant specific activity 
that did not change the scope of the original application and 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 December 16, 1993. No significant hazards 
consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

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

    Date of application for amendments: September 7, 1993
    Brief description of amendments: The amendments revise the 
Technical Specification to (a) reduce the slope of the axial power 
imbalance penalty in the overtemperature-delta temperature reactor 
protection system trip setpoint equation, and (b) increase the boron 
concentration limits in the cold leg accumulators, the refueling water 
storage tank, the reactor coolant system, and refueling canal during 
MODE 6 conditions. These changes reflect the reloading of Unit 1 with 
Mark BW fuel for Cycle 8 including an increase in cycle length from 350 
effective full power days (EFPD) to 390 EFPD.
    Date of issuance: December 17, 1993
    Effective date: December 17, 1993
    Amendment Nos.: 112 and 106
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 27, 1993 (58 FR 
57847) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 17, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

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

    Date of application for amendments: October 25, 1993, as 
supplemented December 3 and 6, 1993
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) Figure 2.1-1, certain TS Table 2.2-1 
factors in the equation for the OVERTEMPERATURE delta T and OVERPOWER 
delta T setpoints, and Figure 3.2-1 to reflect a reduction in the 
required minimum measured reactor coolant system (RCS) flow rate from 
385,000 gallons per minute (gpm) to 382,000 gpm for Unit 1. Catawba 
Unit 2 values are unchanged and, accordingly, certain TS pages were 
modified to retain the current TS values in effect for Unit 2.
    The need for these changes is attributed to the effects of steam 
generator tube plugging and to a hot leg temperature streaming 
phenomenon. The application also proposed to revise the text of TS 
2.1.1 and the definition for TS Figure 2.1-1. These changes are not 
related to the changes in RCS flow rate. The staff is continuing to 
review these proposed changes and, accordingly, they are not dealt with 
in this amendment.
    Date of issuance: December 17, 1993
    Effective date: Effective within 30 days of its date of issuance.
    Amendment Nos.: 113 and 107
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59747) The December 3 and 6, 1993, letters provided clarifying 
information that did not change the scope of the October 25, 1993, 
application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 17, 1993. No significant hazards 
consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

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

    Date of application for amendment: May 20, 1993
    Brief description of amendment: The amendment removed unnecessary 
operability requirements for the Intermediate Range Monitors (IRMs) and 
the Average Power Range Monitors (APRMs) during plant shutdown 
operations.
    Date of issuance: December 13, 1993
    Effective date: December 13, 1993
    Amendment No: 109
    Facility Operating License No. NPF-29. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 23, 1993 (58 FR 
34077) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 13, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Judge George W. Armstrong 
Library, Post Office Box 1406, S. Commerce at Washington, Natchez, 
Mississippi 39120.

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: July 23, 1993
    Brief description of amendments: The amendments are necessary to 
implement new Standards for Protection Against Radiation (10 CFR Part 
20).
    Date of issuance: December 16, 1993
    Effective date: December 16, 1993
    Amendment Nos.: 125, 63
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 1, 1993 (58 
FR 46234) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 16, 1993. 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

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

    Date of application for amendment: October 8, 1993
    Brief description of amendment: The amendment deletes portions of 
the Oyster Creek Nuclear Generating Station Radiological Effluent 
Technical Specifications and relocates them to controlled programs in 
accordance with the guidance contained in NRC Generic Letter 89-01, 
dated January 31, 1989.
    Date of issuance: December 13, 1993
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 166
    Facility Operating License No. DPR-16. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59749) The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated December 13, 1993. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, New Jersey 
08753.

