[Federal Register Volume 59, Number 61 (Wednesday, March 30, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-20330]


[[Page Unknown]]

[Federal Register: March 30, 1994]


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UNITED STATES 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 March 7, 1994, through March 18, 1994. The 
last biweekly notice was published on March 16, 1994.

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 April 29, 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: February 18, 1994
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) Figure 3.2-1, ``REACTOR COOLANT 
COLD LEG vs CORE POWER LEVEL,'' of TS 3/4.2.6, ``REACTOR COOLANT COLD 
LEG TEMPERATURE,'' for Units 1 and 3 to include the cold leg 
temperature between 552 deg.F and 562 deg.F at core power levels 
between 90 percent and 100 percent within the AREA OF ACCEPTABLE 
OPERATION. Also, the proposed amendment would modify TS 3/4.1.1.4, 
``MINIMUM TEMPERATURE FOR CRITICALITY,'' and BASES 3/4.1.1.4, ``MINIMUM 
TEMPERATURE FOR CRITICALITY,'' for all units to allow the minimum 
temperature for criticality to be established at 545 deg.F, rather than 
the current value of 552 deg.F, to establish the surveillance 
temperature at 552 deg.F, rather than the current 557 deg.F, and to 
clarify the BASES for this TS.
    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:
    Standard 1 -- Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    This amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. The 
analyses performed confirmed that the existing safety analysis for 
cycle 5 of all three PVNGS [Palo Verde Nuclear Generating Station] 
units remains valid for a 10 deg.F reduction in RCS [reactor coolant 
system] temperature.
    Standard 2 -- Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    This amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. The 
analyses performed demonstrated that the current licensing basis 
analyses results remain valid with a 10 deg.F reduction in RCS [reactor 
coolant system] temperature, and that the safety system settings remain 
unchanged.
    Standard 3 -- Involve a significant reduction in a margin of 
safety.
    This amendment request will not involve a significant reduction in 
a margin of safety. There is no reduction in the margin of safety since 
the changes apply only to the reactor coolant cold leg temperature and 
the minimum temperature for criticality, the safety analyses have been 
reevaluated (and reperformed where necessary) using the new 
temperature, and the results remain valid. All other safety limits and 
safety system settings remain unchanged. Therefore, there is no 
reduction in any margin of safety.
    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

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

    Date of amendment request: February 4, 1994
    Description of amendment request: The proposed amendment revises 
the Action Statement of Technical Specification 3.6.5, Vacuum Relief 
System, to require that in Modes 1-4 with one vacuum relief system 
inoperable the system be restored to operable status within seventy-two 
hours or be in at least hot standby within the next six hours.
    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 amendment does not physically alter the plant in 
any manner. The proposed amendment does not introduce any new 
equipment nor does it require any existing equipment or systems to 
perform a different type of function than they are currently 
designed to perform. The proposed amendment to Technical 
Specification 3.6.5 allows additional time to restore an inoperable 
containment vacuum relief system to operable status. Changing the 
completion time to seventy-two hours remainsquite conservative for 
this non-ESF system since a seventy-two hour restoration time is 
specified for two-train ESF systems which mitigate Final Safety 
Analysis Report (FSAR) Chapter 15 accidents. The CVRS [containment 
vacuum relief system] is designed to protect the structural 
integrity of containment during an inadvertent actuation of the 
containment spray system, which is not an FSAR Chapter 15 accident. 
Therefore, there would be no 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 amendment does not introduce any new equipment nor 
does it require any existing equipment or systems to perform a 
different type of function than they are currently designed to 
perform. 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 amendment does not involve a significant 
reduction in the margin of safety.
    The proposed amendment to Technical Specification 3.6.5 allows 
additional time to restore an inoperable containment vacuum relief 
system to operable status. Changing the completion time to seventy-
two hours remains conservative since a seventy-two hour restoration 
time is specified for two-train ESF systems which mitigate FSAR 
Chapter 15 accidents. The CVRS is designed to protect the structural 
integrity of containment during an inadvertent actuation of the 
containment spray system, which is not an FSAR Chapter 15 accident. 
Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety as defined in the Technical 
Specifications of FSAR.
    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: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
    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

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: February 24, 1994
    Description of amendment request: The proposed amendments would 
provide surveillance requirements for a planned modification to the 
Keowee emergency power generators' underground power path breaker 
closing logic. The planned modification would provide an automatic 
close feature for the underground path breakers under certain specified 
conditions. The modification is needed to correct a design deficiency 
which resulted in a single failure vulnerability when both Keowee units 
are in their normal alignment. The single failure vulnerability is 
being prevented by means of administrative controls pending 
implementation of a permanent corrective action. The proposed 
amendments would add an annual operability test to Technical 
Specification 4.6, Emergency Power Periodic Testing, of the automatic 
close feature.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    The Keowee Hydro units provide the main source of emergency 
power for the Oconee Nuclear units, but they are not accident 
initiators. The FSAR [Final Safety Analysis Report] Loss of Electric 
Power Accident assumes two types of events: (1) Loss of load (unit 
trip) and (2) Loss of all system and station power. The changes 
performed by the modification that added the automatic closure 
circuitry do not increase the likelihood of either. Also, the 
modifications to the Keowee operating logic will not adversely 
affect the ability to mitigate LOOP [Loss of Offsite Power], LOCA 
[Loss of Coolant Accident], and LOCA/LOOP accidents as described in 
the FSAR. The loss of all station power accident analysis 
assumptions are still valid. This modification has no adverse impact 
on the ability of the Keowee Units to satisfy their design 
requirements to achieving rated speed and voltage within 23 seconds 
of receipt of an emergency start signal.
    The surveillance change that is included in [the] amendment 
request is provided to assure the availability of the electrical 
power systems for mitigation of Design Basis Accidents (DBAs). As 
described within the technical justification [from the licensee's 
application], the Keowee breaker circuitry was modified to allow the 
Keowee Unit that is aligned to the overhead power path to 
automatically close to the underground power path if the postulated 
fault occurs. The surveillance change is an additional restriction 
not presently included in the Technical Specifications. [The] 
amendment will ensure the operability of the Keowee Unit ACB [Air 
Circuit Breaker] automatic close feature and will assure that proper 
testing requirements are maintained.
    Based on the above and the technical justification provided in 
[the amendment application], there is no significant increase in the 
probability of a DBA as a result of this change, nor is there a 
significant increase in the consequences of a DBA as a result of 
this change since the proposed amendment assures the availability of 
the electrical power system.
    (2) Create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated:
    The proposed change makes physical changes to the plant 
configuration. However, the modification simply changes the Keowee 
control logic to remove the possibility of a certain postulated 
failure from causing a loss of emergency power to the Oconee nuclear 
units. The Keowee emergency power systems will remain operable and 
available to mitigate accidents. Operation of ONS [Oconee Nuclear 
Station] in accordance with [the] Technical Specifications will not 
create any failure modes not bounded by previously evaluated 
accidents. Consequently, this change will not create the possibility 
of a new or different kind of accident from any kind of accident 
previously evaluated.
    (3) Involve a significant reduction in a margin of safety:
    Margins of safety associated with [the] Technical Specifications 
have been evaluated. No safety or design limits are adversely 
affected, so margins of safety as defined in the bases to any 
Technical Specifications are not reduced as a result of the Keowee 
modification. The design basis of the auxiliary electrical system is 
to supply the required ES [Engineered Safeguards] loads of one Unit 
and safe shutdown loads of the other two units. The Technical 
Specification amendment includes an additional surveillance 
restriction not presently included in the Technical Specifications. 
The proposed amendment assures the continued availability of the 
electrical power systems; thus preserving the existing margin of 
safety. Therefore, there will be no 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: 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: David B. Matthews, Director

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

    Date of amendment request: September 7, 1993, as supplemented 
February 8, 1994
    Description of amendment request: The proposed amendment would 
revise the Physical Security Plan (PSP).
    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.
    The accident mitigation features of the plant are not affected by 
the proposed compensatory measures for protecting the site during 
periods when security systems are degraded and therefore no decrease 
occurs in the effectiveness of the security program to protect against 
radiological sabotage or increased risk to the public health and 
safety. This is due to continued compliance with existing regulatory 
requirements and other commitments within the security plan. These 
changes have no impact on the design basis security threat and 
accordingly do not create the possibility of a new or different kind of 
accident. New systems, modes of equipment operation, failure modes or 
other plan situations are not introduced by these changes. The proposed 
changes allow flexibility for the use of compensatory measures and do 
not change any safety limits, LCOs, or surveillance requirements on 
equipment to operate the plant.
    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: February 14, 1994
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to reflect changes that have 
been made to 10 CFR Part 20 AND 10 CFR 50.36a.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed revisions to the liquid and gaseous concentration 
release rate limits will not involve a significant increase in the 
probability or consequences of an accident previously evaluated 
because there will be no change in the types and amounts of 
effluents that will be released, nor will there be an increase in 
individual or cumulative occupational radiation exposures.
    The administrative changes for definitions, terminology, 
paragraph references, and record keeping requirements are necessary 
so that the Waterford 3 Technical Specifications will remain 
consistent with the revised federal regulations (i.e., 10CFR20 and 
10CFR50.36a). Record retention and reporting requirements will 
continue to meet NRC regulations. These changes are administrative 
in nature and do not affect plant hardware or operation.
    Restricting access to high radiation areas via guards rather 
than locked doors provides operational flexibility while continuing 
to meet the underlying intent of precluding unauthorized access.
    Therefore, the proposed changes will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    Changes to the liquid and gaseous concentration limits are 
necessary to provide adequate operational flexibility. Operational 
history at Waterford 3 has demonstrated that the use of 
concentration values associated with the old 10CFR20.106 
requirements has resulted in calculated maximum individual doses to 
a member of the public that are small percentages of the limits of 
10CFR50, Appendix I. The proposed revisions will not create the 
possibility of a new or different kind of accident from any 
previously evaluated because the revisions will not change the types 
and amounts of effluent that will be released.
    The administrative changes for definitions, terminology, 
paragraph references, and record keeping are necessary so that the 
Technical Specifications will remain consistent with the revised 
federal regulations (i.e., 10CFR20 and 10CFR50.36a). Record 
retention and reporting requirements will continue to meet NRC 
regulations. These changes are administrative in nature and do not 
affect plant hardware or operation.
    Restricting access for ALARA [as low as reasonably achievable] 
with guards rather than locked doors will continue to meet the 
underlying intent of the TS. These changes do not involve plant 
hardware or operation.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    The proposed revisions do not involve any changes in the types 
or increases in the amounts of effluents released off site. The 
methodology used to control radioactive effluents and calculate 
effluent monitor setpoints will result in the same effluent release 
rate as the current methodology. The basic requirements for TS 
concerning effluent releases (10CFR50.36a) indicate that compliance 
with TS will keep average annual release to small percentages of 
10CFR20 limits. For liquid effluent releases, the annual dose of 500 
mrem, that is the bases for the concentrations in the new 10CFR20. 
The 50.36a requirements further indicate that operational 
flexibility is allowed, compatible with considerations of health and 
safety, which may temporarily result in release higher than such 
small percentages, but still within the limits specified in the old 
10CFR20.106 that references Appendix B maximum permissible 
concentrations (MPCs). For gaseous effluent releases, the limits 
associated with the gaseous release rate TS will be maintained at 
the current instantaneous dose rate limits. Compliance with the 
limits of the new 10CFR20.1301 will be demonstrated by operating 
within the limits of 10CFR50, Appendix I, and 40CFR190. The revision 
will not change the types and amounts of effluent that will be 
released.
    The administrative changes for definitions, terminology, 
paragraph references, and record keeping are necessary so that the 
Technical Specifications will remain consistent with the revised 
federal regulations (i.e., 10CFR20 and 10CFR50.36a). Record 
retention and reporting requirements will continue to meet NRC 
regulations. These changes are administrative in nature and do not 
affect plant hardware or operation.
    Controlling access to high radiation areas for ALARA can be 
performed effectively by guards in place of locked doors. These 
changes do not involve plant hardware or operation.
    Therefore, the proposed changes will not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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

