[Federal Register Volume 59, Number 148 (Wednesday, August 3, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-18741]


[[Page Unknown]]

[Federal Register: August 3, 1994]


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NUCLEAR REGULATORY COMMISSION
Biweekly Notice

 

Applications and Amendments to Facility Operating 
LicensesInvolving No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from July 11, 1994, through July 22, 1994. The 
last biweekly notice was published on July 20, 1994 (59 FR 37060).

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

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

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson 
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: June 29, 1994
    Description of amendment request: The proposed amendment will 
delete the requirement to perform alternate train testing to 
demonstrate that other, similar, safety-related components are operable 
when components are found, or made, inoperable in the safety injection, 
residual heat removal, and containment spray systems. The surveillance 
requirements, which the licensee refers to as accelerated testing 
requirements, affect the following components:
    (a)Safety Injection (SI) pumps TS 3.3.1.2.b)
    (b) Residual Heat Removal (RHR) Pumps (TS 3.3.1.2.c)
    (c) SI and RHR flow paths (TS 3.3.1.2.e)
    (d) Containment Spray (CS) (TS 3.3.2.2.a and b)
    (e) CS flow paths (TS 3.3.2.2.c)
    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 involve a significant increase 
in the probability of an accident previously evaluated because the 
availability of the subject components will not be reduced and the 
design and performance of the components are not being changed. The 
subject components are provided to mitigate the consequences of 
analyzed accidents; therefore their availability has no bearing on 
the probability of occurrence of these accidents.
    The proposed amendment does not involve a significant increase 
in the consequences of an accident previously evaluated. This change 
deletes alternate train testing requirements which, if maintained, 
could result in loss of the safety function. Elimination of the 
requirements will serve to ensure that one train of safety equipment 
is always available to mitigate the consequences of an analyzed 
accident. The remaining surveillance requirements provide adequate 
assurance that the components will be operable when required. 
Therefore the consequences of previously evaluated accidents will 
not be increase.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any previously evaluated 
because these proposed changes do not introduce any new modes of 
operation or testing, and no physical changes are being made to the 
plant. Therefore no new or different kind of accident could be 
initiated by this amendment.
    3. The proposed revisions do not involve a significant reduction 
in the margin of safety since the routine testing requirements that 
remain in the Technical Specifications provide adequate assurance 
that the components will be operable when needed. Since the 
elimination of this accelerated testing will decrease component wear 
and improve availability, the margin of safety should be increased. 
Since accelerated testing may still occur when component problems 
involve a potential common mode failure, margins of safety 
associated with the components' abilities to perform their design 
functions will not be affected. Therefore, the proposed changes do 
not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Hartsville Memorial Library, 
Home and Fifth Avenues, Hartsville, South Carolina 29550
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: David B. Matthews

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

    Date of amendment request: June 13, 1994
    Description of amendment request: The proposed amendment would make 
several changes to the Administrative Controls in Section 6 of 
Technical Specifications (TS) for Byron and Braidwood stations. The 
proposed changes include: (1) a change to the submittal frequency of 
the Radiological Effluent Release Report, (2) a revision to the Shift 
Technical Advisor description, (3) clarification of the Shift 
Engineer's responsibilities, and (4) editorial changes. The references 
to the Semiannual Radiological Effluent Release Report are also revised 
in other sections of 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, which is presented below:
    A. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes to Section 6 of Technical Specifications do 
not affect any accident initiators or precursors and do not change 
or alter the design assumptions for the systems or components used 
to mitigate the consequences of an accident.
    The proposed changes are administrative in nature and provide 
clarification. These changes provide consistency with station 
procedures, programs, the Code of Federal Regulations, other 
Technical Specifications, and Standard Technical Specifications. 
These changes do not impact any accident previously evaluated in the 
UFSAR.
    B. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not affect the design or operation of 
any system, structure, or component in the plant. There are no 
changes to parameters governing plant operation; no new or different 
type of equipment will be installed. The proposed changes are 
considered to be administrative changes. All responsibilities 
described in Technical Specifications for management activities will 
continue to be performed by qualified individuals.
    C. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes do not affect the margin of safety for any 
Technical Specification. The initial conditions and methodologies 
used in the accident analyses remain unchanged, therefore, accident 
analysis results are not impacted.
    The proposed changes are administrative in nature and have no 
impact on the margin of safety of any Technical Specification. They 
do not affect any plant safety parameters or setpoints. The 
descriptions for the Shift Technical Advisor and Shift Engineer are 
clarified, however, include no reduction to their responsibilities.
    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: 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
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: July 6, 1994
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3/4.4.5, ``Steam Generators,'' and 
the associated bases. Previously the NRC granted amendments to the TSs 
which authorized the use of selected steam generator sleeving 
processes. In authorizing use of the processes, the amendments cited 
references to specific NRC approved vendor technical reports, including 
revision number. The proposed changes reference the reports in generic 
terms as those that have been approved by the NRC, subject to 
limitations and restrictions as noted by the NRC staff. While the 
licensee will still have to request NRC approval for application of the 
technologies as referenced in vendor reports, the licenses will not 
have to be amended each time.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes to the Steam Generator section of Technical 
Specifications do not affect any accident initiators or precursors 
and do not alter the design assumptions for the systems or 
components used to mitigate the consequences of an accident. These 
changes are editorial changes to the requirements currently 
identified in the Technical Specifications. The requirements 
approved by the NRC will not be reduced by this request. The 
proposed change maintains the administrative controls necessary to 
ensure safe plant operation.