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

    Date of application for amendment: October 18, 1993
    Brief description of amendment: The amendment revises the Technical 
Specifications to delete requirements to demonstrate by testing, that a 
redundant system/component is operable when a system/component is 
declared inoperable. In lieu of testing the redundant system/component 
to demonstrate its operability the Technical Specifications are being 
revised to require an administrative check of plant records to verify 
operability of the redundant system/component. Confirming changes are 
made to Definition 1.1 ``Operable-Operability.''
    Date of issuance: December 21, 1993
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 167
    Facility Operating License No. DPR-16. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59749) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 21, 1993. No 
significant hazards consideration comments received: Yes. Comments were 
provided by letter dated December 10, 1993, from the State of New 
Jersey, Department of Environmental Protection and Energy, Division of 
Environmental Safety, Health and Analytical Programs. The comments and 
the NRC staff's response are addressed in the Commission's Safety 
Evaluation dated December 21, 1993.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, New Jersey 
08753

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

    Date of application for amendment: September 20, 1993, as 
supplemented on October 1, 1993.
    Brief description of amendment: The amendment revises the plant 
Technical Specifications to reflect a partial GPU Nuclear 
reorganization to become effective when Three Mile Island, Unit 2 (TMI-
2), enters the Post-Defueling Monitored Storage (PDMS) mode. This 
reorganization includes deleting TMI-2 as a Division and incorporating 
those functions and responsibilities required to maintain the PDMS 
condition and requirements into the current TMI-1 Division. The TMI-1 
Division will be renamed the TMI Division.
    Date of issuance: December 13, 1993
    Effective date: No specific date has been specified by the staff 
for the effectiveness of this amendment. The amendment will become 
fully effective at such time as the Vice President - TMI has been 
delegated the full responsibility of the overall safe operation of both 
TMI-1 and TMI-2.
    Amendment No.: 179
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 13, 1993 (58 FR 
52987). The October 1, 1993, submittal provided clarifying and 
corrected TS pages which did not change the initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of this amendment is contained in a Safety 
Evaluation dated December 13, 1993. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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

    Date of application for amendment: August 9, 1993
    Brief description of amendment: The amendment revises the plant 
Technical Specifications to be consistent with a major revision to 10 
CFR Part 20 that is to be implemented by January 1, 1994.
    Date of issuance: December 21, 1993
    Effective date: As of the date of issuance to be implemented on 
January 1, 1994.
    Amendment No.: 180
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59751). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 21, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105

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

    Date of application for amendment: August 24, 1993.
    Brief description of amendment: The amendment revises the plant 
Technical Specifications to adopt the Standard Technical Specification 
(STS) provision that allows a period up to 24 hours to complete a 
surveillance requirement upon the discovery that the surveillance has 
been missed.
    Date of issuance: December 22, 1993
    Effective date: As of its date of issuance, to be implemented 
within 30 days of issuance.
    Amendment No.: 181
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59751). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 22, 1993. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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

    Date of amendment request: January 13, 1993, as supplemented by 
letter dated October 18, 1993.
    Brief description of amendment: The amendment revises the River 
Bend, Unit 1 operating license to reflect a change in ownership of Gulf 
States Utilities (GSU). GSU, which owns a 70 percent undivided interest 
in the River Bend Station, will become a wholly-owned subsidiary 
company of Entergy Corporation.
    Date of issuance: December 16, 1993
    Effective date: December 6, 1993, to be implemented within 180 days 
of issuance.
    Amendment No.: Amendment No. 69
    Facility Operating License No. NPF-47: The amendment revised the 
license.
    Date of initial notice in Federal Register: July 7, 1993 (58 FR 
36435) The October 18, 1993, supplemental letter provided additional 
clarifying information and did not change the initial no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
December 16, 1993. No significant hazards consideration comments 
received: Yes. Comments and a request for hearing were received from 
Cajun Electric Power Cooperative of Baton Rouge, Louisiana.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803