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

    Date of amendment request: February 22, 1994
    Description of amendment request: The proposed amendment modifies 
the minimum stored borated water inventory requirements for Operational 
Modes 1 through 4 by revising Figure 3.1-1 and Limiting Condition for 
Operation (LCO) 3.1.2.8 of the unit Technical Specifications (TS). The 
associated bases for TS 3/4.1.2 are also revised to reflect the 
bounding borated water makeup volumes, as a function of boric acid 
concentration, which define the proposed inventory requirements. The 
proposed amendment will significantly improve operational flexibility 
with no risk to plant safety and will provide for consistency of 
operation between the two St. Lucie units.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Pursuant to 10 CFR 50.92, a determination may be made that a 
proposed license amendment involves no significant hazards 
consideration if operation of the facility in accordance with the 
proposed amendment would not: (1) involve a significant increase in 
the probability or consequences of an accident previously evaluated; 
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. Each standard is discussed as 
follows:
    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendment will reduce the minimum borated water 
inventory required to be stored in the Boric Acid Makeup Tanks 
(BAMT) during unit operation in Modes 1 through 4. The reduction in 
BAMT inventory will not affect any equipment postulated to 
malfunction in the Updated Final Safety Analysis Report (UFSAR) to 
initiate an accident nor will it impact the operation of any other 
equipment whose malfunction could adversely affect safety-related 
structures, systems, or components. Credit is not taken for boron 
addition to the Reactor Coolant System from the BAMTs for purposes 
of reactivity control in accidents analyzed in the UFSAR. The 
minimum required capability to achieve and maintain safe shutdown 
for such events has not been altered. Therefore, operation of the 
facility in accordance with the proposed amendment will not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The reduction in minimum required BAMT inventory does not change 
the boration system function, configuration, operation, or design 
basis as described in the UFSAR. The proposed change does not alter 
the modes of plant operation and does not affect the operation of 
safety-related structures, systems, or components. Therefore, 
operation of the facility in accordance with the proposed amendment 
would not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The reduced BAMT minimum inventory requirements are defined by 
analyses that utilize an approved plant cooldown scenario and 
conservative physics parameters representative of the present and 
future planned reactor core designs for St. Lucie Unit 1. The 
analytical methodology employed to determine the revised inventory 
requirements is the same as that used to establish the existing 
inventory requirements. The existing reactivity control Limiting 
Conditions for Operation (LCO) related to safe shutdown margins and 
redundant boron flow paths have not been altered. Sufficient 
quantities of borated water will continue to be stored in the BAMTs 
to assure compliance with these LCOs during the prescribed plant 
operating modes. Therefore, operation of the facility in accordance 
with the proposed amendment would not involve a significant 
reduction in a margin of safety.
    Based on the discussion presented above and on the supporting 
Evaluation of Proposed TS Changes, FPL has concluded that this 
proposed license amendment involves no significant hazards 
consideration.
    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: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: Harold F. Reis, Esquire, Newman and 
Holtzinger, 1615 L Street, NW, Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: February 18, 1994
    Description of amendment request: The licensee proposes to change 
Turkey Point Units 3 and 4 Technical Specifications (TS) by deleting 
the frequencies specified for audits performed under the cognizance of 
the Company Nuclear Review Board (CNRB). The periodicity of the audits 
for these activities will be controlled as described in the licensee's 
Topical Quality Assurance Report (FPLTQAR), wherein the minimum audit 
frequency for any activity is established as biennial unless the audit 
is otherwise required to be performed more frequently by the TS, Code 
of Federal Regulations, or other licensing commitments. Periodic audits 
of selected aspects of operational phase activities are performed with 
a frequency commensurate with safety significance. During the interval 
between the periodic audits, continuing performance evaluations are 
conducted of activities important to plant safety.
    In addition, the licensee proposes to revise the TS in accordance 
with Generic Letter 93-07. Generic Letter 93-07, ``Modifications of the 
Technical Specifications Administrative Control Requirements for 
Emergency and Security Plans,'' issued December 28, 1993, provided 
guidance for changes to the TS to remove the audit of the emergency and 
security plans and implementing procedures from the list of 
responsibilities of the company nuclear audit and review group. The 
basis of this change is that Parts 50 and 73 of Title 10 of the Code of 
Federal Regulations (10 CFR) include provisions that are sufficient to 
address these requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments relocate the administrative control 
criteria for minimum audit frequencies from the facility TS to the 
FPL Quality Assurance (QA) Program. The QA Program is described in 
the FPL Topical Quality Assurance Report pursuant to 10 CFR 50, 
Appendix B. In addition, the proposed amendments in accordance with 
Generic Letter 93-07, changes the TS to remove the audit of the 
emergency and security plans and implementing procedures from the 
list of responsibilities of the Company Nuclear Review Board. The 
changes being proposed are administrative in nature and do not 
affect assumptions contained in plant safety analyses, the physical 
design and/or operation of the plant, nor do they affect the TS that 
preserve safety analysis assumptions. Therefore, operation of the 
facility in accordance with the proposed amendments would not affect 
the probability or consequences of an accident previously analyzed.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The changes being proposed are administrative in nature and will 
not change the physical plant or the modes of operation defined in 
the Facility License. The change does not involve the addition or 
modification of equipment nor does it alter the design or operation 
of plant systems. Therefore, operation of the facility in accordance 
with the proposed amendments would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The changes being proposed are administrative in nature and do 
not alter the bases for assurance that safety-related activities are 
performed correctly or the basis for any TS that is related to the 
establishment of or maintenance of a safety margin. Therefore, 
operation of the facility in accordance with the proposed amendments 
would not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199
    Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer, 
P.C., 1615 L Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

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, Unit 1, 
Matagorda County, Texas

    Date of amendment request: March 14, 1994
    Description of amendment request: The licensee proposes to make a 
change to the technical specifications to add a new Limiting Condition 
For Operation (LCO), 3.0.6. LCO 3.0.6 will allow equipment removed from 
service or declared inoperable to comply with actions to be returned to 
service, under administrative controls, solely to perform testing. The 
new LCO will permit non-compliance with the applicable Action statement 
to perform the post-maintenance and surveillance testing required to 
demonstrate the operability of the equipment being returned to service 
or the operability of other equipment.
    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 change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The implementation of LCO 3.0.6 will allow the orderly and 
judicious return to service of inoperable equipment. This LCO will 
permit equipment removed from service to comply with required actions 
to be returned to service under administrative controls to verify the 
component or system will perform its safety function. The 
administrative controls will ensure the time involved will be limited 
to only the time required to demonstrate the component or system's 
operability. The implementation of this new LCO will provide an 
acceptable method of testing technical specification equipment prior to 
its return to operable service following required maintenance. These 
actions will ensure that the equipment being returned to service is 
capable of performing its designed safety function prior to being 
declared operable. Therefore, this action will ensure the probability 
or consequences of an accident previously evaluated are not 
significantly increased.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The equipment is only being tested in its designed configuration or 
being returned to service to allow testing of another component or 
system. Therefore, the use of this new LCO will not result in 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.
    The use of the new LCO will only allow the return to service of 
equipment that is expected to operate as designed. The use of the LCO 
will be limited to the performance of testing on the equipment being 
returned to service or on other equipment that is dependent on the 
equipment being returned to service. This testing is limited to post-
maintenance testing and the testing necessary to prove operability. 
Since the equipment will be controlled by administrative requirements 
that will ensure all necessary actions are taken, this change does not 
involve a significant reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: 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 Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of amendment requests: February 22, 1994
    Description of amendment requests: The proposed amendments would 
modify the technical specifications to reduce surveillance requirements 
for testing during power operation. This modification was proposed to 
licensees in NRC Generic Letter 93-05, ``Line Item Technical 
Specifications Improvements to Reduce Surveillance Requirements for 
Testing During Power Operation.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1
    Although the surveillance requirements are lessened by these 
proposed changes, the changes are consistent with those found 
acceptable by the NRC in Generic Letter 93-05. The proposed changes 
have been determined to be compatible with our plant operating 
experience. Based on these considerations, it is concluded that the 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Criterion 2
    The proposed changes do not involve physical changes to the 
plant or changes in plant operating configuration. The changes only 
involve frequency of testing required to be performed. The changes 
are consistent with those found to be acceptable by the NRC in 
Generic Letter 93-05. Thus, it is concluded that the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Criterion 3
    Although the surveillance requirements are lessened by these 
proposed changes, the changes are consistent with those found 
acceptable by the NRC in Generic Letter 93-05. The proposed changes 
have been determined to be compatible with our plant operating 
experience. Based on these considerations, it is concluded that the 
changes do not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Ledyard B. Marsh

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

    Date of amendment request: February 15, 1994
    Description of amendment request: The proposed amendment would 
delete from the Technical Specifications the operational and 
surveillance requirements for the turbine overspeed protection system. 
The licensee intends to continue testing of the overspeed protection 
system as part of plant procedures.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (a) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed amendment does not involve a significant increase 
in the probability or consequences for an accident previously 
evaluated. The proposed deletion of the turbine overspeed protection 
T/S [technical specification] will not significantly change the 
surveillance tests on the Unit 2 turbine. The surveillance schedule 
and tests will be under administrative procedures outside of the TSs 
similar to that of Unit 1 and will be in line with operating 
experience at Cook Nuclear Plant and applicable industry experience. 
The Unit 2 turbine is now operating in its ninth operating cycle 
with over 90,000 hours of operation. Turbine overspeed protection 
surveillance results have been very good since unit startup in 1978. 
In 1983, a wear problem was found with the overspeed plungers. 
Replacement plungers were installed. Then in 1988, these plungers 
were replaced with parts having stellited (hardened) surfaces. There 
have been no subsequent problems. Our expectation is that the 
turbine overspeed protection system will remain available to perform 
its function of preventing excessive turbine overspeed. Lastly, the 
STS [Standard Technical Specifications] developed by the MERITS 
program in NUREG-1431 do not include a T/S for turbine overspeed 
protection. The omission of an overspeed protection T/S in NUREG-
1431 indicates that a T/S is not needed to ensure an adequate level 
of safety for a nuclear facility. This view is supported by WCAP 
11618 which uses the NRC's ``Interim Policy Statement Criteria'' to 
evaluate the need for a turbine overspeed protection T/S and 
concludes that it is not needed. For these reasons, we believe that 
deleting the turbine overspeed protection T/S will not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    (b) Create the possibility of a new or different kind of 
accident from any previously analyzed.
    The proposed amendment does not create the possibility of a new 
or different kind of accident from any previously evaluated. This 
request to delete the turbine overspeed protection T/S eliminates a 
control on the surveillance testing of the Unit 2 turbine. The 
design function of the turbine overspeed protection and the 
operation of the turbine/generator remain the same. The operating 
history of the Unit 2 surveillance results to date and our continued 
testing support the view that the turbine overspeed protection will 
remain available. For these reasons, we believe that the proposed 
changes will not create the possibility of a new or different kind 
of accident from any previously analyzed.
    (c) Involve a significant reduction in a margin of safety.
    The proposed amendment does not involve a significant reduction 
in the margin of safety. Turbine overspeed protection surveillance 
results have been excellent since 1983. The years of operating data 
well within acceptance criteria on Unit 2 turbine overspeed 
protection provide ample evidence that there is no significant 
degradation of the system to perform its function. The reliability 
of the overspeed protection was improved by the replacement of the 
plungers with parts having stellited surfaces. The surveillance 
schedule and tests will be based on operating experience at Cook 
Nuclear Plant and applicable industry experience. Surveillance 
testing will continue under an administrative program outside of 
TSs. Thus the turbine overspeed protection is expected to remain 
available. Also by eliminating this T/S we will be reducing the 
potential for shutting down the unit because of difficulties 
performing this T/S surveillance unrelated to the functionality of 
the valves and overspeed trip protection. Lastly, the STS developed 
by the MERITS program in NUREG-1431 do not include a T/S for turbine 
overspeed protection. The omission of an overspeed protection T/S in 
NUREG-1431 indicates that a T/S is not needed to ensure an adequate 
level of safety for a nuclear facility.
    This view is supported by WCAP 11618 which uses the NRC's 
``Interim Policy Statement Criteria'' to evaluate the need for a 
turbine overspeed protection T/S and concludes that it is not 
needed. For these reasons, we believe that the turbine overspeed 
protection system will remain operable and so this proposed 
amendment does not involve a significant reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document 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: Ledyard B. Marsh

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

    Date of amendment request: February 22, 1994
    Description of amendment request: The proposed amendment would 
revise the reactor coolant system heatup and cooldown curves in the 
Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to the P-T [pressure-temperature] curves 
are being updated as a result of the Unit 2 Capsule U analysis, 
WCAP-13515. The analysis was required per the removal schedule 
established in Table 4.4-5 of the Cook Nuclear Plant Technical 
Specifications. The analysis was performed based on guidance from R/
G 1.99 [Regulatory Guide 1.99, ``Radiation Embrittlement of Reactor 
Vessel Materials''], Revision 2. The change only involves a revised 
time frame for material qualification from 12 EFPY [effective full-
power years] to 15 EFPY as supported by the aforementioned 
Westinghouse analysis. Therefore, we conclude that the changes will 
not involve a significant increase in the probability or 
consequences of a previously evaluated accident, nor will the 
changes involve a significant reduction in a margin of safety.
    (2) Create the possibility of a new or different kind of 
accident from any previously analyzed.
    The proposed changes do not involve any physical modifications 
to the plant. Therefore, the changes should not create the 
possibility of a new or different kind of accident from any 
previously analyzed or evaluated.
    (3) Involve a significant reduction in a margin of safety.
    See the response to (1) above.
    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: 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: Ledyard B. Marsh

Long Island Power Authority, Docket No. 50-322, Shoreham Nuclear 
Power Station, Unit 1 (SNPS), Wading River, New York

    Date of application for amendment: Amendment No. 11, November 4, 
1993 (Reference LSNRC-2115)
    Brief description of amendment: This license amendment request 
(LSNRC-2115) proposes to delete from the Possession-Only License (POL) 
the requirements associated with the safe storage and handling of 
irradiated fuel, the accompanying Appendix A of SNPS Technical 
Specifications, and Appendix B of SNPS Environmental Protection Plan 
(non-radiological). This proposed amendment will update the SNPS POL to 
reflect the status of the facility after irradiated fuel removal from 
the site. SNPS License Condition No. 3 prohibits this amendment from 
being implemented until all the fuel has been removed from SNPS, and 
the licensee has certified to the NRC that all the fuel has been 
removed.
    Basis for the proposed no significant hazards consideration 
determination: In accordance with the requirements and standards in 10 
CFR 50.92(c), the licensee has provided an analysis of the issues 
related to the no significant hazards consideration.
    The licensee's analysis of the issues related to no significant 
hazards consideration are presented below:
    a. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes will become effective after the fuel and 
its related hazards are removed from the site.
    Therefore, the proposed changes will update the SNPS license to 
reflect the facility status after the removal of irradiated fuel. 
This action will not increase the probability or consequences of any 
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 changes will update the license by deleting 
requirements which will no longer apply to SNPS and will not have an 
adverse impact on the operation of the remaining plant systems and 
components.
    Therefore, the proposed change does not create the possibility 
for an accident or malfunction different from any previously 
analyzed.
    c. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change will update the license to reflect the 
status of the facility after the removal of irradiated fuel from the 
site.
    Therefore, the proposed changes will not reduce the margins of 
safety for the remaining plant systems and components.
    The NRC staff has reviewed the licensee's analysis and based on 
this review the three standards of 50.92(c) are satisfied. The NRC 
staff agrees with the licensee's analysis and has determined that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Shoreham Wading River Public 
Library, Shoreham Wading River High School, Route 25A, Shoreham, NY 
11792
    Attorney for licensee: Mr. W. Taylor Reveley, III, Hunton and 
Williams, Riverfront Plaza, East Tower, 951 East Byrd Street, Richmond 
VA 23219-4074
    NRC Branch Chief: John H. Austin