    The original amendment requested tubesheet sleeves and tube 
support plate sleeves as an alternate tube repair method for Bryon 
and Braidwood Units 1 and 2. The steam generator sleeves approved 
for installation use the Westinghouse process (laser welded joints) 
or the Babcock & Wilcox Nuclear Technologies (BWNT) process of 
kinetically welded joints. The sleeve configuration was designed and 
analyzed in accordance with the criteria of Regulatory Guide (RG) 
1.121 and the design requirements of Section III of the American 
Society of Mechanical Engineers (ASME) Code. Fatigue and stress 
analyses of the sleeved tube assemblies for both processes produced 
acceptable results documented in the current Westinghouse and BWNT 
Technical Reports. The proposed Technical Specifications change to 
allow the use of the current NRC approved laser welded or 
kinetically welded sleeving process does not adversely impact any 
other previously evaluated design basis accident or the results of 
these analyses. Therefore, the editorial changes to the referenced 
sleeving Technical Reports will not increase the probability of 
occurrence of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes are considered to be administrative 
changes. All the requirements described in Technical Specifications 
``Acceptance Criteria'' for the Steam Generators will continue to be 
implemented as described in the current Technical Reports.
    Referencing the current Westinghouse or BWNT Sleeving Technical 
Reports currently approved by the NRC and subject to the limitations 
and restrictions as noted by the NRC, has no effect upon any design 
transient or accident analyses. The proposed changes do not affect 
the design or operation of any system, structure, or component in 
the plant. There are no changes to parameters governing plant 
operation and no new or different type of equipment will be 
installed.
    The use of the proposed sleeving processes will not introduce 
significant or adverse changes to the plant design basis. Stress and 
fatigue analyses of the repair have shown the ASME Code and
    RG 1.121 allowable values are met. Implementation of the 
currently approved laser welded or kinetically welded sleeving will 
continue to maintain the overall tube bundle structural integrity at 
a level consistent with that of the originally supplied tubing. 
Repair of a tube with a sleeve does not provide a mechanism which 
would result in an accident outside of the area affected by the 
sleeve. Any hypothetical accident as a result of potential tube or 
sleeve degradation in the repaired portion of the tube is bounded by 
the existing steam generator tube rupture accident analysis. The 
tube rupture accident analysis accounts for the installation of 
sleeves and the impact on current plugging level analyses. The 
sleeve design does not affect any other component or location on the 
tube outside of the immediate area repaired.
    Thus, the possibility of a new or different type of accident 
from any accident previously evaluated is not created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change is administrative in nature and has no 
impact on the margin of safety of any Technical Specification. 
Specific technical reports are no longer referenced in Technical 
Specifications. An editorial change is made to TS referencing the 
current NRC approved vendor Technical Report, subject to the 
limitations and restrictions noted by the NRC. The initial 
conditions and methodologies used in the accident analyses remain 
unchanged.
    The laser welded and kinetically welded sleeving repair of 
degraded steam generator tubes has been shown by analysis to restore 
the integrity of the tube bundle to its original design basis 
condition. The safety factors used in the design of sleeves for the 
repair of degraded tubes are consistent with the safety factors in 
the ASME Boiler and Pressure Vessel Code used in steam generator 
design. The design of the tube sleeves has been verified by testing 
to preclude leakage during normal and postulated accident 
conditions. Installation of either type of vendor sleeve using the 
current approved process will continue to maintain the structural 
integrity of the steam generator tubes.
    Thus, these changes do not involve a significant reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
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: 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
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: June 9, 1994
    Description of amendment request: The proposed amendments would 
revise the LaSalle County Station, Units 1 and 2, Technical 
Specifications (TS), Appendix A, in order to facilitate implementation 
of the Thermal Limits portion of the General Electric Average Power 
Range Monitor (APRM)/Rod Block Monitor (RBM)/TS Improvement Program 
(ARTS).
    Specifically, the proposed TS change will create power and flow 
dependent Minimum Critical Power Ratio (MCPR) and Maximum Average 
Planar Linear Heat Generation Rate (MAPLHGR) limits, and other 
administrative changes.
    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 increase in 
the probability or consequences of an accident previously evaluated 
because:
    The probability of an accident previously evaluated will not 
increase as a result of this change, because no changes to plant 
systems will occur. All changes are related to core monitoring 
software, and there will be no physical changes to equipment.
    The consequences of an accident previously evaluated will not 
increase as a result of the proposed changes. The power- and flow-
dependent MCPR and MAPLHGR limits incorporate sufficient 
conservatism so the safety limit MCPR [SLMCPR] (operating limit MCPR 
[OLMCPR] for automatic flow control) and the fuel thermal-mechanical 
limits will not be violated for any power and flow condition. 
Because these limits are protected during normal operation, the 
consequences of any transient will not increase with this change in 
limit definition. General Electric has verified in Attachment E that 
the introduction of Arts will not cause any change in the Licensing 
Basis PCT [Peak Centerline Temperature] resulting from a Loss-Of-
Coolant Accident [LOCA], nor any change in the results satisfying 
the other LOCA acceptance criteria of 10 CFR 50.46 and Section 
15.6.5 of NUREG-0800 (Standard Review Plan), which are: cladding 
oxidation, metal-water reaction (hydrogen generation), coolable 
geometry and long-term cooling.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because:
    Since no physical changes to any plant system are occurring, 
there will be no new or different types of accidents created by this 
change. No interactions between equipment systems will be changed in 
any manner.