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: January 13, 1993, as supplemented by 
letter dated June 29, 1993.
    Brief description of amendment: The amendment revises the River 
Bend Station, Unit 1 operating license to include as a licensee, 
Entergy Operations, Inc. (EOI), and to authorize EOI to use and operate 
River Bend and to possess and use related licensed nuclear materials.
    Date of issuance: December 16, 1993
    Effective date: December 16, 1993 to be implemented within 180 days 
of issuance.
    Amendment No.: 70
    Facility Operating License No. NPF-47: The amendment revised the 
license.
    Date of initial notice in Federal Register: July 7, 1993 (58 FR 
36436) The June 29, 1993, supplemental letter provided additional 
clarifying information and did not change the initial no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
December 16, 1993. No significant hazards consideration comments 
received: Yes. Comments and a request for hearing were received from 
Cajun Electric Power Cooperative of Baton Rouge, Louisiana.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803

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

    Date of application for amendment: June 18, 1993
    Brief description of amendment: The proposed changes to Technical 
Specifications 6.2.3.1, ``Independent Safety Engineering Group (ISEG) 
Function;'' 6.2.3.4, ``ISEG Records;'' 6.4.1, ``Training;'' and 
6.5.2.2, ``Nuclear Review and Audit Group (NRAG) Composition'' are 
editorial changes reflecting recent administrative/organizational 
changes which occurred at Clinton Power Station.
    Date of issuance: November 29, 1993
    Effective date: Immediately, to be implemented within 30 days.
    Amendment No.: 86
    Facility Operating License No. NPF-62. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 18, 1993 (58 FR 
43927) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 29, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: The Vespasian Warner Public 
Library District, 310 N. Quincy Street, Clinton, Illinois 61727

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook 
Nuclear Plant, Unit No. 2, Berrien County, Michigan

    Date of application for amendment: April 16, 1993, as supplemented 
September 28 and December 3, 1993
    Brief description of amendment: The amendment revises Technical 
Specifications to allow certain tests normally designated as 18-month 
surveillances to be delayed until the next refueling outage scheduled 
to begin August 6, 1994.
    Date of issuance: December 22, 1993
    Effective date: December 22, 1993
    Amendment No.: 158
    Facility Operating License No. DPR-74. Amendments revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 4, 1993 (58 FR 
41505) The supplemental letters provided clarifying information which 
did not change the staff's initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated December 22, 1993. 
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. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: June 7, 1993
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3/4.8.1, ``AC Sources-Operating,'' and associated 
Bases to eliminate unnecessary diesel generator testing when a diesel 
generator or an offsite power source becomes inoperable. The amendment 
is intended to increase diesel generator reliability and the overall 
level of plant safety by reducing the stresses on the diesel generators 
caused by unnecessary testing. The amendment also makes additional 
changes to TS 3/4.8.1 to further enhance diesel generator reliability 
and incorporate certain administrative changes.
    Date of issuance: December 15, 1993
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 54
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 7, 1993 (58 FR 
36440) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 15, 1993. 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.

North Atlantic Energy Service Corporation, Docket No. 50-443, 
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: November 11, 1992, as supplemented by 
letters dated July 2, 1993, and November 24, 1993.
    Description of amendment request: The amendment modifies the 
Seabrook Station Technical Specifications to allow the use of either 
the fixed incore detector system or the movable incore detector system 
to perform technical specification surveillances. Specifically, the 
amendment modifies Technical Specification sections 3.1.3, 4.2.2, 
4.2.3, 4.2.4, and 3.3.3.
    Date of issuance: December 22, 1993
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 27
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 3, 1993 (58 FR 
7002). The licensee's letters dated July 2, 1993, and November 24, 
1993, provide additional information and clarification to the 
application but do not change the initial proposed no significant 
hazards consideration determination and do not provide information 
outside the scope of the original Federal Register notice. The 
licensee's November 24, 1993, letter provides a commitment to acquire, 
through the end of Cycle 4, a limited number of flux maps using the 
movable incore detector system for comparison to flux maps obtained 
using the fixed incore detector system. Additionally, the licensee 
committed to provide a report to the NRC at the end of Cycle 4 
regarding comparison of the data obtained from both systems. The NRC's 
approval of the requested TS changes is conditioned upon the licensee's 
implementing the commitment. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated December 22, 1993. 
No significant hazards consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 47 
Front Street, Exeter, New Hampshire 03833.