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

    Date of amendment requests: February 14, 1994
    Description of amendment requests: The proposed amendments would 
revise Technical Specifications to reflect the new configuration for 
the Unit 1 480V safeguards bus arrangement (two 480V safeguards buses 
fed by each 4160V safeguards bus). This would make the specifications 
the same for both units since the configuration for the two units will 
become the same during the 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, which is presented below:
    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    SBO/ESU [Station Blackout/Electrical Safeguards Upgrade] Project 
modifications as reflected in the proposed Technical Specifications 
changes were evaluated to determine their impact, if any, on 
potential transients and accidents as described in the Prairie 
Island USAR [Updated Safety Analysis Report]. Each transient and 
accident was evaluated in terms of the mitigating actions described 
or assumed in the USAR analysis. The role of the modified systems in 
mitigating the event was analyzed in order to evaluate whether the 
modification:
    (1) changed, degraded or prevented actions described or assumed 
in the USAR analysis;
    (2) altered any assumptions made in evaluating the radiological 
consequences of the accident;
    (3) played a direct part in mitigating the radiological 
consequences of the accident; or
    (4) affected any fission product barrier.
    The evaluation demonstrated that the USAR transient and accident 
analyses remain valid and bounding.
    As part of the evaluation, the revised emergency diesel 
generator load sequence was analyzed and found to be bounded by the 
existing analyses.
    In particular, the USAR analyses of the loss of offsite power 
(LOOP) event and the large break loss of coolant accident (LBLOCA) 
remain valid and bounding. In addition, the current USAR analysis 
for the radiological consequences of a LBLOCA remains valid.
    Further, the plant response to a loss of AC power event is not 
degraded as a result of these changes but, in fact, is significantly 
improved.
    In order to determine the effect of the modifications upon the 
probability and consequences of an accident, the following items 
were specifically evaluated:
    (1) the applicable design, material and construction standards;
    (2) instrumentation accuracies and response times;
    (3) the equipment operating and design limits, including 
electrical bus loading, emergency diesel generator loading and 
battery loading;
    (4) the system interfaces;
    (5) voltage margins; and
    (6) coordination of protective devices.
    Structures, systems and components involved in the modifications 
were evaluated as follows:
    (1) The design specifications for the new structures, systems 
and components were considered for the following requirements:
    - seismic;
    - separation including control/power circuit interaction, 
redundancy/separation of systems, and isolation between safety and 
non-safety circuits;
    - environmental parameters;
    - severe meteorological events;
    - missiles; and
    - fire protection.
    All structures, systems and components meet the appropriate 
design requirements for their respective classifications.
    (2) Structures, systems and components were additionally 
evaluated for the following:
    - Structural loads were determined for new cable runs in the 
existing plant and for new cable penetrations in the existing 
structures.
    - New electrical loads requirements were determined.
    - System/equipment protection features have been maintained in 
the modification.
    - Support system performance was specified to maintain the 
safety function of the equipment.
    - System/equipment redundancy and independence is maintained.
    - The frequency of operation of existing equipment was evaluated 
and determined not to be affected.
    - The testing requirements imposed on new structures, systems 
and components are in accordance with their safety classification.
    Failures of systems and components involved in the modifications 
were analyzed, and it was determined that all safety functions were 
maintained.
    Required engineered safeguards features loads are accommodated 
with the improved auxiliary electrical systems configuration; and, 
as demonstrated by the performance of a failure modes and affects 
analysis, no single failure will prevent the modified plant from 
performing its required safety function in the event of an accident 
on either unit.
    For the reasons discussed above, the proposed amendment does not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The SBO/ESU Project modifications as reflected in the proposed 
Technical Specifications changes were evaluated to determine if they 
could create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The modifications were evaluated to determine the types of 
accidents which could result from malfunction of the new/modified 
structures, systems and components. It was determined that no new or 
different kinds of accidents from those previously evaluated are 
created. USAR analyses remain bounding.
    For these reasons, 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 will not involve a significant 
reduction in the margin of safety.
    The new Unit 1 480V safeguards configuration provides additional 
circuit breakers for improved motor control center (MCC) feeder 
circuit coordination by eliminating subfed 480V MCCs from safeguards 
480V buses. The proposed Technical Specification changes identify 
the new 480V buses and require the operability of both of the buses 
per train rather than the one bus per train of the current 
configuration and current Technical Specification requirements.
    Since the operability requirements are not decreased nor are the 
allowed out-of-service times increased by the proposed changes, the 
margin of safety is maintained.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: Ledyard B. Marsh

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

    Date of amendment requests: February 16, 1994 (Reference LAR 94-03)
    Description of amendment request: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant Unit Nos. 1 and 2 to revise TS 4.6.1.2, ``Containment 
Integrity.'' The specific TS changes proposed are as follows:
    (1) The requirement to conduct three Type A tests specifically at 
40 plus or minus 10 month intervals during each 10-year service period 
would be replaced with a requirement to conduct three Type A tests at 
approximately equal intervals during each 10-year service period.
    (2) The requirement to conduct the third Type A test of each set 
during the shutdown for the 10-year plant inservice inspection would be 
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:
    a. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes do not affect the initiation of any 
accident, nor do the proposed changes involve modifications to any 
plant equipment.
    The proposed change to the schedule provides flexibility in 
meeting the current requirement for 3 tests in 10 years and is 
consistent with the intent of the 10 CFR 50, Appendix J requirement 
to perform Type A tests at approximately equal intervals. The test 
type and test method used for Type A testing would not be changed. 
The Type A test acceptance criteria would not be changed, and 
containment leakage will continue to be maintained within the 
required limits.
    Elimination of the requirement to perform the third Type A test 
during the shutdown for the 10-year plant ISI does not involve any 
modification to plant equipment or affect the operation or design 
basis of the containment. These surveillances are independent of 
each other and provide assurance of different plant characteristics. 
The Type A tests assure the required leak-tightness of the 
containment to demonstrate compliance with the guidelines of 10 CFR 
100. The 10-year ISI program provides assurance of the integrity of 
plant structures, systems, and components and verifies the 
operational readiness of pumps and valves in accordance with 10 CFR 
50.55a.
    Therefore, the proposed changes do 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 changes do not involve modifications to any 
existing equipment or affect the operation or design basis of the 
containment. The proposed changes do not affect the response of the 
containment during a design basis accident.
    Therefore, the proposed changes do 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 changes to the schedule provide flexibility in 
meeting the Type A testing schedule requirements. These proposed 
changes do not affect or change any limiting conditions for 
operation (LCO) or any other surveillance requirements in the TS and 
the Bases for the surveillance requirement remains unchanged. The 
testing method, acceptance criteria, and bases are not changed and 
still provide assurance that the containment will perform its 
intended function.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Project Director: Theodore R. Quay

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

    Date of amendment request: January 10, 1994
    Description of amendment request: The proposed amendment would 
relocate the seismic monitoring instrumentation Limiting Condition for 
Operation, Surveillance Requirements, and associated tables and Bases 
contained in TS sections 3.3.7.2 and 4.3.7.2 to the Updated Final 
Safety Analysis Report (UFSAR).
    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 function of the seismic monitoring instrumentation system is 
to monitor the magnitude and effect of a seismic event only, and 
cannot initiate or mitigate an accident previously evaluated. 
Furthermore, the proposed TS changes to relocate the seismic 
monitoring instrumentation requirements from TS to the UFSAR are in 
accordance with the criteria for determining those requirements that 
should remain in the TS as defined by the NRC in its final policy 
statement, ``Final Policy Statement on Technical Specifications 
Improvements for Nuclear Power Reactors,'' dated July 22, 1993. The 
seismic monitoring instrumentation LCO, SRs, and associated tables 
and Bases proposed for relocation from TS to the LGS UFSAR will 
continue to be implemented by administrative controls that will 
satisfy the applicable requirements of TS section 6 ``Administrative 
Controls.'' Those requirements include a review of changes to plant 
systems and equipment and to the applicable administrative controls 
in accordance with the provisions of 10CFR50.59.
    Criterion 2 of the July 22, 1993 NRC final policy statement 
states, ``A process variable, design feature, or operating 
restriction that is an initial condition of a Design Basis Accident 
or Transient analysis that either assumes the failure of or presents 
a challenge to the integrity of a fission product barrier.'' The 
seismic monitoring instrumentation system is not a system that 
monitors a process variable that is an initial condition for 
accident or transient analyses. The seismic monitoring 
instrumentation is also not a design feature or an operating 
restriction that is an initial condition of Design Basis Accident or 
transient analyses since it only provides information regarding the 
magnitude of and the plant equipment response to a Design Basis 
earthquake. Therefore, the current LGS seismic monitoring 
instrumentation TS requirements do not meet Criterion 2 of the July 
22, 1993 NRC final policy statement.
    Criterion 3 of the July 22, 1993 NRC final policy statement 
states, ``A structure, system, or component that is part of the 
primary success path and which functions or actuates to mitigate a 
Design Basis Accident or Transient that either assumes the failure 
of or presents a challenge to the integrity of a fission product 
barrier.'' The LGS seismic monitoring instrumentation system does 
not provide a function or actuate in order to mitigate the 
consequences of a Design Basis Accident or transient. Therefore, the 
current LGS seismic monitoring instrumentation TS requirements do 
not meet Criterion 3 of the July 22, 1993 NRC final policy 
statement.
    Criterion 4 of the July 22, 1993 NRC final policy statement 
states, ``A structure, system or component which operating 
experience or probabilistic safety assessment has shown to be 
significant to public health and safety.'' Operating experience has 
shown that the LGS seismic monitoring instrumentation system has no 
impact on public health and safety as defined by the NRC final 
policy statement. Furthermore, LGS specific probabilistic risk 
assessment (PRA) does not credit the seismic monitoring 
instrumentation system as a significant factor in the plant response 
to an accident. Therefore, the current LGS seismic monitoring 
instrumentation TS requirements do not meet Criterion 4 of the July 
22, 1993 NRC final policy statement for determining those 
requirements that should remain in TS. This conclusion is consistent 
with the function of the seismic monitoring instrumentation system 
stated above.
    These proposed TS changes will maintain the current operation, 
maintenance, testing, and system operability controls of the seismic 
monitoring instrumentation system. Furthermore, any future changes 
to the seismic monitoring instrumentation system will be evaluated 
for the effect of those changes on system reliability as required by 
10CFR50.59. The seismic monitoring instrumentation system 
performance will not decrease due to these proposed TS changes and 
the system will continue to be administratively controlled in 
accordance with TS Section 6, including the requirements of 
10CFR50.59, thereby precluding a future decrease in its performance.
    In accordance with the current TS Section 3.3.7.2, with the 
seismic monitoring instrumentation inoperable, the plant would not 
be required to shutdown and the provisions of TS Section 3.0.3 
(i.e., plant shutdown) would not be applicable. Therefore, the 
inoperability of this system and therefore the consequences of an 
accident while this system is inoperable, was previously evaluated 
as not significant enough to require a change to the plant operating 
condition.
    Since the seismic monitoring instrumentation system does not 
monitor a process variable that is an initial condition for an 
accident or transient analyses, or actuates any accident mitigation 
feature, and since the operation, maintenance, testing, and 
modification of the seismic monitoring instrumentation system will 
continue to be administratively controlled, including the 
requirements of 10CFR50.59; therefore, maintaining the reliability 
of the system, 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 function of the seismic monitoring instrumentation system is 
to monitor the magnitude and effect of a seismic event only. The 
proposed TS changes to relocate the seismic monitoring instruments 
requirements from TS to the UFSAR are in accordance with the 
criteria for determining those requirements that should remain in 
the TS as defined by the NRC in its final policy statement, dated 
July 22, 1993. The seismic monitoring instrumentation system does 
not monitor a process variable that is an initial condition for an 
accident or transient analyses. The seismic monitoring 
instrumentation is also not a design feature or an operating 
restriction that is an initial condition of a Design Basis Accident 
or transient analyses since it only provides information regarding 
the magnitude of and the plant equipment response to a Design Basis 
earthquake.
    These proposed TS changes to relocate the TS requirements to the 
UFSAR will not alter the operation of the plant, or the manner in 
which the seismic monitoring instrumentation system will perform its 
function, and any future changes will continue to be 
administratively controlled in accordance with TS Section 6, 
including the requirements of 10CFR50.59.
    These proposed TS changes will not impose new conditions nor 
result in new types of equipment which will result in different 
types of malfunctions of equipment important to safety than any type 
previously evaluated.
    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.
    These proposed TS changes to relocate the seismic monitoring 
instrumentation requirements from TS to the UFSAR are in accordance 
with the criteria for determining those requirements that should 
remain in the TS as defined by the NRC in final policy statement, 
dated July 22, 1993.
    Criterion 1 of the NRC final policy statement states, 
``Installed instrumentation that is used to detect, and indicate in 
the control room, a significant abnormal degradation of the reactor 
coolant pressure boundary.'' The NRC final policy statement explains 
that ''...This criterion is intended to ensure that Technical 
Specifications control those instruments specifically installed to 
detect excessive reactor coolant leakage. This criterion should not, 
however, be interpreted to include instrumentation to detect 
precursors to reactor coolant pressure boundary leakage or 
instrumentation to identify the source of actual leakage (e.g., 
loose parts monitor, seismic instrumentation, valve position 
indicators).'' Based on the above NRC guidance, the LGS UFSAR, and 
TS Bases 3.3.7.2, the seismic monitoring instrumentation does not 
detect, and indicate in the control room, a significant abnormal 
degradation of the reactor coolant pressure boundary. Therefore, the 
current LGS seismic monitoring instrumentation TS requirements do 
not meet Criterion 1. Furthermore, operating experience has shown 
that the LGS seismic instrumentation system has no impact on public 
health and safety as defined by the NRC final policy statement. In 
addition, the LGS specific PRA does not credit the seismic 
monitoring instrumentation system as a significant factor in the 
plant response to accidents.
    The seismic monitoring instrumentation LCO, SRs, and associated 
tables and Bases proposed for relocation to the LGS UFSAR will 
continue to be implemented by administrative controls that will 
satisfy the applicable requirements of TS section 6 ``Administrative 
Controls.'' Those requirements include a review of future changes to 
the system and applicable administrative controls in accordance with 
the provisions of 10CFR50.59.
    Accordingly, based on the above discussion of NRC specific 
guidance, operating experience, and continued imposition of 
administrative controls, 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: Charles L. Miller