    The proposed changes do not involve a significant reduction in a 
margin of safety because:
    The power- and flow-dependent MCPR and MAPLHGR limits will 
sufficiently protect the SLMCPR (OLMCPR for automatic flow control) 
and the fuel thermal-mechanical limits at all power and flow 
conditions. The ARTS limits conservatively assure that all licensing 
criteria are satisfied without setdown of the flow referenced APRM 
scram and rod block trips. The limits were developed using NRC 
approved methods, and satisfy the same NRC approved criteria that 
the APRM setdown requirement does.
    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: Public Library of Illinois 
Valley Community College, Rural Route No. 1, Ogelsby, Illinois 61348
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: July 13, 1994
    Description of amendment request: The requested amendments would 
allow the testing interval in Technical Specification Surveillance 
Requirement 4.6.2 for the air or smoke flow test through each 
containment spray header to be increased from 5 to 10 years. The 
licensee states that the proposed amendments are consistent with NRC 
staff guidance contained in NUREG-1366, ``Improvements to Technical 
Specifications Surveillance Requirements,'' and Generic Letter 93-05, 
``Line-Item Technical Specifications Improvements to Reduce 
Surveillance Requirements for Testing During Power Operation.'' In 
addition, the amendments would also remove an obsolete footnote related 
to the Catawba Unit 1 first refueling. The licensee's application 
jointly addressed both its Catawba and McGuire Nuclear Stations. This 
notice addresses those aspects applicable only to the Catawba Station.
    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
    The requested amendments will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. Increasing the surveillance interval of TS [Technical 
Specification] 4.6.2d from five to ten years will have no impact 
upon the probability of any accident, since the NS [containment 
spray] system is not accident initiating equipment. Also, since 
Catawba's... flow test histor[y] support[s] making the proposed 
change, system response following an accident will not be adversely 
affected. Therefore, the requested amendments will not result in 
increased accident consequences. Deletion of the obsolete footnote 
as indicated in the Catawba TS markup is purely an administrative 
change, and therefore will have no impact upon either the 
probability or consequences of any accident.
    Criterion 2
    The requested amendments will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. As stated above, the NS system is not accident initiating 
equipment. No new failure modes can be created from an accident 
standpoint. The plant will not be operated in a different manner. 
Deletion of the Catawba obsolete footnote has no bearing on any 
accident initiating mechanisms.
    Criterion 3
    The requested amendments will not involve a significant 
reduction in a margin of safety. Plant safety margins will be 
unaffected by the proposed changes. The NS system will still be 
capable of fulfilling its required safety function, since plant 
operating experience supports the proposed change. Finally, the 
proposed amendments are consistent with the NRC position and 
guidance set forth in NUREG-1366 and Generic Letter 93-05. Deletion 
of the Catawba obsolete footnote will not result in any impact to 
plant safety margins.
    Based upon the preceding analyses, Duke Power Company concludes 
that the requested amendments 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: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: July 13, 1994
    Description of amendment request: The requested amendments would 
allow the testing interval in Technical Specification Surveillance 
Requirement 4.6.2 for the air or smoke flow test through each 
containment spray header to be increased from 5 to 10 years. The 
licensee states that the proposed amendments are consistent with NRC 
staff guidance contained in NUREG-1366, ``Improvements to Technical 
Specifications Surveillance Requirements,'' and Generic Letter 93-05, 
``Line-Item Technical Specifications Improvements to Reduce 
Surveillance Requirements for Testing During Power Operation.'' The 
licensee's application jointly addressed both its Catawba and McGuire 
Nuclear Stations. This notice addresses those aspects applicable only 
to the McGuire Station.
    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
    The requested amendments will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. Increasing the surveillance interval of TS [Technical 
Specification] 4.6.2d from five to ten years will have no impact 
upon the probability of any accident, since the NS [containment 
spray] system is not accident initiating equipment. Also, since... 
McGuire's flow test histor[y] support[s] making the proposed change, 
system response following an accident will not be adversely 
affected. Therefore, the requested amendments will not result in 
increased accident consequences. ...
    Criterion 2
    The requested amendments will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. As stated above, the NS system is not accident initiating 
equipment. No new failure modes can be created from an accident 
standpoint. The plant will not be operated in a different manner. 
...
    Criterion 3
    The requested amendments will not involve a significant 
reduction in a margin of safety. Plant safety margins will be 
unaffected by the proposed changes. The NS system will still be 
capable of fulfilling its required safety function, since plant 
operating experiences supports the proposed change. Finally, the 
proposed amendments are consistent with the NRC position and 
guidance set forth in NUREG-1366 and Generic Letter 93-05. ...
    Based upon the preceding analyses, Duke Power Company concludes 
that the requested amendments 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: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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: December 8, 1993, as supplemented April 
20, 1994.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 3.4 to address the need to bypass 
automatic initiation of the Emergency Feedwater (EFW) system when the 
main feedwater pump discharge pressure is below actuation setpoint 
during startup and shutdown in order to prevent inadvertent actuation. 
The proposed amendment is in response to NRC Inspection Report 50-269, 
50-270, 50-287/90-30 (Inspector Followup Item 90-30-02), which 
determined that the existing TSs regarding initiation circuitry for the 
EFW system were inadequate. The amendments would also delete 
operability requirements for the Emergency Condenser Cooling Water 
(ECCW) system. The licensee determined that the ECCW system is not 
required to remove decay heat following any design basis event.