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 19, 1993
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) to reflect staff positions and improvements to the 
TS in response to Generic Letter 90-06, ``Resolution of Generic Issue 
70, `Power-Operated Relief Valve and Block Valve Reliability, and 
Generic Issue 94, `Additional Low-Temperature Overpressure Protection 
for Light Water Reactors.''' Generic Issue 94 was closed out by 
Amendment 80 dated July 12, 1993. With the issuance of this TS 
amendment, we consider the licensee's response to Generic Letter 90-06 
and Generic Issue 70 (TAC No. M77362) complete for the Millstone 
Nuclear Power Station, Unit No. 3.
    Date of issuance: December 16, 1993
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 88
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 9, 1993 (58 FR 
32388) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 16, 1993. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.

Pennsylvania Power and Light Company, Docket No. 50-387, 
Susquehanna Steam Electric Station, Unit 1, Luzerne County, 
Pennsylvania

    Date of application for amendment: July 21, 1993
    Brief description of amendment: The amendment revised the Technical 
Specifications to modify the requirement for acquisition of baseline 
data on single-loop operation from during startup testing following 
each refueling outage to at least once per 18 months.
    Date of issuance: December 10, 1993
    Effective date: December 10, 1993
    Amendment No.: 131
    Facility Operating License No. NPF-14: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 4, 1993 (58 FR 
41509) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 10, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.

Portland General Electric Company, et al., Docket No. 50-344, 
Trojan Nuclear Plant, Columbia County, Oregon

    Date of application for amendment: April 1, 1993 as supplemented 
June 9 and August 5, 1993.
    Brief description of amendment: This amendment relocates the 
Radiological Effluent Technical Specifications (RETS) to the Offsite 
Dose Calculation Manual (ODCM) and Process Control Program (PCP) in 
accordance with NRC staff Generic Letter 89-01, and changes the 
required frequency for submittal of the radioactive Effluent Release 
Report from semiannual to annual in accordance with 10 CFR 50.36(a).
    Date of issuance: December 6, 1993
    Effective date: 30 days from date of issuance
    Amendment No.: 193
    Facility Operating License No. NPF-1: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 7, 1993 (58 FR 
36442) The supplements proposed additional changes and clarification to 
the TS regarding the frequency of effluent reporting. The changes were 
within the scope of the action described in the notice and did not 
change the initial no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated December 6, 1993. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Branford Price Millar Library, 
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151, 
Portland, Oregon 97207

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: September 25, 1992
    Brief description of amendment: The amendment to the Technical 
Specifications (TSs) deletes the surveillance requirements for the 
iodine analyzer portion of the drywell atmosphere Continuous Atmosphere 
Monitoring system from TS Table 4.6-2 and makes accompanying changes to 
TS Bases Section 3.6/4.6.D. These changes are consistent with the 
guidance in Regulatory Guide 1.45, ``Reactor Coolant Boundary Leakage 
Detection Systems.''
    Date of issuance: December 9, 1993
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 200
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 25, 1993 (58 FR 
16229) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 9, 1993. 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 Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
County, New Jersey

    Date of application for amendments: May 26, 1992
    Brief description of amendments: These amendments increase the 
shutdown margin requirements for the current operating cycle at both 
units; reduce the containment pressure high-high setpoint and allowable 
value; and change the containment spray system, containment fan cooler 
and service water system response times. These changes were 
necessitated by the discovery of containment fan cooler unit and 
containment spray system response times greater than originally assumed 
for Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB) 
analysis, and auxiliary feedwater system flow greater than assumed for 
the MSLB analysis.
    Date of issuance: December 16, 1993
    Effective date: December 16, 1993
    Amendment Nos. 149 and 127
    Facility Operating License Nos. DPR-70 and DPR-75. These amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 19, 1992 (57 FR 
37571) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 16, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