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

    Date of amendment request: February 3, 1994
    Description of amendment request: The licensee commenced operating 
on a 24-month fuel cycle, instead of the previous 18-month fuel cycle, 
with fuel cycle 9. Fuel cycle 9 started in August 1992; however, the 
facility shut down in February 1993 for a ``Performance Improvement 
Plan'' outage and a restart date has not yet been established. In order 
to accommodate operation on a 24-month cycle after the facility 
restarts, the licensee requested an amendment to the Technical 
Specifications (TSs) to incorporate the changes listed in items 1-7 
below:
    (1) The licensee proposed changing the calibration frequency for 
the reactor coolant temperature instrument channels (specified in TS 
Table 4.1-1) to accommodate operation on a 24-month cycle.
    (2) The licensee proposed changing the calibration frequency for 
the steam generator level instrument channels (specified in TS Table 
4.1-1) to accommodate operation on a 24-month cycle.
    (3) The licensee proposed changing the calibration frequency for 
the containment pressure instrument channels (specified in TS Table 
4.1-1) to accommodate operation on a 24-month cycle.
    (4) The licensee proposed changing the calibration frequency for 
the steam line pressure instrument channels (specified in TS Table 4.1-
1) to accommodate operation on a 24-month cycle.
    (5) The licensee proposed changing the calibration frequency for 
the turbine first stage pressure instrument channels (specified in TS 
Table 4.1-1) to accommodate operation on a 24-month cycle.
    (6) The licensee proposed changing the calibration frequency for 
the turbine trip low auto stop oil pressure instrument channels 
(specified in TS Table 4.1-1) to accommodate operation on a 24-month 
cycle.
    (7) The licensee proposed changing the calibration frequency for 
the 480V bus undervoltage and alarm relays (specified in TS Table 4.1-
1) to accommodate operation on a 24-month cycle.
    These proposed changes follow the guidance provided in Generic 
Letter 91-04, ``Changes in Technical Specification Surveillance 
Intervals to Accommodate a 24-Month Fuel Cycle,'' as applicable.
    The licensee also requested the following additional changes:
    (1) The addition to TS Table 3.5-5 of limiting conditions for 
operation (LCO) requirements for a wide range containment pressure 
variable.
    (2) The addition of a quarterly functional test surveillance 
requirement to Item 4 of TS Table 4.1-1 for the low average temperature 
actuation circuits of the reactor coolant temperature channels.
    (3) The addition of a second line to Item 14 of TS Table 4.1-1 to 
specify surveillance requirements for the wide range containment 
pressure instrumentation.
    (4) The revision of Item 20 to TS Table 4.1-1 to clarify that both 
the reactor trip and the engineered safety features (ESF) actuation 
relay logic channels are functionally tested.
    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:
    Consistent with the criteria of 10 CFR 50.92, the enclosed 
application is judged to involve no signicant hazards based on the 
following information:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of any accident 
previously evaluated?
    Response:
    The proposed changes do not involve a significant increase in 
the probability or consequences of any accident previously 
evaluated. The proposed changes extend the calibration intervals 
(given in [TS] Table 4.1-1) for the reactor coolant loop temperature 
instrumentation used for Engineered Safety Features Actuation 
Systems (ESFAS) and Post Accident Monitoring (PAM) functions, the 
steam generator (SG) level instrumentation used for ESFAS and PAM 
functions, the containment pressure instrumentation used for ESFAS 
and PAM functions, the steam line pressure instrumentation used for 
ESFAS and PAM functions, the turbine first stage pressure 
instrumentation used for ESFAS functions, the 480V bus undervoltage 
and alarm relays used for ESFAS functions, and the turbine trip low 
auto stop oil pressure instrumentation. These changes are being made 
to accommodate a 24 month operating cycle. Other changes include: 1) 
the addition to [TS] Table 3.5-5 of limiting conditions for 
operation (LCO) requirements for the wide range containment pressure 
channels; 2) the addition of a quarterly functional test 
surveillance requirement to Item 4 of [TS] Table 4.1-1 for the low 
Tavg [average temperature] actuation circuits of the reactor 
coolant temperature channels; 3) the addition of a second line to 
Item 14 of [TS] Table 4.1-1 to specify the surveillance requirements 
for the wide range containment pressure channels; and 4) the 
revision of Item 20 to [TS] Table 4.1-1.
    Extension of the calibration intervals in question were 
evaluated and the results documented in the ESFAS and Indicating 
Instrument Surveillance Test Extension reports (References 7 and 8 
[Engineered Safety Features Actuation Systems Surveillance Test 
Extensions, NYPA document IP3-RPT-ESS-00400, dated May 10, 1993 and 
Indicating Instruments Surveillance Test Extensions, NYPA document 
IP3-RPT-MULTI-00424, dated May 5, 1993]). ESFAS and indicating 
instrument drift analyses were performed to evaluate actual past and 
projected future instrument drift. Revised safety system loop 
accuracy/setpoint calculations, which include any additional 
instrument uncertainties resulting from the proposed calibration 
interval extensions, show that sufficient margin exists between the 
analytical and field trip settings for the low Tavg, the SG 
low-low level, the high and high-high containment pressure, the high 
differential steam line pressure, the low steam line pressure, the 
high steam flow (dependent upon turbine first stage pressure), the 
turbine trip low auto stop oil pressure, and the 480V bus 
undervoltage trip functions. Safety analyses are not affected. 
Additionally, postulated uncertainties associated with the extended 
calibration intervals for the wide range reactor coolant loop 
temperature, the narrow and wide range SG level, and the steam line 
pressure instrumentation will be accomodated by changes to the 
Emergency Operating Procedure (EOP) settings. Extension of the 
calibration interval for the narrow range containment pressure 
instrumentation channels does not affect EOP settings. Safety 
analyses are not affected by the EOP setting changes.
    The results of the changes to: 1) add to [TS] Table 3.5-5 LCO 
requirements for the wide range containment pressure channels, 2) 
add a quarterly functional test surveillance requirement to Item 4 
of [TS] Table 4.1-1 for the low Tavg actuation circuits of the 
reactor coolant temperature channels, and 3) add a monthly channel 
check surveillance requirement to [TS] Table 4.1-1 for the wide 
range containment pressure channels are consistent with Westinghouse 
Standard Technical Specifications (W STS - Reference 12 [NUREG-1431, 
Revision O, ``Standard Technical Specifications - Westinghouse 
Plant,'' dated September 28, 1992]). The addition of LCO 
requirements to [TS] Table 3.5-5 for the wide range containment 
pressure instrumentation, the addition of a quarterly functional 
test requirement to Item 4 of [TS] Table 4.1-1 for the low Tavg 
actuation circuits, and the separation of surveillance requirements 
for the narrow and wide range containment pressure instrumentation 
into two lines on [TS] Table 4.1-1 consitute additional technical 
specification controls. Changes which consitute additional technical 
specification limitations and controls are classified by Federal 
Register dated April 6, 1983 (48 FR 14870, April 6, 1983) as not 
likely to involve significant hazards considerations. The change to 
[TS] Table 3.5-5 ensures conistency with the Authority's commitment 
to Regulatory Guide (RG) 1.97 [``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident''] for the containment pressure 
variable.
    The current surveillance requirement specified by Item 20 has 
been interpreted by Indian Point 3 as including on-line testing of 
both the reactor trip and engineered safety features (ESF) actuation 
logic channels, but since the wording may be confusing, this 
application proposes to change the wording to clarify that both the 
reactor trip and the ESF actuation logic channels are functionally 
tested at least every two months on a staggered basis (i.e., one 
train per month). The change is consistent with W STS and only 
involves a wording change which strengthens the Technical 
Specification requirement. The change does not involve hardware, 
procedural, or operational changes, and, therefore, does not affect 
safety analyses.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any previously 
evaluated?
    Response:
    The proposed changes do not create the possibility of a new or 
different kind of accident from any previously evaluated. Extension 
of the calibration intervals in question were evaluated and the 
results documented in the ESFAS and Indication Instrument 
Surveillance Test Extension reports. ESFAS and indicating instrument 
drift analyses were performed to evaluate actual past and projected 
future instrument drift. Revised safety system loop accuracy/
setpoint calculations and EOP setting calculations show that, 
although some EOP setting changes will be made to accommodate 
postulated drift associated with the extended calibration intervals, 
safety analyses are not affected.
    The changes to 1) specify LCO and surveillance requirements for 
the wide range containment pressure instrumentation channels, 2) add 
a quarterly functional test surveillance requirement to Item 4 of 
[TS] Table 4.1-1 for the low Tavg actuation circuits of the 
reactor coolant temperature channels, and 3) clarify that both 
reactor trip and ESF actuation logic channels are functionally 
tested constitute additional technical specification limitation and 
controls. Additionally, these changes are consistent with W STS.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response:
    The proposed changes do not involve significant reductions in 
margins of safety. Loop accuracy/setpoint calculations show that 
sufficient margin exists between the analytical and field trip 
settings for the low Tavg, the SG low-low level, the high and 
high-high containment pressure, the high differential steam line 
pressure, the low steam line pressure, the high steam flow 
(dependent upon turbine first stage pressure), the turbine trip low 
auto stop oil pressure, and the 480V bus undervoltage trip functions 
to accommodate postulated uncertainties associated with the extended 
calibration intervals. And, although changes to EOP settings will be 
made to accommodate the postulated uncertainties associated with the 
extended calibration intervals for the wide range reactor coolant 
loop temperature, the narrow and wide range SG level, and the steam 
line pressure instrumentation, the EOP setting changes do not in any 
way adversely affect the analytical limits established by safety 
analyses.
    Extension of the calibration intervals in question do not affect 
safety analyses. The other changes being made in this application 
involve additional technical specification limitations and controls 
and are consistent with W STS. None of the changes involve 
significant reductions in margins of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle, 
New York, New York 10019.
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: March 4, 1994
    Description of amendment request: The proposed change would modify 
Sections 5.3 and 5.6 of the Technical Specifications (TSs) to allow the 
use of Westinghouse Vantage+ fuel with ZIRLO cladding. The present TSs 
require the fuel rod cladding to be Zircaloy-4, which is used in the 
Westinghouse Standard and Vantage 5H fuel designs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed changes to Technical Specifications 5.3.1 and 5.6.1 
for Salem Generating Station (SGS) Unit Nos. 1 and 2:
    1. do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The fuel cladding design criteria for SGS would remain the same 
for ZIRLO clad fuel as it is for Zircaloy-4 clad fuel. All fuel 
design and performance criteria will continue to be met using NRC-
approved methods and no new single failure mechanisms will be 
introduced. The use of ZIRLO clad fuel does not introduce any 
changes to plant equipment or operation that would adversely affect 
accident initiators or precursors. The proposed changes would not 
result in any changes to compliance with licensing basis safety 
limits.
    2. do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed changes would require that NRC[-]approved methods 
be used in fuel assembly design. No new operating configurations 
potentially resulting in the occurrence of a previously unanalyzed 
event would be allowed by the proposed change.
    3. do not involve a significant reduction in a margin of safety.
    The proposed change would continue to require that NRC[-
]approved methods are used to ensure compliance with the fuel design 
and safety limits which ensure that an acceptable margin of safety 
is maintained relative to fuel assembly design.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, New Jersey 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: Charles L. Miller