    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 amendment request would not] involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Each accident analysis addressed within the Oconee FSAR [Final 
Safety Analysis Report] has been examined with respect to changes 
proposed within this amendment request. The design basis of the 
Emergency Feedwater (EFW) System is to supply feedwater to the steam 
generators in the event Main Feedwater is lost. The EFW system 
provides the required flow rate to cool the RCS [reactor coolant 
system] down to the point at which the Decay Heat Removal System is 
designed to operate. The EFW system is also designed to cool the RCS 
following a small break LOCA [loss of coolant accident]. Changes 
included within this amendment request are provided to clarify 
requirements for the operability of EFW. Specifically, theses 
changes clarify that automatic initiation circuitry due to low main 
feedwater pump discharge pressure or low hydraulic oil pressure may 
be bypassed when the reactor is shutdown to prevent inadvertent 
actuation. In addition, these changes provide that if an EFW pump is 
inoperable due only to the inoperability of automatic initiation, 
cooldown to below 250 deg.F is not necessary after the reactor is 
shutdown. Accident analysis for the loss of main feedwater, and 
subsequent initiation of EFW, assumes initial conditions of the 
reactor at full power operation. The utilization of criticality for 
this specific automatic initiation circuitry to be operable ensures 
the EFW system is operated within the boundaries of design basis for 
Oconee while also providing a reasonable margin to prevent 
inadvertent actuation. It is not possible to place this automatic 
initiation circuitry in service prior to exceeding 250 deg.F because 
the main feedwater pump discharge pressure is well below the 
initiation setpoint at this value. Manual initiation circuitry 
operability is required prior to exceeding an RCS temperature of 
250 deg.F. This change only clarifies existing configuration and 
control for the Oconee units and does not increase the probability 
or consequences of any accident previously evaluated.
    This change also removes the requirement for Emergency Condenser 
Cooling Water (ECCW) System operability for the removal of decay 
heat using the secondary systems. The ability to provide flow 
through the condenser from the ECCW system is a preferred method for 
decay heat removal. However, this mode of operation is not necessary 
to prevent or mitigate any accident previously evaluated. The 
primary success path for decay heat removal following loss of 
station power events, and thus loss of normal CCW flow, is the use 
of the turbine driven EFW pump providing flow to the steam 
generators and heat removal via the main steam safety relief valves 
to the atmosphere. Analysis has shown that sufficient inventory 
exists in secondary systems, as limited by Technical Specification 
3.4.4, to provide for decay heat removal.
    Therefore, this proposed change deletes the requirement for ECCW 
for secondary systems decay heat removal. The probability or 
consequences of any design basis accident are not increased by this 
change. As such, this change will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    (2) [The amendment request would not] create the possibility of 
a new or different kind of accident from any kind of accident 
previously evaluated.
    Changes included within this amendment request are provided to 
clarify existing requirements for operability of the EFW System and 
remove the requirement for ECCW flow through the condenser for decay 
heat removal. Operation of Oconee units in accordance with these 
Technical Specifications will not create any failure modes not 
bounded by previously evaluated accidents. Previously evaluated 
accidents assume an initial condition of power operation for loss of 
main feedwater events. Providing for automatic initiation prior to 
criticality ensures operation within the bounds of design analysis. 
Previously evaluated accidents also assume the removal of decay 
heat, following loss of normal CCW flow, to be via the main steam 
safety relief valves to the atmosphere which eliminates the need for 
ECCW operability. Consequently, this change will not create the 
possibility of a new or different kind of accident from any kind of 
accident previously evaluated.
    (3) [The amendment request would not] involve a significant 
reduction in a margin of safety.
    The design basis of the EFW system is to supply feedwater to the 
steam generators in the event Main Feedwater is lost. By providing 
clarification that manual initiation circuitry is operable prior to 
exceeding an RCS temperature of 250 deg.F and automatic initiation 
circuitry, due to low main feedwater discharge pressure or low 
hydraulic oil pressure, is operable prior to criticality, there is 
no significant reduction in the margin of safety associated with 
this amendment request. The ECCW system is designed to provide a 
means to remove decay heat without a loss of secondary side 
inventory. However, analysis has shown that sufficient secondary 
side inventory exist, as specified by Technical Specification 3.4.4, 
to provide for coping with loss of station power events. 