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

    Date of application for amendment: June 28, 1991
    Brief description of amendment: The amendment revised Technical 
Specification 3.5.1 to add a required action to periodically monitor 
alternative indication if one or both automatic depressurization system 
(ADS) safety related instrument air header(s) low pressure alarm system 
instrumentation channels become inoperable.
    Date of issuance: December 13, 1993
    Effective date: December 13, 1993
    Amendment No.: 52
    Facility Operating License No. NPF-58. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 7, 1991 (56 FR 
37590) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 13, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

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

    Date of application for amendment: March 19, 1991
    Brief description of amendment: The amendment revised Technical 
Specification Table 3.3.3-1 to make the required actions for the 
automatic depressurization system (ADS) consistent with the as-built 
configuration of the system. An editorial change to Action statement 32 
was added to this amendment to make Action statement 32 consistent with 
Action statements 30 and 33.
    Date of issuance: December 17, 1993
    Effective date: December 17, 1993
    Amendment No.: 53
    Facility Operating License No. NPF-58. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 15, 1991 (56 FR 
22480) Additional clarifying information was provided verbally by the 
utility on November 5, 1993, that did not change the initial proposed 
no significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated December 17, 1993. No significant hazards consideration comments 
received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.

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

    Date of application for amendment: November 9, 1992 as supplemented 
on November 22, 1993
    Brief description of amendment: The amendment revises the Technical 
Specifications to allow the de-energization of the borated water 
storage tank outlet isolation valves in the open position during 
operational Modes 1, 2, 3, and 4.
    Date of issuance: December 16, 1993
    Effective date: December 16, 1993
    Amendment No. 182
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1993 (58 FR 
28061) The supplemental letter provided additional information that did 
not change the proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a safety evaluation dated December 16, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

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

    Date of application for amendment: August 1, 1991
    Brief description of amendment: The amendment revises the Technical 
Specification Sections 3.3.3.6, 3.6.4.1, 4.11.2.5, 6.2.2, and Tables 
3.3-4 and 3.3-10 to correct typographical errors and make editorial 
changes.
    Date of issuance: December 21, 1993
    Effective date: December 21, 1993
    Amendment No.: 86
    Facility Operating License No. NPF-30. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 18, 1991 (56 
FR 47244) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 21, 1993. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity For a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 
application for amendment, (2) the amendment to Facility Operating 
License, 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 20555, 
and at the local public document room for the particular facility 
involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By February 4, 1994, 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 20555 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. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    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 20555, 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 the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Commonwealth Edison Company, Docket No. STN 50-456, Braidwood 
Station, Unit 1, Will County, Illinois

    Date of application for amendment: November 12, 1993, as 
supplemented by letters dated November 18 and December 6, 1993
    Brief description of amendment: The amendment changes the existing 
technical specifications (TS) by adding a footnote to TS 4.4.5.0 to 
address steam generator (SG) operability requirements. The change 
references an unscheduled inspection of the 1C SG which occurred due to 
a tube leak in that SG. The amendment was required because the 
circumstances of the inspection were not covered by the existing TS 3/
4.4.5. It will allow SG operability requirements to be satisfied until 
the next SG inservice inspection, scheduled to begin no later than 
March 5, 1993.
    Date of issuance: December 16, 1993
    Effective date: December 16, 1993
    Amendment No.: 43
    Facility Operating License No. NPF-72. The amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: No. The Commission's related 
evaluation of the amendment, finding of emergency circumstances, and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated December 16, 1993.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    Local Public Document Room location: Wilmington Township Public 
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.
    NRC Project Director: James E. Dyer
    Dated at Rockville, Maryland, this 28th of December 1993.
    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects - III/IV/V, Office of 
Nuclear Reactor Regulation
[Doc. 94-53 Filed 1-4-94; 8:45 am]
BILLING CODE 7590-01-F