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 amendment request: March 1, 1993
    Description of amendment request: The proposed amendment would 
clarify Technical Specification 3.6.1.2, Primary Containment Leakage, 
and revise the ``as-found'' value of the overall integrated primary 
containment leakage rate which is used when determining the test 
schedule for future Type A tests within Surveillance Requirement 
4.6.1.2.b. This amendment also requests an exemption from the 
requirements of 10 CFR 50 Appendix J, Primary Reactor Containment 
Leakage Testing for Water-Cooled Power Reactors.
    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.
    These proposed changes clarify Technical Specification 3.6.1.2 
by providing a more definitive action to take if the leakage rate 
limit(s) specified in the LCO are not being met. The current Action 
is not clear on what actions are necessary if the leakage rate 
limits (e.g., Type B and C limits) are known to be exceeded while 
the reactor coolant system (RCS) temperature is above 200 deg. F, 
which has caused compliance difficulties. The revised Action is 
modeled after the one in the Primary Containment Integrity 
Specification, which (through the definition of Primary Containment 
Integrity) includes a provision that the containment leakage rates 
be in compliance with the requirements of Specification 3.6.1.2.
    Surveillance Requirement 4.6.1.2.b has been revised to reflect 
the actual plant design basis leakage rate of La as the value 
against which the ``as-found'' Type A test results are compared when 
determining the test schedule for future Type A tests. The 
probability of exceeding the maximum allowable leakage rate, 
La, is not significantly increased since the ``as-left'' 
leakage rate requirement of 0.75 La (which must be met during 
startup from any outage in which a Type A test has been performed) 
is still imposed through LCO 3.6.1.2.a, Action 3.6.1.2.a and 
Surveillance Requirement 4.6.1.2.a. The Applicability of 
Specification 3.6.1.2 has been modified to resolve an existing 
conflict with the current Action, which requires that a reactor 
coolant system temperature of 200 deg. F not be exceeded with a 
leakage rate greater than 0.75 La (during startups from outages 
in which a Type A ILRT has been performed). With the modified 
Applicability and the retained LCO requirement for the ``as-left'' 
leakage rate to be less than or equal to 0.75 La, the 
requirement of the current Action (not to exceed to 200 deg. F) is 
implicitly maintained, due to the provisions contained within 
specification 3.0.4. This maintains the same margin for degradation 
between performances of the periodic Type A tests as is provided in 
the current specification. Since the analysis leakage limit of 
La has not changed, the offsite radiological consequences of an 
accident assumed in the safety analyses have not been affected.
    The deletion of the current link between Specifications 3.6.1.2 
and 3.10.1 is an administrative change only, made because the two 
Specifications no longer overlap and the link is therefore 
unnecessary.
    In summary, there is no change in the probability or 
consequences of any accident since the clarifications of the 
existing LCO, Applicability, Actions, Surveillance Requirements and 
the revised ``as-found'' acceptance criterion do not change the 
design of the plant, nor the operational characteristics of any 
plant system, nor the procedures by which the Operators run the 
plant.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed Action to address situations when the leakage rate 
limit(s) cannot be met in Operational Conditions 1, 2 and 3, with 
the reactor coolant system temperature greater than 200 deg. F, does 
not create the possibility of a new or different kind of event - it 
only provides the measures to be taken following determination of 
increased containment leakage. The clarification to the existing 
Applicability simply resolves an existing conflict between the 
Applicability and the Action, and ensures that the same requirements 
that were contained within the former Action are maintained 
following implementation of the change, by preventing plant startup 
above a RCS temperature of 200 deg. F (following an outage in which 
a Type A test has been performed), unless the leakage rate is below 
the 0.75 La test acceptance criterion. Additional changes are 
being made to clarify the application of Appendix J requirements.
    Revising the ``as-found'' value of La does not create the 
possibility of a new or different kind of event - since the analysis 
limit value, La, has not been increased and no new mode of 
operation has been introduced.
    In summary, the proposed changes do not create the possibility 
of a new or different kind of accident, since no design changes are 
being made that would create a new type of accident or malfunction, 
and the method and manner of plant operation remains unchanged.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed Action simply imposes a more definitive action to 
take when a leakage rate limit(s) is exceeded, consistent with the 
Primary Containment Integrity Specification. The changes to the 
Surveillance Requirements to reflect the ``as-found'' value of 
La are consistent with the intent of the requirements specified 
in Appendix J, and similar requirements have been provided for other 
plants. The current requirement for ``as-left'' leakage rates to be 
less than or equal to 0.75 La before increasing the reactor 
coolant system temperature above 200 deg. F from outages in which a 
Type A ILRT has been performed has been retained since the proposed 
Action now includes a shutdown requirement, and in accordance with 
Technical Specification 3.0.4, ``Entry into an OPERATIONAL CONDITION 
or other specified condition shall not be made when the conditions 
for the LCO are not met and the associated ACTION requires a 
shutdown if they are not met within a specified time interval.'' 
Since the new Action includes a shutdown provision and the LCO 
retains the current limit of 0.75 La, a change into the new 
Applicability of Specification 3.6.1.2 cannot occur if 0.75 La 
is exceeded. This ensures that the same margin as currently exists 
today is maintained for possible degradation between performance of 
the periodic Type A tests. The other changes are clarifications and 
are administrative in nature. Therefore, the proposed changes do not 
involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, D.C. 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment request: February 10, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification Table 2.2-1 and Bases Section 2.2.1. The 
Functional Unit 14 of Table 2.2-1 would be revised to correct the Total 
Allowance, reflecting the undervoltage relay span and to correct the 
Allowable Value, reflecting the rack measurement and test equipment 
(M&TE) uncertainty. The Bases would be revised to clarify the 
relationship between the Trip Setpoint and Allowable Value, expressed 
in voltage, and the Total Allowance, Z and S values, expressed in 
percent of the undervoltage relay span (calibrated span of 70-100 
volts).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed changes do not involve a significant hazards 
consideration because operation of the Callaway Plant with these 
changes would not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Overall protection system performance will remain within the 
bounds of the accident analyses documented in FSAR Chapter 15, WCAP-
10961-P, and WCAP-11883 since no hardware changes are proposed.
    The RCP undervoltage reactor trip function is a primary trip 
function and is credited in FSAR Section 15.3.2, Complete Loss of 
Forced Reactor Coolant Flow. The trip setpoint is designed to ensure 
plant operation within the DNB design basis. There will be no effect 
on this analysis, or any other accident since the safety analysis 
limit and trip response time are unaffected and remain the same as 
discussed in FSAR Section 15.0.6 and FSAR Table 15.0.4.
    The RCP undervoltage reactor trip will continue to function in a 
manner consistent with the above analysis assumptions and the plant 
design basis. As such, there will be no degradation in the 
performance of nor an increase in the number of challenges to 
equipment assumed to function during an accident situation.
    These Technical Specification revisions do not involve any 
hardware changes nor do they affect the probability of any event 
initiators. There will be no change to normal plant operating 
parameters, ESF actuation setpoints, accident mitigation 
capabilities, accident analysis assumptions or inputs. Therefore, 
these changes will not increase the probability of an accident 
previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any previously evaluated.
    As discussed above, there are no hardware changes associated 
with these Technical Specification revisions nor are there any 
changes in the method by which any safety-related plant system 
performs its safety function.
    Changes to the Total Allowance and Allowable Value terms in 
Technical Specification Table 2.2-1 will require only minor changes 
to the acceptance criteria sections of a few surveillance 
procedures. The normal manner of plant operation is unaffected. If 
an undervoltage relay setpoint is found to be below the nominal trip 
setpoint in Table 2.2-1, entry into Action Statements a or b of 
Specification 2.2.1 will be affected insofar as the Allowable Value 
is being lowered and the Total Allowance value contained in Equation 
2.2-1 is being raised. However, the nominal trip setpoint is 
unchanged and the required plant condition for exiting the Action 
Statements, i.e. adjusting the trip setpoint consistent with the 
Table 2.2-1 value, is likewise unchanged. The revisions to the Total 
Allowance and Allowable Value correct errors in their derivation and 
were calculated using the previously approved Westinghouse setpoint 
methodology. The setpoint equations cited in that methodology are 
unchanged; however, inputs to those equations have been revised to 
reflect the undervoltage relay span and the rack M&TE uncertainty.
    No new accident scenarios, transient precursors, failure 
mechanism, or limiting single failures are introduced as a result of 
these changes. There will be no adverse effect or challenges imposed 
on any safety-related system as a result of these changes. 
Therefore, the possibility of a new or different type of accident is 
not created.
    (3) Involve a significant reduction in a margin of safety.
    There will be no change to the DNBR Correlation Limit, the 
design DNBR limits, or the safety analysis DNBR limits discussed in 
Bases Section 2.1.1.
    As discussed previously, the response time of the RCP 
undervoltage reactor trip function will remain within the 
assumptions used in the accident analyses. The analysis of the 
complete loss of flow accident will remain as presented in FSAR 
Section 15.3.2.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. There will be no impact on DNBR limits, 
FQ, F-delta-H, LOCA PCT, peak local power density, or any other 
margin of safety. The safety analysis limit, 9384 Vac at the RCP 
motor, and the nominal trip setpoint, 10,584 Vac, remain the same as 
before.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
    NRC Project Director: John N. Hannon

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

    Date of amendment request: February 17, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specifications 3/4.5.1 and Bases Section 3/4.5.1. A 
new Action Statement a. would be added to Specification 3.5.1 to 
provide a 72 hour allowed outage time (AOT) for one accumulator 
inoperable due to its boron concentration not meeting the 2300-2500 ppm 
band. The AOT for Action Statement b. would be changed to 24 hours in 
lieu of the current AOT of 1 hour. Surveillances 4.5.1.1.a.1) and 
4.5.1.1.b would be revised and Surveillance 4.5.1.2 would be deleted 
per the guidance of NRC Generic Letter 93-05. Bases Section 3/4.5.1 
would be revised to discuss the 72 hour and 24 hour AOTs for Action 
Statements a. and b. above.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed changes to the Technical Specifications do not 
involve a significant hazards consideration because operation of 
Callaway Plant in accordance with these changes would not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Overall protection system performance will remain within the 
bounds of the accident analyses documented in FSAR Chapter 15, WCAP-
10961-P, and WCAP-11883 since no hardware changes are proposed.
    The safety injection (SI) accumulators are credited in FSAR 
Section 15.6.5 for large and small break LOCA. There will be no 
effect on these analyses, or any other accident analysis, since the 
analysis assumptions are unaffected and remain the same as discussed 
in FSAR Section 15.6.5. Design basis accidents are not assumed to 
occur during allowed outage times covered by the Technical 
Specifications. As such, the ECCS Evaluation Model equipment 
availability assumptions made in FSAR Section 15.6.5 remain valid.
    The SI accumulators will continue to function in a manner 
consistent with the above analysis assumptions and the plant design 
basis. As such, there will be no degradation in the performance of 
nor an increase in the number of challenges to equipment assumed to 
function during an accident situation.
    These Technical Specification revisions do not involve any 
hardware changes nor do they affect the probability of any event 
initiators. There will be no change to normal plant operating 
parameters, ESF actuation setpoints, accident mitigation 
capabilities, accident analysis assumptions or inputs. The effect on 
the Callaway core damage frequency has been quantified as 
insignificant. Therefore, these changes will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any previously evaluated.
    As discussed above, there are no hardware changes associated 
with these Technical Specification revisions nor are there any 
changes in the method by which any safety-related plant system 
performs its safety function. The normal manner of plant operation 
is unaffected.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of these changes. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of these changes. 
Therefore, the possibility of a new or different type of accident is 
not created.
    (3) Involve a significant reduction in a margin of safety.
    There will be no change to the DNBR Correlation Limit, the 
design DNBR limits, or the safety analysis DNBR limits discussed in 
Bases Section 2.1.1.
    As discussed previously, the performance of the SI accumulators 
will remain within the assumptions used in the large and small break 
LOCA analyses, as presented in FSAR Section 15.6.5.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. There will be no impact on DNBR limits, 
FQ, F-delta-H, LOCA PCT, peak local power density, or any other 
margin of safety.
    Based upon the preceding information, it has been determined 
that the proposed changes to the Technical Specifications do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated, create the possibility of a new or 
different kind of accident from any accident previously evaluated, 
or involve a significant reduction in a margin of safety. Therefore, 
it is concluded that the proposed changes meet the requirements of 
10CFR50.92(c) and do not involve a significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
    NRC Project Director: John N. Hannon

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: March 1, 1994
    Description of amendment request: The proposed change would revise 
the Technical Specifications (TS) for the North Anna Power Station, 
Units No. 1 and No. 2 (NA-1&2). Specifically, the change would 
eliminate certain surveillance requirements for the emergency diesel 
generators which have been determined to be unnecessary.
    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, ``Line-
Item Technical Specifications Improvements to Reduce Surveillance 
Requirements for Testing During Power Operation,'' dated September 27, 
1993.
    The safety function of the Emergency Diesel Generators (EDGs) is to 
supply AC electrical power to plant safety systems whenever the 
preferred AC power supply is unavailable. Consistent with Generic 
Letter 93-05, Item 10.1 and NUREG-1366, the licensee is requesting a 
change to the testing requirements of an operable EDG when the 
alternate safety buses' EDG is inoperable or an offsite circuit is 
inoperable, the separation of the hot restart test of an EDG from the 
24 hour loaded run, and the elimination of fast loading of EDGs except 
for the 18 month surveillance test of the Loss of Offsite Power (LOOP) 
capability.
    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.
    Modifying the operability testing requirements for an inoperable 
EDG or inoperable offsite AC source(s), gradual loading of EDGs 
during surveillance testing, and separating the hot restart test of 
an EDG from the 24 hour load run test of EDGs does not affect the 
probability of occurrence or consequences of any previously 
evaluated accidents. Surveillance testing of the EDG in accordance 
with Revision 2 of Regulatory Guide 1.9 (December 1979) will 
continue to ensure that the EDGs will be capable of performing their 
intended safety functions. Therefore, modifying the operability 
testing requirements for an inoperable EDG or inoperable offsite AC 
source(s), gradual loading of EDGs during surveillance testing, and 
separating the hot restart test of an EDG from the 24 hour load run 
test of EDGs 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.
    Modifying the operability testing requirements for an inoperable 
EDG or inoperable offsite AC source(s), gradual loading of EDGs 
during surveillance testing, and separating the hot restart test of 
an EDG from the 24 hours load run test of EDGs does not involve any 
physical modifications of the plant or result in a change in a 
method of operation. Surveillance testing of the EDG in accordance 
with Revision 2 of Regulatory Guide 1.9 (December 1979) will 
continue to ensure that the EDGs will be capable of performing their 
intended safety functions. Therefore, a new or different type of 
accident is not made possible.
    (3) Involve a significant reduction in a margin of safety.
    Modifying the operability testing requirements for an inoperable 
EDG or inoperable offsite AC source(s), gradual loading of EDGs 
during surveillance testing, and separating the hot restart test of 
an EDG from the 24 hour load run test of EDGs does not affect any 
safety limits or limiting safety systems settings. System operating 
parameters are unaffected. The availability of equipment required to 
mitigate or assess the consequences of an accident is not reduced. 
Surveillance testing of the EDG in accordance with Revision 2 of 
Regulatory Guide 1.9 (December 1979) will continue to assure that 
the EDGs will be capable of performing their intended safety 
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

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington

    Date of amendment request: May 10, 1993
    Description of amendment request: The amendment proposes to modify 
the Technical Specifications (TS) to incorporate new power to flow 
limits based on core power stability calculations performed for Cycle 
9. In addition, the proposed amendment would clarify the maximum 
measured decay ration permitted during operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's evaluation of 
the licensee's analysis is presented below:
    1. Does the amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change related to the instability regions on the power 
to flow map is based on new calculations using a new code (STAIF) while 
maintaining a decay ration of 0.9 or less as required in IEB 88-07. The 
closer the operators come to a decay ration of 1.0, the closer the core 
comes to potential core power instabilities. By ensuring that the decay 
ratio is maintained below 0.9, the operators reduce the likelihood of 
core power instabilities. The result of the revised calculations using 
this new code is that the restricted regions are expanded over the 
regions contained in the current TS. This increase in restricted 
regions results in plant operation further from potential core power 
instabilities compared to the restricted regions in the current TS, 
resulting in a decreased probability of core power oscillations. The 
power to flow map regions are operating restrictions that, for the core 
power oscillation restricted regions, are intended to reduce the 
likelihood of the onset of oscillations. The core power oscillation 
restricted regions on the power to flow map do not contribute to any 
mitigative actions or plant response after a power oscillation occurs, 
thus the proposed change does not change the consequences of any 
accidents previously evaluated.
    The proposed amendment would also change the wording of the 
technical specifications to clarify that action must be taken to reduce 
the measured decay ration if any two neutron signals of ``greater than 
or equal to 0.75,'' as opposed to the current wording ``greater than 
.75,'' are measured. This would not have any measurable effect on the 
implementation of the affected TS, and would, if anything, result in 
action being taken at a maximum lower value than the current TS. This 
proposed amendment would not, therefore, involve a change in the 
probability or consequences of an accident previously evaluated.
    2. Does the amendment create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change; modifies existing restrictions on the power to 
flow map, and does not involve any modifications to plant systems or 
components or the manner in which they are operated.
    Changing the wording of the TS to require that action be taken to 
reduce the measured decay ration if any two neutron signals of 
``greater than or equal to 0.75,'' as opposed to the current wording 
``greater than .75,'' are measured, does not involve any modifications 
to plant systems or components or the manner in which they are 
operated.
    Based on these considerations, this does not create or increase the 
possibility of a new or different kind of accident.
    3. Does the amendment involve a significant reduction in a margin 
of safety?
    The margin of safety related to the proposed TS change is the core 
power vs. core flow restrictions on the power to flow map. These 
restrictions are currently based on maintaining a decay ratio less than 
0.9, which provides a margin of at least a decay ration of 0.1 from 
what is defined as a decay ration (1.0) that would result in an 
unstable core. Since the revised curves are based on ensuring decay 
rations of less than 0.9 are maintained, the existing margin of safety 
is maintained.
    Changing the wording of the TS to require that action be taken to 
reduce the measured decay ratio if any two neutron signals of ``greater 
than or equal to 0.75,'' as opposed to the current wording ``greater 
than .75,'' are measured, would not have any measurable effect on the 
implementation of the affected TS, and would, if anything, result in 
action being taken at a maximum lower value than the current TS. This 
would not have any significant impact on how close the plant was 
allowed to operate to potential core power instability, and would not, 
therefore, have a significant effect ont he margin of safety related to 
the proposed TS.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Attorney for licensee: Nicholas S. Reynolds, Esq., Winston & 
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: Theodore R. Quay