Furthermore, even though this method of decay heat removal is 
desirable, Oconee PRA [probabilistic risk assessment] studies do not 
model the loss of ECCW for accident precursors since it is not 
required and margins of safety are not reduced if it is not 
available. Changes included within this amendment request clarify 
existing requirements for the operability of secondary system for 
decay heat removal based on previously evaluated accidents. As such, 
all margins of safety are preserved. Therefore, there will be no 
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, 
200 17th Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: June 2, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) sections 3.4.6.1 and 3.4.6.2 
related to reactor coolant system (RCS) operational leakage and leakage 
detection instrumentation. The proposed amendment would revise the TSs 
to be in accordance with the standard TSs in NUREG-1431 in so far as 
the plant-specific design will allow. The proposed changes relate to 
the limiting conditions for operation and the surveillance requirements 
for the four primary instruments used to detect RCS leakage. Changes 
are also proposed for the index and definition sections. A new TS, 
section 3/4.5.4, is proposed to address reactor coolant pump seal 
injection flow.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The probability of occurrence of a previously evaluated 
accident, i.e., loss of coolant accident (LOCA), is not increased 
because the ability of the plant operators to detect RCS leakage and 
take appropriate corrective action is not changed. The proposed 
change will continue to ensure that diverse means for detecting 
extremely small leaks are available to plant operators. In
    addition, the proposed amendment does not change the operational 
leakage limits. The seal injection flow limit is not affected by 
this proposed change. Due to these three factors, the probability of 
occurrence of a LOCA is not increased. The consequences of an 
accident previously evaluated are not significantly increased 
because the proposed changes do not affect the ability of the 
various safety systems to perform their intended function. The 
leakage detection monitors do not initiate any automatic function to 
mitigate the consequences of a LOCA. They provide an early 
indication of RCS leakage. The operational leakage limits are not 
affected by this proposed change and they do not initiate any 
automatic function to mitigate the consequences of a LOCA. The 
proposed change to the seal injection flow requirement will continue 
to ensure that ECCS flow will be as assumed in the accident 
analyses.
    Therefore, based on the continued ability of the leakage 
detection monitors and independent monitoring capabilities to detect 
extremely small leaks, the fact that this proposed amendment does 
not change the operational leakage limits, the seal injection flow 
limit is not affected by this proposed change, and that the proposed 
changes do not affect the ability of the various safety systems to 
perform their intended functions, this proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.2. Does the change create the 
possibility of a new or different kind of accident from any accident 
previously evaluated?
    The proposed amendment does not change the plant configuration 
in a way which introduces a new potential hazard to the plant. Since 
design requirement[s] continue to be met and the integrity of the 
RCS pressure boundary is not challenged, no new failure mode has 
been created. As a result, an accident which is different than any 
already evaluated in the Updated Final Safety Analysis Report 
(UFSAR) will not be created due to this change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change does not involve a significant reduction in 
a margin of safety since the operational leakage limits will not be 
affected. Continued plant operation will not be permitted if 
operational leakage exceeds the current technical specification 
limits. The operational leakage limits establish limits which ensure 
that any RCS leakage does not compromise safety. The protection of 
the RCS pressure boundary from degradation and the core form [from] 
inadequate cooling, in addition to preventing the accident analyses 
radiation release assumptions from being exceeded, is the main 
purpose of the operational leakage limits. The ability to detect and 
quantify operational leakage allows plant operators to perform 
actions to place the plant in a safe condition when leakage rate 
indicates possible RCS pressure boundary degradation. The proposed 
change will continue to ensure that diverse measurement means are 
available to provide the plant operators with an early indication of 
extremely small RCS leakage. Therefore, [the change is] allowing 
action to be taken to place the plant in a safe condition when RCS 
leakage indicates possible RCS pressure boundary leakage.
    The proposed addition of the separate seal injection 
specification will not change the flow limit on seal injection. The 
new specification will continue to ensure that seal injection flow 
is limited. This will ensure that sufficient flow to the reactor 
core is provided during accident conditions. The proposed 
elimination of the Mode 4 applicability, for seal injection flow 
specification, will not involve a significant reduction in the 
margin of safety since high seal injection flow is less critical as 
a result of the lower initial RCS pressure and decay heat removal 
requirements in Mode 4.
    Therefore, this proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B.F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
    Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg, 
Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., 
Washington, DC 20037.
    NRC Project Director: Walter R. Butler

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

    Date of amendment request: June 6, 1994
    Description of amendment request: The proposed admendment would 
revise the technical specifications (TSs) by relocating the seismic and 
meteorological monitoring instrumentation and their associated 
requirements from the TSs to the Waterford 3 updated final safety 
analysis report and plant procedures pursuant to the NRC final policy 
statement on TSs improvements for nuclear power reactors. The final 
policy statement was published in the Federal Register on Thursday, 
July 22, 1993.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change relocates Seismic and Meteorological 
Monitoring Instrumentation requirements from the TS to licensee 
controlled documents consistent with the NRC Policy Statement on 
Technical Specification Improvements. Criterion 1 of the Policy 
Statement indicates that the TS should include installed 
instrumentation that is used to detect, and indicate in the control 
room, a significant abnormal degradation of the reactor coolant 
pressure boundary. This criterion is intended to ensure that the TS 
control those instruments specifically installed to detect excessive 
reactor coolant system leakage. This criterion is not interpreted to 
include instrumentation used to detect precursors to reactor coolant 
pressure boundary leakage (e.g., loose parts monitor, seismic 
instrumentation, valve position indicators). Combustion Engineering 
and the NRC have previously determined that relocating Seismic and 
Meteorological Monitoring Instrumentation requirements from the TS 
does not affect any material condition of the plant that could 
directly contribute to causing or mitigating the effects of an 
accident.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    The proposed change will not involve any design change or 
modification to the plant. The proposed change will not alter the 
operation of the plant or the manner in which it is operated. Any 
subsequent change to the Seismic or Meteorological Monitoring 
Instrumentation requirements will undergo a review in accordance 
with the criteria of 10 CFR 50.59 to ensure that the change does not 
involve an unreviewed safety question.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed change will relocate Seismic and Meteorological 
Monitoring Instrumentation requirements from the TS to licensee 
controlled documents subject to the criteria of 10 CFR 50.59. The 
proposed change will have no adverse impact on any protective 
boundary or safety limit.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: June 22, 1994
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) to change three plant 
protection system (PPS) trip setpoints to be consistent with the 
current setpoint/uncertainty methodology being implemented at Waterford 
3. The change adjusts the affected TSs values in a more conservative 
direction.
    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:
    Implementing the proposed change will not affect any design 
basis accident. The revised trip and actuation setpoints are based 
upon the same analytical limits that form the basis for the current 
trip and actuation setpoints. The design basis for each trip and 
actuation setpoint was verified to be consistent with the 
appropriate accident analyses as part of the process of revising the 
PPS setpoint analysis. The proposed changes in trip and actuation 
setpoints are all in the conservative (away from the analytical 
limits) direction. Therefore, the proposed change will not involve a 
significant increase in the probability or consequences of any 
previously analyzed accident.
    Plant operation and the manner in which the plant is operated 
will not be altered as a result of implementing the proposed change 
since no new system or design change is being implemented. 
Therefore, the proposed change will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The current safety margins of the affected trip setpoints and 
allowable values is preserved by the proposed change. This is 
assured by retaining the current analytical limit for the affected 
parameters. Since the analytical limits are not affected and the 
total channel uncertainty is increased, the margin of safety for the 
affected trip setpoints and allowable values is preserved. 
Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: June 22, 1994
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) to replace the generic 
control room outside air intake (CROAI) radiation alarm/trip setpoint 
(less than or equal to 2x background) with a specific setpoint (less 
than or equal to 4.09E-5). The new setpoint is based on radioactive 
material concentrations in the control room not exceeding the derived 
air concentrations (DAC) occupational values listed in 10 CFR Part 20, 
Appendix B, Table 1, Column 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change replaces the current CROAI radiation monitor 
alarm/trip setpoint of less than or equal to 2x background with a 
fixed value independent of background radiation. The new setpoint 
will continue to provide protection to plant personnel such that 
occupational radiation exposure is maintained within the limits of 
10 CFR 20 during normal plant operation, anticipated operational 
occurrences or design basis accidents.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    The proposed change will replace the generic CROAI radiation 
monitor alarm/trip setpoint with a setpoint derived from a site-
specific calculation. The proposed change will not alter the 
operation of the plant or the manner in which it is operated.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed change will replace the current CROAI radiation 
monitor setpoint with a new setpoint that will ensure occupational 
radiation exposure will not exceed the DAC limits of 10 CFR 20. The 
proposed change has no adverse impact on protective boundaries, 
safety limits, or margin of safety.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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: July 19, 1994
    Description of amendment request: The licensee proposes to change 
Turkey Point Units 3 and 4 Technical Specifications by relocating 
cycle-specific parameter limits from the Technical Specifications to 
the Core Operating Limits Report (COLR). Presently, the parameter 
limits for Turkey Point Units 3 and 4 are calculated using NRC-approved 
methodologies. These limits are evaluated for every reload cycle and 
may be revised by a license amendment as appropriate, to reflect 
changes to cycle-specific variables.The curves to be relocated include 
(a) TS Figure 3.1-2, Rod Bank Insertion Limits versus Thermal Power 
curve, and (b) TS Figure 3.2-2, K(Z) Normalized FQ(Z) as a 
Function of Core Height curve.
    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 removal of cycle-specific Rod Bank Insertion limits and the 
K(Z) curve from the Turkey Point Units 3 and 4 Technical 
Specifications is administrative in nature and has no impact on the 
probability or consequences of any Design Bases Event (DBE) 
occurrences which was previously evaluated. The determination of the 
Rod Bank Insertion limits and K(Z) curve will be performed using 
methodology approved by the NRC and poses no significant increase in 
the probability or consequences of any accident previously 
evaluated.
    The Rod Bank Insertion limits and K(Z) curve will be evaluated 
every cycle to ensure proper compliance with the Updated Final 
Safety Analysis Report (UFSAR). These limits will be evaluated in 
accordance with 10 CFR Sec. 50.59, which ensures that the reload 
will not involve an increase in the probability of occurrences or 
consequences of an accident previously evaluated. 10 CFR Sec. 50.59 
(2) states that a proposed change involves an unreviewed safety 
question (i) if the probability of occurrence or the consequences of 
an accident or malfunction of equipment important to safety 
previously evaluated in the safety analysis report may be increased. 
Consequently, since any change to the reload core design analysis 
must be evaluated relative to the more restrictive evaluation 
criterion of 10 CFR Sec. 50.59, then 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.
    (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 removal of the Rod Bank Insertion limits and K(Z) curve from 
the Technical Specifications is administrative in nature and has no 
impact, nor does it contribute in any way to the possibility of a 
new or different kind of accident from any accident previously 
evaluated. No new accident scenarios, failure mechanisms or limiting 
single failure events are introduced as a result of the proposed 
change.
    The generation of the Rod Bank Insertion limits and K(Z) curve 
will be performed using NRC-approved methodology and are submitted 
to the NRC, as a revision to the COLR, to allow the NRC staff to 
trend. The Technical Specifications will continue to require 
operation within the core operating limits and appropriate actions 
will be taken if these limits are exceeded.