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington

    Date of amendment request: July 29, 1993, with supplemental 
information provided March 11, 1994 and March 17, 1994
    Description of amendment request: The amendment proposes to modify 
the Technical Specifications (TS) to reflect a new refueling platform. 
Specifically, the amendment would add new values for protective 
features in the TS to reflect the new refueling platform. Values for 
the old refueling platform are retained in the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The staff's evaluation of the licensee's analysis is 
presented below:
    1. Does the amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The only accident evaluation affected by the proposed changes are 
those associated with the Fuel Handling Accident (FHA) analyses 
presented in WNP-2 Final Safety Analysis Report (FSAR) section 15.7.4. 
As discussed therein, the fuel handling accident event that produces 
the largest number of failed spent fuel rods is the drop of a spent 
fuel bundle into the reactor core when the reactor vessel head is off. 
The probability of dropping a spent fuel assembly onto other fuel 
assemblies in the reactor vessel does not increase with the new design. 
The NF500 mast functions identically to the old mast when grappling, 
lifting, or moving a fuel assembly. It does not degrade platform design 
features such as grapple fail-safe on loss of air, dual lifting cables, 
backup cable reel brake, and the grapple engaged loaded interlock, all 
of which serve to protect against a fuel drop event. The new mast is 
more rigid than the previous mast design and, therefore, is less prone 
to mast bowing. The consequences of dropping a fuel assembly are also 
unaffected because the weight of the mast is not considered in existing 
FHA analysis. The number of postulated fuel pins which fail as a result 
of the FHA is unaffected since the energy imparted by the dropped 
assembly is independent of the mast design, and mitigating systems will 
function as previously analyzed. Further, analysis by GE of a 
postulated accident in which the exposed portion of the NF500 mast is 
struck by a missile and severed while lifting a fuel bundle with both 
falling onto the top of the core has been conducted, showing that the 
consequences of the increased weight of the mast and bundle are bounded 
by the current WNP-2 FSAR analysis for the fuel bundle only FHA. 
Retaining the ability to use the old mast does not introduce any 
changes to the current TS that reflect the analysis of the old mast. 
The proposed change would not, therefore, significantly increase the 
probability or consequences of a previously analyzed accident.
    2. Does the amendment create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No new failure modes are introduced as a result of the proposed 
changes. The NF500 mast in intended as an exact replacement for the 
currently installed mast, and is designed to match or exceed the 
strength and performance of the NF400 mast in all areas. No new fuel 
handling methods or surveillance procedures will be necessary as a 
result of installation of the new mast. The proposed change does not 
affect the manner in which protective interlocks operate. Limits on 
fuel travel in all directions are unchanged. Retaining the ability to 
use the NF400 mast presents no new accident possibilities since no 
changes in fuel mast operation would result from use of the existing 
mast. The proposed change would not, therefore, create the possibility 
of a new or different kind of accident from any previously analyzed.
    3. Does the amendment involve a significant reduction in a margin 
of safety?
    The changed refueling mast cutoff and interlock values account for 
the increase weight of the mast, or a portion thereof, and do not 
affect the margins related to the fuel bundle drop analyses. The new 
mast has the same single failure protection as the old mast. The 
proposed change would not, therefore, involve a significant reduction 
in a margin of safety.
    The NRC staff has determined that 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: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Attorney for licensee: Nicholas S. Reynolds, Esq., Winston & 
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: Theodore R. Quay

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington

    Date of amendment request: December 20, 1993
    Description of amendment request: The amendment proposes to modify 
the Technical Specifications (TS) to address new containment purge and 
vent valves to be installed in the 1994 refueling outage. The TS are 
being modified to remove the requirement to ensure the remaining 
existing-design valves' position remains at less than or equal to 
70 deg. because the valves have a permanently installed mechanical stop 
to limit the open position to ensure adequate closure times. In 
addition, this modification is being requested because the current TS 
are too limiting for the new valves, which are designed to close from a 
90 deg. open position. The TS are also being modified to change the 
containment leak testing requirements for the new valves from 6 months 
to 2 years, to reflect the improved seat design of the replacement 
valves. Additional administrative changes are proposed to delete an 
out-of-date note, and to relocate an action statement requirement from 
surveillance section of the TS to the action statements section.
    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. Does the amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Regarding the removal of the requirement to ensure the remaining 
existing-design valves' position remains at less than or equal to 
70 deg.: The maximum open position of the containment purge and vent 
valves is not one of the initiating events for any previously evaluated 
accident in the WNP-2 FSAR. Thus the proposed change will not affect 
the probability of an accident previously evaluated. The containment 
purge and vent valves' position is considered in the accident analyses, 
and could affect the analyzed consequences of events. The current 
limiting condition for operation (LCO) and Action Statement requires, 
and the surveillance verifies, that the permanently installed 70 deg. 
block is in place and effective. If the existing valves were open 
further than 70 deg., the valves may not close in time. The valves have 
a welded mechanical stop installed that limits the position to no more 
than 70 deg. open, which is a fixed condition that can only be changed 
by plant modification requiring evaluation against the requirements of 
10 CFR 50.59. The licensee considers the mechanical stop as sufficient 
to ensure the existing valves will remain within existing analysis 
bounds for a design basis loss of coolant accident (LOCA). In addition, 
the new valves are qualified to close within the 5 seconds assumed in 
the design basis LOCA. The licensee considers, therefore, that the 
existing and new valves will operate as required for accident 
mitigation with the proposed change, and that the proposed change will 
not affect the consequences of accidents previously evaluated.
    Regarding the modification of the containment leak testing 
requirements to reflect the new design valves: The containment purge 
and vent valves are not one of the initiating events for any previously 
evaluated accident in the WNP-2 FSAR. Thus the proposed change will not 
affect the probability of an accident previously evaluated. The metal 
to metal seat valves meet the Appendix J criteria necessary to be 
tested as type C valves. Type C valves can be tested every 2 years, 
compared to every 6 months for the current valves. The testing 
frequency is based on the performance of the valve types to ensure that 
they are capable of maintaining the necessary leak tightness over the 
test interval. The new valves' design has been certified to provide the 
same leak tightness over 2 years that the current valves provide over 6 
months, thus the consequences of analyzed events remains unaffected by 
the proposed change.
    Regarding the administrative changes: The proposed change would (1) 
delete a note that was applicable only through April 10, 1988, and (2) 
move an action that is currently stated in the SURVEILLANCE 
REQUIREMENTS section of a TS to the ACTIONS section of the same TS. 
These changes do not affect the design or operation of the plant or the 
implementation of the affected TS, and as such would not affect the 
probability or consequences of previously analyzed events.
    2. Does the amendment create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Regarding the removal of the requirement to ensure the remaining 
existing-design valves' position remains at less than or equal to 
70 deg.: No aspect of the design or plant operation is affected by 
deletion of the surveillance or removal of the reference to the block 
from the LCO and Action Statement, no new modes of plant operation are 
introduced, and the proposed change does not require physical 
modification of the plant. The valves not being replaced will continue 
to be limited from opening greater than 70 deg. by the welded and non-
adjustable blocking feature. The capability of these valves to close 
within 5 seconds to meet the limiting design basis accident (LOCA) will 
remain unchanged. The replacement valves will be capable of closing 
within the same 5 seconds from a full-open position of 90 deg.. Since 
the proposed change does not introduce any new component, system, or 
plant operating conditions, the change does not create the possibility 
of a new or different kind of accident from any previously analyzed.
    Regarding the modification of the containment leak testing 
requirements to reflect the new design valves: The proposed change in 
surveillance frequency for the replacement valves does not introduce 
any new mode of plant operation, nor does it involve plant 
modifications. The new valves operate in the same manner as the old 
valves, only the seating surfaces are different. This does not affect 
the way the valves operate to perform their function. The proposed 
change would not, therefore, involve any new or different kinds of 
accidents from any previously evaluated.
    Regarding the administrative changes: The proposed changes do not 
introduce any new modes of plant or equipment operation, nor do they 
involve physical modification of the plant. The proposed change would 
not, therefore, create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the amendment involve a significant reduction in a margin 
of safety?
    Regarding the removal of the requirement to ensure the remaining 
existing-design valves' position remains at less than or equal to 
70 deg.: The margin of safety of concern with the proposed change is 
the need for the containment purge and vent valves to close in 5 
seconds, which will ensure that part 100 limits for design basis events 
are not exceeded. The proposed change does not affect the maximum open 
position of the existing valves, thus the valves will still close 
within 5 seconds. In addition, the new valves, with their maximum full 
open position of 90 deg., are a new design that will still close within 
the five seconds from the full open position, thereby preserving the 
existing margin of safety.
    Regarding the modification of the containment leak testing 
requirements to reflect the new design valves: The margin of safety 
involved in the proposed TS change is the amount of leakage that may 
occur due to plant degradation that may affect the design basis 
accident assumptions for leakage. The new design valves have been 
certified to provide the same leak tightness over 2 years that the 
current valves provide over 6 months, thus the leakage assumptions for 
design basis events is unaffected. The proposed change would not, 
therefore, affect the margin of safety provided by the TS.
    Regarding the administrative changes: There are no margins of 
safety affected by the administrative changes.
    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: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: Theodore R. Quay

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington

    Date of amendment request: January 6, 1994
    Description of amendment request: The amendment proposes to modify 
the Technical Specifications (TS) to remove the requirements for the 
Seismic Monitoring Instrumentation from the TS and relocate them to the 
FSAR and plant procedures. The requirements described in the 
specifications will be maintained in the FSAR and plant procedures.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The staff's evaluation of the licensee's analysis is 
presented below:
    1. Does the amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The seismic monitors only provide monitoring and recording of 
seismic events that might occur in the vicinity of WNP-2. The 
instrumentation are not relied upon in current accident analyses for 
any automatic or manual initiation of safety systems in response to a 
seismic event. The proposed change would not, therefore, significantly 
increase the probability or consequences of a previously analyzed 
accident.
    2. Does the amendment create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not affect the manner in which the plant 
is operated, maintained, or tested. The proposed change would not, 
therefore, create the possibility of a new or different kind of 
accident from any previously analyzed.
    3. Does the amendment involve a significant reduction in a margin 
of safety?
    The seismic monitors provide monitoring and recording functions 
only, and are not relied upon in accident analyses for automatic or 
manual initiation of any safety system. Thus the results of analyzed 
events, and the associated margins of safety, are unaffected by the 
administrative removal of the seismic monitors from the TS. The 
proposed change would not, therefore, involve a significant reduction 
in a margin of safety.
    The NRC staff has determined that 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: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Attorney for licensee: Nicholas S. Reynolds, Esq., Winston & 
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: Theodore R. Quay

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington

    Date of amendment request: February 17, 1994
    Description of amendment request: The amendment proposes to modify 
the Technical Specifications (TS) to support hydrostatic testing of the 
reactor coolant system. Specifically, the proposed amendment would: (1) 
add a Special Test Exception that would allow Mode 4 (Cold Shutdown) 
operation up to 212 deg.F, compared to the current limit of 200 deg.F, 
without shutdown cooling in operation, to conduct hydrostatic testing, 
and (2) add a new reactor metal temperature vs reactor vessel pressure 
(P/T) limit curve that is applicable up to 8 effective full power years 
(EFPY), for use during hydrostatic testing and non-nuclear plant 
heatup.
    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. Does the amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Regarding the proposed Special Test Exception: The proposed change 
would allow performance of hydrostatic testing in OPERATIONAL CONDITION 
4 at temperatures greater than 200 deg.F but less than or equal to 
212 deg.F. Operating in this condition is only allowed if specified 
OPERATIONAL CONDITION 3 secondary containment requirements are met. The 
operating condition is not considered as an initiator for any event 
analyzed in the FSAR, therefore the proposed change would not affect 
the probability of an accident previously evaluated.
    The specified OPERATIONAL CONDITION 3 requirements compensate for 
the allowed temperature increase and assure that the consequences of a 
potential leak will be conservatively bounded by the existing FSAR 
accident analyses, as discussed below.
    The hydrostatic test is conducted near water solid, all rods in, 
and temperature less than or equal to 212 deg.F. The stored energy in 
the core will be very low (approximately 43 days of shutdown conditions 
and partial core replacement during refueling) and the potential for 
failed fuel and a subsequent increase in coolant activity above 
Technical Specification limits is minimal. In addition, secondary 
containment will be OPERABLE and capable of handling airbone 
radioactivity from leaks that could occur during the performance of the 
testing. Maintaining the temperature less than or equal to 212 deg.F 
will ensure that any leak will not flash to steam, thereby ensuring the 
potential for airborne activity remains low. Requiring the standby gas 
treatment system (SGTS) to be OPERABLE will conservatively ensure that 
any airborne radiation from leaks will be processed by the SGTS thereby 
limiting releases to the environment. Existing pipe breaks analyzed in 
Chapter 15 of the FSAR are bounding for the proposed condition. In the 
event of a large break, the reactor would rapidly depressurize, 
allowing the low pressure ECCS subsystems to operate. The capability of 
the subsystems required for OPERATIONAL CONDITION 4 would be adequate 
to keep the core flooded under this condition. Small system leaks would 
be detected by leakage inspections before significant inventory loss 
occurred. Thus the consequences of previously analyzed accidents are 
not increased by the proposed amendment.
    Regarding the proposed P/T limit curve: The proposed change would 
modify the P/T limit curves that are based on prevention of brittle 
fracture of the reactor vessel. The proposed change would result in 
plant operation closer to the actual brittle fracture condition of the 
reactor vessel, potentially making a brittle fracture more likely. This 
condition is offset by the slow heatup conducted using only pump heat, 
which would result in lower stresses in the reactor vessel than are 
assumed in the brittle fracture analyses. The resulting P/T limit curve 
based on 8 EFPY would have sufficient conservatism from the actual 
vessel brittle fracture condition to make vessel failure as unlikely as 
the original 32 EFPY curve.
    The potential reactor vessel failure mechanisms are not affected by 
the proposed change, therefore the consequences of previously analyzed 
accidents are unaffected by the proposed change.
    2. Does the amendment create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Regarding both the proposed Special Test Exception and the proposed 
P/T limit curve: The proposed change introduces no new failure modes, 
involves no physical modification to the plant or change in system 
configurations, nor does it involve changes in plant, system, or 
component operation. The proposed change, therefore, does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the amendment involve a significant reduction in a margin 
of safety?
    Regarding the proposed Special Test Exception: The hydrostatic test 
is conducted with low stored energy in the reactor, which is bounded by 
the assumed decay heat in current safety analyses. In the unlikely 
event that a leak from the reactor coolant system were to occur, the 
RPV would depressurize and the low pressure systems would be available 
to keep the core flooded. This would ensure that the fuel peak clad 
temperature would not exceed 2200 deg.F, which is the design basis that 
provides the margin of safety for the reactor itself. In addition, 
secondary containment will be maintained during the hydrostatic test, 
which would ensure that any potential airborne activity that might 
occur would be filtered through the SGTS. This would ensure that the 
current margins to the 10 CFR Part 100 limits remain bounded by current 
analyses. The proposed change would, therefore, not involve a 
significant reduction in the margins of safety.
    Regarding the proposed P/T limit curve: The proposed new curves 
would allow plant operation closer to the actual brittle fracture 
condition of the reactor vessel during hydrostatic test conditions 
only. This would result in a reduced margin in the protection afforded 
by the P/T curve. The new curves would, however, allow a lower 
temperature for conduct of the hydrostatic test, which would increase 
the heat sink available in the RCS, and increase the margin to decay 
heat loads assumed in accident analyses. This would result in reduced 
potential for extensive flow from any break, reduce the time for 
initiation of low pressure ECCS systems, and reduce the available 
radioactive decay products that are available for release during any 
postulated accident condition. The overall impact of the conditions 
increases the margin to 10 CFR Part 100 limits that are the design 
margin of safety for postulated loss of coolant accidents. The overall 
effect of the proposed change would not involve a significant reduction 
in the overall margins of safety.
    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: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: Theodore R. Quay