    10 CFR Sec. 50.59 permits a licensee to make changes in the 
facility as described in the safety analysis report without prior 
Commission approval, provided that the proposed changes does not 
involve an unreviewed safety question. 10 CFR Sec. 50.59 (2) states 
that a proposed change involves an unreviewed safety question (ii) 
if a possibility for an accident or malfunction of a different type 
than any evaluated previously in the safety analysis report may be 
created. Consequently, since any change to the reload core design 
analysis must be evaluated relative to the more restrictive 
evaluation criterion of 10 CFR Sec. 50.59, then 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 margin of safety is not affected by the removal of the Rod 
Bank Insertion limits and K(Z) curve from the Technical 
Specifications. The methodology for the reload core design analysis 
have been approved by the NRC and does not constitute a significant 
reduction in the margin of safety.
    The supporting Technical Specification values are defined by the 
accident analyses which are performed to conservatively bound the 
operating conditions defined by the Technical Specifications. The 
development of the limits for future reloads will continue to 
conform to the methodology described in NRC approved documentation. 
In addition, each future reload will involve a 10 CFR 50.59 review 
to assure that operation of the units within the cycle specific 
limits will not involve a reduction in a margin of safety. 10 CFR 
Sec. 50.59 (2) states that a proposed change involves an unreviewed 
safety question (iii) if the margin of safety as defined in the 
basis for any technical specification is reduced. Consequently, 
since any change to the reload core design analysis must be 
evaluated relative to the more restrictive evaluation criterion of 
10 CFR Sec. 50.59, then 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: Victor M. McCree, Acting

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: July 19, 1994
    Description of amendment request: The licensee proposes to change 
Turkey Point Units 3 and 4 Technical Specifications and its associated 
BASES, which address the maximum allowed reactor thermal power 
operation with inoperable main steam safety valves (MSSVs). 
Westinghouse issued Nuclear Safety Advisory Letter (NSAL) 94-001 which 
notified the licensee of a deficiency in the basis of the Turkey Point 
Technical Specification 3/4.7.1, which allows the plant to operate at 
reduced power levels with a specified number of MSSVs inoperable. This 
amendment request corrects the allowable power level with inoperable 
MSSVs.
    The licensee also proposed changes to the TS 3.7.1.1 applicability 
statement to indicate that, for mode 3 only, the actions are required 
when the Reactor Trip System breakers are in the closed position and 
the Control Rod Drive System is capable of rod withdrawal.
    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 changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The new power range neutron flux high setpoint values will ensure 
that the secondary side steam pressure will remain below 110 percent 
of the design value following a Loss of Load/Turbine Trip event, 
when one or more main steam safety valves (MSSVs) are declared 
inoperable. The proposed change will not impact the classification 
of the Loss of Load/Turbine Trip event as a Condition II probability 
event (faults of moderate frequency) per ANSI - N18.2, 1973. 
Accordingly, since the new power range neutron flux setpoints will 
maintain the capability of the MSSVs to perform their pressure 
relief function associated with a Loss of Load/Turbine Trip event, 
there will be no effect on the probability or consequences of an 
accident previously evaluated.
    In addition, the proposed change to the applicability statement 
of TS 3.7.1.1, will not effect the probability or consequences of an 
accident previously evaluated, since the proposed plant condition 
with the reactor trip breakers open and the rod control system not 
capable of withdrawing rods is an analyzed safe shutdown condition.
    (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 proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed changes do not involve any change to the configuration 
or method of operation of any plant equipment, and no new failure 
modes have been defined for any plant system or component. The new 
power range neutron flux high setpoints will maintain the capability 
of the MSSVs to perform their pressure relief function to ensure the 
secondary side steam design pressure is not exceeded following a 
Loss of Load/Turbine Trip event. Therefore, since the function of 
the MSSVs is unaffected by the proposed changes, the possibility of 
a new or different kind of accident from any accident previously 
evaluated is not created.
    In addition, the proposed change to the applicability statement 
of TS 3.7.1.1, will not create the possibility of a new or different 
kind of accident from any accident previously evaluated, since the 
proposed plant condition with the reactor trip breakers open and the 
rod control system not capable of withdrawing rods is an analyzed 
safe shutdown condition.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed changes to the Technical Specifications [do] not 
involve a significant reduction in a margin of safety. The algorithm 
methodology used to calculate the new power range neutron flux high 
setpoints is conservative and bounding since it is based on a number 
of inoperable MSSVs per loop; i.e., if only one MSSV in one loop is 
out of service, the applicable power range setpoint would be the 
same as if one MSSV in each loop were out of service. Another 
conservatism with the algorithm methodology is with the assumed 
minimum total steam flow rate capability of the operable MSSVs. The 
assumption is that if one or more MSSVs are inoperable per loop, the 
inoperable MSSVs are the largest capacity MSSVs, regardless of which 
capacity MSSVs are actually inoperable. Therefore, since the power 
range neutron flux setpoints calculated for the proposed changes 
using the algorithm methodology are more conservative and ensure the 
secondary side steam design pressure is not exceeded following a 
Loss of Load/Turbine Trip event, this proposed license amendment 
will not involve a significant reduction in a margin of safety.
    In addition, the proposed change to the applicability statement 
of TS 3.7.1.1, does not involve a significant reduction in a margin 
of safety, since the proposed plant condition with the reactor trip 
breakers open and the rod control system not capable of withdrawing 
rods is an analyzed safe shutdown condition.
    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: Victor M. McCree, Acting

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: July 19, 1994
    Description of amendment request: The licensee proposes to change 
Turkey Point Units 3 and 4 Technical Specifications (TS) by revising 
Surveillance Requirements 4.8.1.1.2e. and 4.8.1.1.2f., to delete the 
specific reference in the TS of the American Society for Testing and 
Materials (ASTM) testing standard being used to meet TS testing 
requirements. The Emergency Diesel Generator (EDG) fuel oil TS 
Surveillance Requirements will be replaced with a requirement to test 
the EDG fuel oil in accordance with the Turkey Point Units 3 and 4 
Diesel Fuel Oil Testing Program.