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

    Date of amendment request: February 23, 1994
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 
6.8.c by removing the requirement to conduct a biennial review of plant 
procedures in accordance with American National Standards Institute 
(ANSI) N18.7-1976. The licensee proposes using alternate programs, that 
are already in place, to ensure that procedures are periodically 
reviewed and maintained current. A biennial review of the Integrated 
Plant Emergency Operating Procedures (IPEOPs), however, would continue. 
The requirements for these alternate programs and for the IPEOP review 
would be added to the Operational Quality Assurance Program Description 
(OQAPD).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    The proposed changes were reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed changes will not:
    1) involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The likelihood that an accident will occur is neither increased 
or decreased by eliminating the periodic reviews of routine 
administrative and technical procedures. Sufficient controls are 
established to ensure that procedures impacting safety-related 
structures, systems, and components are maintained current, 
accurate, and usable. This TS change will therefore not impact the 
function or method of operation of plant equipment. Thus, a 
significant increase in the probability of a previously analyzed 
accident does not result due to this change. No systems, equipment, 
or components are affected by the proposed changes. Thus, the 
consequences of a malfunction of equipment important to safety 
previously evaluated in the Updated Safety Analysis Report (USAR) 
are not increased by this change. The proposed changes do not affect 
equipment or its operation, and, thus, do not affect the 
probabilities or consequences of an accident. Therefore, WPSC 
concludes that this change does not significantly increase the 
probability or consequences of an accident.
    2) create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not involve changes to the physical 
plant or operations. Since periodic procedure reviews do not 
contribute to accident initiation, a change related to such an 
activity does not produce a new accident scenario or produce a new 
type of equipment malfunction. Also, this change does not alter any 
existing accident scenarios. The proposed changes do not affect 
equipment or its operation, and thus, do not increase the 
possibility of a new or different kind of accident.
    3) involve a significant reduction in the margin of safety.
    The proposed changes do not affect equipment or its operation, 
and thus, do not involve any reduction in the margin of safety. 
Therefore, use of the proposed Technical Specification would not 
involve any 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 Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497.
    NRC Project Director: John N. Hannon

Previously Published Notices of Consideration of Issuance of 
Amendments to Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, and Opportunity for a Hearing

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

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

    Date of application for amendment: February 17, 1994
    Brief description of amendment request: The proposed amendments 
would revise the combined Technical Specifications (TS) for the Diablo 
Canyon Power Plant Unit Nos. 1 and 2 to revise TS 3/4.3.2, ``Engineered 
Safety Feature Actuation System Instrumentation,'' as follows: (1) 
Table 3.3-3, functional unit 6.c.2), channels to trip, would be changed 
from 2/steam generator in one steam generator to 2/steam generator in 
any 2 steam generators to correct an administrative error. (2) Table 
3.3-4 would be changed as follows: a. functional unit 4.6., Negative 
Steam Pressure Rate - High, trip setpoint and allowable value, would be 
changed from -100 psi/sec and -105.4 psi/sec to 100 psi and 105.4 psi, 
respectively; b. a note would be added stating that the time constants 
utilized in the rate-lag controller for Negative Steam Pressure Rate - 
High, are equal to 50 seconds.
    Date of individual notice in Federal Register: March 1, 1994 (59 FR 
9789)
    Expiration date of individual notice: March 31, 1994
    Local Public Document Room location: California Polytechnical State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

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

    Date of amendment request: March 4, 1994
    Brief description of amendment request: The proposed amendment 
would add a new Section 3/4.10.8, ``Inservice Leak and Hydrostatic 
Testing,'' and the Bases. The new section would allow Hope Creek to 
remain in OPERATIONAL CONDITION 4 with reactor coolant temperatures up 
to 212 *F to facilitate inservice leak and hydrostatic testing.
    Date of publication of individual notice in Federal Register: March 
16, 1994 (59 FR 12384)
    Expiration date of individual notice: April 15, 1994
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

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

    Date of amendment request: April 28, August 12, and November 17, 
1993, and February 2, 1994
    Brief description of amendment request: The proposed changes 
increase the spent fuel pool capacities for Salem 1 and 2 from the 
current 1170 fuel assemblies to 1632 fuel assemblies. Also, the decay 
time for refueling operations is being extended from 100 hours to 168 
hours.
    Date of publication of individual notice in Federal Register: March 
4, 1994 (59 FR 10440)
    Expiration date of individual notice: April 4, 1994
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.

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.

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

    Date of application for amendment: September 3, 1993, as 
supplemented February 1, 1994
    Brief description of amendment: The amendment revises the heatup 
and cooldown curves and the low-temperature overpressure protection 
(LTOP) controls. The changes to the LTOP controls support proposed 
modifications to allow a variable-setpoint (VLTOP) protection system. 
The VLTOP system will increase the allowable operating pressure band in 
the LTOP region and increase the flexibility in the use of the reactor 
coolant pumps.
    Date of issuance: March 15, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 185
    Facility Operating License No. DPR-53: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 29, 1993 (58 
FR 50963) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 15, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendments: September 17, 1993, as 
supplemented on January 4, 1994
    Brief description of amendments: The amendments implement the 
recommendations provided in Generic Letter 88-16, ``Removal of Cycle-
Specific Parameter Limits From Technical Specifications,'' by removing 
cycle specific values from the Technical Specifications (TSs) and 
incorporating them in a separate document. The amendments also include 
two other changes. One is the removal of outdated references to power 
operation with less than four reactor coolant pumps in operation and 
the other includes administrative changes to clarify the existing TSs, 
but do not alter the existing requirements.
    Date of issuance: March 17, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 186 and 163
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 27, 1993 (58 FR 
57844) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated March 17, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, 
Will County, Illinois

    Date of application for amendments: August 27, 1993, as 
supplemented February 21, 1994
    Brief description of amendments: The amendments revise the 
requirements for snubber visual inspection intervals and corrective 
actions in accordance with Generic Letter 90-09. The amendments also 
remove two of the options for determining the sample size to be used 
for snubber functional testing.
    Date of issuance: March 11, 1994
    Effective date: March 11, 1994
    Amendment Nos.: 60, 60, 48, and 48
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1994 (59 FR 
4935) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 11, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: For Byron, the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois 
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, 
Units 1 and 2, Rock Island County, Illinois

    Date of application for amendments: June 1, 1992
    Brief description of amendments: The amendments update the leakage 
test requirements of the drywell airlock to the standards of 10 CFR 
Part 50, Appendix J, Section III.D.2.
    Date of issuance: March 11, 1994
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 125, 119, 145, and 141
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29, and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 28, 1992 (57 FR 
48818) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 11, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: For Dresden, The Morris Public 
Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities, 
The Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021.

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: February 25, 1993
    Brief description of amendments: The amendments change Technical 
Specification 4.8.1.1.2.c to update the diesel fuel oil testing 
requirements to the standards of ASTM D4057-88 (new fuel oil test); 
ASTM D975-88 (water and sediment content testing); and ASTM D2276-89 
(impurity levels). The updated standards will be referenced in the 
Technical Specification Bases.
    Date of issuance: March 10, 1994
    Effective date: March 10, 1994
    Amendment Nos.: 97 and 81
    Facility Operating License Nos. NPF-11 and NPF-18. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 7, 1993 (58 FR 
36431) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 10, 1994. 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.

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 28, 1993, supplemented 
by letter dated January 21, 1994.
    Brief description of amendments: The amendments revise the ECCS 
injection valve stroke times and ECCS response times to allow the 
licensee to perform Motor Operated Valve modifications that slow down 
injection valve stroke times. As part of this change, a limited break 
spectrum Loss-Of-Coolant Accident analysis was performed to evaluate 
the impact of the slower response on the Peak Cladding Temperatures and 
to update the plants licensing bases.
    Date of issuance: March 9, 1994
    Effective date: March 9, 1994
    Amendment Nos.: 96 and 80
    Facility Operating License Nos. NPF-11 and NPF-18. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1994 (59 FR 
4937) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 9, 1994. 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.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of application for amendments: October 21, 1993
    Brief description of amendments: The amendments delete the 
requirements for demonstrating the operability of redundant equipment 
when emergency core cooling system equipment is found to be inoperable, 
or made inoperable for maintenance. The changes are consistent with the 
guidance provided by the NRC staff in Generic Letter 93-05, dated 
September 27, 1993.
    Date of issuance: March 8, 1994
    Effective date: March 8, 1994
    Amendment Nos.: 144 and 140
    Facility Operating License Nos. DPR-29 and DPR-30. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59747) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 8, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.

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: April 21, 1993
    Brief description of amendment: This amendment deleted License 
Condition 2.C(36), Attachment 1, Item (c)(4) which implemented the 
requirements of Regulatory Guide 1.97, ``Instrumentation For Light-
Water-Cooled Nuclear Power Plants to Assess Plant and Environs 
Conditions During and Following an Accident,'' for the Grand Gulf 
Nuclear Station because analysis shows that these requirements are 
being met by alternative methods.
    Date of issuance: March 7, 1994
    Effective date: March 7, 1994
    Amendment No: 112
    Facility Operating License No. NPF-29. Amendment revises the 
license.
    Date of initial notice in Federal Register: May 12, 1993 (58 FR 
28056) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 7, 1994. 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.

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

    Date of amendment request: May 8, 1991, as supplemented by letters 
dated March 6, 1992, and January 28, 1993
    Brief description of amendment: The amendment revised the Technical 
Specifications by revising the fuel oil amounts in the feed and storage 
tanks for the emergency diesel generators, clarifying the testing for 
the interconnecting piping, and revising the specific gravity of the 
fuel oil.
    Date of issuance: March 16, 1994
    Effective date: March 16, 1994
    Amendment No.: 92
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 26, 1991 (56 FR 
29274), as revised April 14, 1993 (58 FR 19478) The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated March 16, 1994. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

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: June 21, 1993
    Brief description of amendments: These amendments will change 
Technical Specifications Section 6.0, ``Administrative Controls,'' by 
(a) revising unit staff titles to those of the current FPL Nuclear 
Division organization, (b) revising the composition of the Facility 
Review Group (FRG) to broaden the scope of available expertise, and (c) 
making minor editorial corrections.
    Date of issuance: March 2, 1994
    Effective date: March 2, 1994
    Amendment Nos.: 126, 65
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 21, 1993 (58 FR 
39050) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 2, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

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

    Date of application for amendments: August 17, 1993, as 
supplemented January 14, 1994.
    Brief description of amendments: These amendments relocate fire 
protection requirements from the Technical Specifications to the Final 
Safety Analysis Report in accordance with Generic Letter 86-10, 
``Implementation of Fire Protection Requirements,'' and amend the 
license conditions accordingly.
    Date of issuance: February 25, 1994
    Effective date: February 25, 1994
    Amendment Nos. 159 and 153
    Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments 
revised the Licenses and Technical Specifications.
    Date of initial notice in Federal Register: September 29, 1993 (58 
FR 50967) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 25, 1994. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

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

    Date of application for amendment: December 16, 1993
    Brief description of amendment: The amendment revises the Technical 
Specifications to clarify the requirements for maintaining secondary 
containment integrity when one or more Reactor Building Ventilation 
supply and exhaust valves are declared inoperable. The Technical 
Specifications add a new Limiting Condition for Operation, Basis 
Statement and Surveillance Requirements for these isolation valves.
    Date of issuance: March 7, 1994
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 168
    Facility Operating License No. DPR-16. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1994 (59 FR 
4938) The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated March 7, 1994. 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. 50-289, Three Mile 
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of application for amendment: August 26, 1993
    Brief description of amendment: The amendment revises the plant 
Technical Specifications (TSs) to accommodate limited fuel 
reconstitution based on NRC Generic Letter (GL) 90-02, Supplement 1. 
Such reconstitution may be appropriate in the event of a leaking fuel 
rod, in which case the fuel rod would be replaced with a stainless 
steel or zirconium alloy filler rod.
    Date of issuance: March 15, 1994
    Effective date: As of its date of issuance, to be implemented 
within 30 days of issuance.
    Amendment No.: 183
    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 March 15, 1994. 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.

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

    Date of application for amendments: April 16, 1991, as supplemented 
January 6, 1993.
    Brief description of amendments: The amendments revise the 
technical specifications to incorporate recommendations from NRC 
Generic Letter 90-06 for power-operated relief valve and block valve 
reliability and low-temperature overpressure protection.
    Date of issuance: March 9, 1994
    Effective date: March 9, 1994
    Amendment Nos.: 176 & 161
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 3, 1993 (58 FR 
12261) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 9, 1994
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

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

    Date of application for amendment: January 25, 1993, as 
supplemented by letters dated November 3 and 23, and December 9, 1993, 
and January 5 and 24, 1994.
    Brief description of amendment: This amendment increases the 
maximum number of spent fuel assemblies that can be stored in the Maine 
Yankee fuel pool to 2019 from 1476. The increase in fuel storage 
capacity is required so that storage space is available for spent fuel 
through the duration of the current operating license, including the 
final full core offload.
    Date of issuance: March 15, 1994
    Effective date: March 15, 1994
    Amendment No.: 144
    Facility Operating License No. DPR-36: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 26, 1993 (58 FR 
16423) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 15, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, Maine 04578.

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

    Date of application for amendment: December 22, 1993
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.4.4.e (Emergency Ventilation System) to permit 
fuel handling operations to continue during refueling beyond 7 days 
with one circuit of the emergency ventilation system inoperable, 
provided the remaining emergency ventilation system circuit is operable 
and in operation. The change to TS 3.4.4.e is consistent with the NRC's 
Improved Standard Technical Specifications, NUREG-1433.
    Date of issuance: March 8, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 146
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1994 (59 FR 
4940) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 8, 1994. 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.

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

    Date of application for amendment: December 27, 1993
    Brief description of amendment: The amendment relocates TS Tables 
3.2.7, ``Reactor Coolant Isolation Valves,'' and 3.3.4, ``Primary 
Containment Isolation Valves,'' from TSs 3.2.7/4.2.7 and 3.3.4/4.3.4, 
respectively, to a plant procedure which governs lists removed from TSs 
per Generic Letter (GL) 91-08, ``Removal of Component Lists from 
Technical Specifications.'' The plant procedure would be subject to the 
requirements specified in the Administrative Controls section of the 
NMP-1 TSs. The proposed amendment would also make conforming changes to 
the TS Bases. These lists of valves will continue to be included in the 
NMP-1 Updated Final Safety Analysis Report. Relocation of these valve 
lists from the NMP-1 TSs to the plant procedure is consistent with NRC 
staff guidance issued in GL 91-08.
    Date of issuance: March 7, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 145
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1994 (59 FR 
4941) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 7, 1994. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: June 11, 1993, supplemented by 
letter dated November 15, 1993.
    Brief description of amendment: The amendment revises the pressure/ 
temperature (P/T) limits for the reactor vessel. Specifically, Figure 
3.4-2, ``Millstone Unit 2 Reactor Coolant System Pressure-Temperature 
Limitations for 12 Full Power Years,'' on page 3/4 4-19, is revised to 
reflect the change in the curves and the title change to ``Millstone 
Unit 2 Reactor Coolant System Pressure-Temperature Limitations for 20 
EFPY.''
    Date of issuance: January 27, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 170
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 21, 1993 (58 FR 
39054) The November 15, 1993, submittal provided information 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 January 27, 1994. 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.

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

    Date of amendment request: May 7, 1993
    Description of amendment request: The amendment changes Technical 
Specification (TS) 3/4 6.1 relating to primary containment integrity. 
Limiting Condition for Operation (LCO) 3.6.1.7 is changed to delete the 
requirements applicable to the 36-inch containment shutdown purge 
supply and exhaust isolation valves in the containment air purge (CAP) 
system. Surveillance Requirement (SR) 4.6.1.7.1 and associated footnote 
and SR 4.6.1.7.2 are deleted also. To maintain document consistency, 
certain other editorial changes were made.
    Date of issuance: March 7, 1994
    Effective date: Not effective until operational MODE 5 is entered 
when commencing the third refueling outage, and is to be implemented 
prior to reentering operational MODE 4.
    Amendment No.: 29
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 23, 1993 (58 FR 
34083). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 7, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 47 
Front Street, Exeter, New Hampshire 03833.

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

    Date of application for amendments: July 7, 1993
    Brief description of amendments: The amendments revise the combined 
Technical Specifications (TS) for the Diablo Canyon Power Plant Unit 
Nos. 1 and 2 to change TS 5.1.3, ``Map Defining Unrestricted Areas and 
Site Boundary for Radioactive Gaseous and Liquid Effluents,'' to be 
consistent with a recent interpretation of the restricted area 
definition in 10 CFR 20.
    Date of issuance: March 3, 1994
    Effective date: March 3, 1994
    Amendment Nos.: 90 & 89
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 18, 1993 (58 FR 
43930) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 3, 1994, and an 
environmental assessment was published in the Federal Register on 
February 25, 1994, (59 FR 9252). No significant hazards consideration 
comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

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

    Date of application for amendments: September 8, 1993 (Reference 
LAR 93-06)
    Brief description of amendments: The amendments revise the combined 
Technical Specifications (TS) for the Diablo Canyon Power Plant Unit 
Nos. 1 and 2. Specifically, TS 1.44, ``Radiological Monitoring and 
Controls Program,'' 3/4.11, ``Radioactive Effluents,'' and 6.14, 
``Radiological Monitoring and Controls Program (RMCP), Offsite Dose 
Calculation Procedure (ODCP) and Environmental Radiological Monitoring 
Procedure (ERMP),'' are revised to change the Semiannual Radioactive 
Effluent Release Report to Annual Radioactive Effluent Release Report. 
The amendment also revises TS 6.2.3, ``Onsite Safety Review Group 
(OSRG),'' 6.5.2, ``Plant Staff Review Committee,'' and 6.5.3.7, 
``Nuclear Safety Oversight Committee Review,'' to implement 
organizational changes.
    Date of issuance: March 7, 1994
    Effective date: March 7, 1994
    Amendment Nos.: 91 and 90
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 27, 1993 (58 FR 
57855) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 7, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

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

    Date of application for amendments: October 8, 1993
    Brief description of amendments: The amendments revised the 
existing definition of CHANNEL CALIBRATION in Technical Specification 
1.4 to allow in-place qualitative methods to be used to verify 
resistance temperature detector or thermocouple sensor behavior.
    Date of issuance: March 8, 1994
    Effective date: March 8, 1994
    Amendment Nos.: 133 and 102
    Facility Operating License Nos. NPF-14 and NPF-22. These amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59754) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 8, 1994. 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.

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

    Date of application for amendments: November 1, 1993 as 
supplemented January 26, 1994 and February 18, 1994.
    Brief description of amendments: These amendments concern the 
Radiation Monitoring Systems - Isolation and Initiation Functions of 
the Technical Specifications and are necessary to support modification 
5281. This modification replaces the obsolete control room ventilation 
radiation monitoring equipment.
    Date of issuance: March 15, 1994
    Effective date: March 15, 1994
    Amendments Nos.: 184 and 189
    Facility Operating License Nos. DPR-44 and DPR-56: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64614) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 15, 1994. No significant 
hazards consideration comments received: No
    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.

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

    Date of application for amendment: July 15, 1993
    Brief description of amendment: The amendment revises the Technical 
Specifications to eliminate the reactor scram and Main Steam Line 
Isolation Valve closure requirements associated with the Main Steam 
Line Radiation Monitors. The changes are consistent with Licensing 
Topical Report NEDO-31400, ``Safety Evaluation for Eliminating the 
Boiling Water Reactor Main Steam Isolation Valve Closure Function and 
Scram Function of the Main Steam Line Radiation Monitor,'' dated May 
1987.
    Date of issuance: March 9, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 207
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 4, 1993 (58 FR 
41513) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 9, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendments: October 16, 1992
    Brief description of amendments: These amendments revise TS 3/
4.3.4, ``Turbine Overspeed Protection,'' to allow one surveillance 
every 31 days for verification of turbine overspeed protection system 
operability. Currently, the surveillance tests are performed at power 
every 7 days and again every 31 days. The 31-day test is performed by 
an operator with an observer at the valve.
    Date of issuance: March 9, 1994
    Effective date: March 9, 1994
    Amendment Nos.: 111 and 100
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 17, 1993 (58 
FR 8783) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 9, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713

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

    Date of application for amendments: January 10, 1992 (TS304)
    Brief description of amendments: The amendments address emergency 
diesel generator availability for the plant shared systems of Standby 
Gas Treatment and Control Room Emergency Ventilation.
    Date of issuance: March 9, 1994
    Effective date: March 9, 1994
    Amendment Nos.: 203, 222 and 176
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 15, 1992
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 9, 1994. No significant hazards 
consideration comments received: None
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: January 8, 1993; which was 
supplemented by submittals dated April 1, May 3, and August 18, 1993; 
and February 22, 1994.
    Brief description of amendments: The amendments remove the 
surveillance requirement to perform reactor vessel nozzle inspections 
at the end of each 10-year inspection interval.
    Date of issuance: March 15, 1994
    Effective date: March 15, 1994
    Amendment Nos.: 177 and 168
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: February 3, 1993 (58 FR 
7007) The Commission's related evaluation of the amendments are 
contained in a Safety Evaluation dated March 15, 1994. No significant 
hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.

    Date of application for amendments: July 2, 1993, as supplemented 
December 10, 1993
    Brief description of amendments: These amendments modify the 
Technical Specifications having cycle-specific parameters limits by 
replacing the values of those limits with a reference to a Core 
Operating Limits Report for the values of those limits.
    Date of issuance: March 2, 1994
    Effective date: March 2, 1994
    Amendment Nos. 189 and 189
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 4, 1993 (58 FR 
41519) The December 10, 1993, submittal did not expand the scope of the 
original application and did not change the proposed no significant 
hazards determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated March 2, 1994. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185

Virginia Electric and Power Company, Docket Nos. 50-280, 50-281, 
50-338, and 50-339, Surry Power Station, Unit Nos. 1 and 2, Surry 
County, Virginia, and North Anna Power Station, Unit Nos. 1 and 2, 
Louisa County, Virginia.

    Date of application for amendments: July 20, 1993
    Brief description of amendments: These amendments delete the 
Technical Specifications requirement for Station Nuclear Safety and 
Operating Committee review of the Emergency and Security Plans. This 
requirement remains in the respective plans. The audit frequencies are 
also being deleted from the TS.
    Date of issuance: March 1, 1994
    Effective date: March 1, 1994
    Amendment Nos. 188, 188, (Surry 1&2) 180, 161 (North Anna 1&2)
    Facility Operating License Nos. DPR-32, DPR-37, NPF-4 and NPF-7: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 1, 1993 (58 
FR 46242) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 1, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room locations: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185, and The Alderman 
Library, Special Collections Department, University of Virginia, 
Charlottesville, Virginia 22903-2498.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.

    Date of application for amendments: December 10, 1993
    Brief description of amendments: These amendments modify the 
surveillance requirements for the Auxiliary Feedwater System pumps and 
valves, define ``staggered test basis,'' and make administrative 
changes to the Technical Specifications.
    Date of issuance: March 7, 1994
    Effective date: March 7, 1994
    Amendment Nos. 190 and 190
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 19, 1994 (59 FR 
2873) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 7, 1994. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185

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 April 29, 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).

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

    Date of application for amendment: February 25, 1994 as 
supplemented on March 11, 1994
    Brief description of amendment: The amendment revised the Technical 
Specifications by adding a footnote to Specification 3/4.4.3.1, 
``Reactor Coolant System Leakage - Leakage Detection Systems,'' to 
permit continued plant operations with inoperable drywell floor drain 
sump flow monitoring instrumentation until the first time the plant is 
required to be brought to COLD SHUTDOWN after March 15, 1994.
    Date of issuance: March 14, 1994
    Effective date: March 14, 1994
    Amendment No.: 89
    Facility Operating License No. NPF-62. The amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazardsconsideration: 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 March 
14, 1994.
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.

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

    Date of application for amendment: December 15, 1993, as 
supplemented February 15 and 24, 1994 (December 15, 1993, application 
supersedes the licensee's March 10, 1993 application.)
    Brief description of amendment: The amendment revises the Technical 
Specifications to allow the continuance of voltage-based steam 
generator tube plugging criteria for outside-diameter stress corrosion 
cracking at tube support plate elevations. The amendment allows the use 
of a 2.0 volt interim repair criterion for Cycle 14 operation.
    Date of issuance: March 15, 1994
    Effective date: March 15, 1994
    Amendment No.: 178
    Facility Operating License No. DPR-58. Amendment revises the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration. Yes. The December 15, 1993, 
application was noticed in the Federal Register on January 5, 1994 (59 
FR 621). The NRC also published a public notice of the proposed 
amendment, issued a proposed finding of no significant hazards 
consideration, and requested that any comments on the proposed finding 
be provided to the staff by the close of business on March 7, 1994. The 
notice was published in the South Haven Tribune on March 1, 1994, and 
in the Herald-Palladium on March 2, 1994. No comments have been 
received.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, consultation with the State of Michigan, and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated March 15, 1994.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    NRC Project Director: Ledyard B. Marsh

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

    Date of application for amendment: February 15, 1994
    Brief description of amendment: The amendment revises the Technical 
Specifications
    Date of issuance: March 14, 1994
    Effective date: March 14, 1994
    Amendment No.: 177
    Facility Operating License No. DPR-58. Amendment revises the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration. Yes. The NRC published a public 
notice of the proposed amendment, issued a proposed finding of no 
significant hazards consideration, and requested that any comments on 
the proposed finding be provided to the staff by the close of business 
on March 7, 1994. The notice was published in the South Haven Tribune 
on March 1, 1994, and in the Herald-Palladium on March 2, 1994. No 
comments have been received.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, consultantion with the State of Michigan, and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated March 14, 1994
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    NRC Project Director: Ledyard B. Marsh
    Dated at Rockville, Maryland, this 23rd day March 1994.
    For the Nuclear Regulatory Commission.
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear 
Reactor Regulation
[Doc. 94-7331 Filed 3-29-94; 8:45 am]
BILLING CODE 7590-01-F