    The licensee proposes the addition of ACTION statements g. and h. 
of TS 3.8.1.1 to address the required action in the event the diesel 
fuel oil does not meet the Diesel Fuel Oil Testing Program limits.
    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 changes to the Technical Specifications will permit 
the Technical Specification required testing of Emergency Diesel 
Generator (EDG) fuel oil in accordance with the Turkey Point Units 3 
and 4 Diesel Fuel Oil Testing Program. The proposed change will 
permit FPL to use more recent editions of the American Society for 
Testing and Materials (ASTM) standards currently listed in Technical 
Specification Surveillance Requirements 4.8.1.1.2e. and 4.8.1.1.2f. 
Prior to changing the Diesel Fuel Oil Testing Program, the proposed 
change will be evaluated pursuant to Title 10 Code of Federal 
Regulations Sec. 50.59 (10 CFR Sec. 50.59), ``Changes, tests, and 
experiments.'' Title 10 CFR Sec. 50.59 permits a licensee to make 
changes in the procedures as described in the safety analysis report 
without prior Commission approval, provided that the proposed 
changes [do] not involve an unreviewed safety question.
    Title 10 CFR Sec. 50.59(a)(2) states that a proposed change 
involves an unreviewed safety question (i) if the probability of 
occurrence or the consequences of an accident or malfunction of 
equipment important to safety previously evaluated in the safety 
analysis report may be increased. Consequently, since any change to 
the Diesel Fuel Oil Testing Program, including the ASTM standard or 
ASTM edition standard to be used to evaluate EDG fuel oil 
acceptability, the change must be evaluated relative to the more 
restrictive evaluation criterion of 10 CFR Sec. 50.59, then 
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 EDG fuel oil 
TS Surveillance Requirements will be replaced with a requirement to 
test the EDG fuel oil in accordance with the Turkey Point Units 3 
and 4 Diesel Fuel Oil Testing Program.
    ACTION statement g. of TS 3.8.1.1 is added to address the 
required action in the event the new fuel oil properties do not meet 
the Diesel Fuel Oil Testing Program limits. A failure to meet the 
API gravity, kinematic viscosity, flash point or clarity limits is 
cause for rejecting the new fuel oil prior to the addition to the 
Diesel Fuel Oil Storage Tanks, but does not represent a failure to 
meet the Limiting Condition for Operation (LCO) of TS 3.8.1.1, since 
the new fuel oil has not been added to the storage tanks. Provided 
these new fuel oil properties are met subsequent to the addition of 
the new fuel oil to the storage tanks, 30 days is provided to 
complete the analyses of the other fuel oil properties specified in 
Table 1 of ASTM-D975-81, except sulfur which may be performed in 
accordance with ASTM-D1552-79 or ASTM-D2622-82. In the event the 
other new fuel oil properties specified in Table 1 of ASTM-D975-81 
are not met, ACTION statement g. of TS 3.8.1.1 provides an 
additional 30 days to meet the Diesel Fuel Oil Testing Program 
limits. This additional 30 day period is acceptable because the fuel 
oil properties of interest, even if they are not within limits, 
would not have an immediate effect on EDG operation.
    ACTION statement h. of TS 3.8.1.1 is added to address the 
required action in the event the stored fuel oil total particulates 
does not meet the Diesel Fuel Oil Testing Program limits. Fuel oil 
degradation during long term storage shows up as an increase in 
particulate, due mostly to oxidation. The presence of particulate 
does not mean the fuel oil will not burn properly in a diesel 
engine. The frequency for performing surveillance on stored fuel oil 
is based on stored fuel oil degradation trends which indicate that 
particulate concentration is unlikely to change significantly 
between surveillances.
    Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil 
Testing Program, FPL will need to determine if the proposed program 
change is at least as, if not more, effective, in detecting 
unsatisfactory fuel oil. The EDGs will thus continue to function as 
designed and the probability or consequences of previously evaluated 
accidents will be unaffected.
    (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 proposed changes to the Technical Specifications will permit 
the Technical Specification required testing of Emergency Diesel 
Generator fuel oil using more recent editions of the American 
Society for Testing and Materials standards listed in Technical 
Specification Surveillance Requirements 4.8.1.1.2e. and 4.8.1.1.2f. 
Prior to changing the edition of the previously approved ASTM 
standard being used to evaluate the EDG fuel oil, the proposed 
edition standard will be evaluated pursuant to 10 CFR Sec. 50.59, 
``Changes, tests, and experiments.'' Title 10 CFR Sec. 50.59 permits 
a licensee to make changes in the procedures as described in the 
safety analysis report without prior Commission approval, provided 
that the proposed changes does not involve an unreviewed safety 
question. Title 10 CFR Sec. 50.59(a)(2) states that a proposed 
change involves an unreviewed safety question (ii) if a possibility 
for an accident or malfunction of a different type than any 
evaluated previously in the safety analysis report may be created. 
Consequently, since any change to the edition of the ASTM standard 
to be used to evaluate EDG fuel oil acceptability must be evaluated 
relative to the more restrictive evaluation criterion of 10 CFR 
Sec. 50.59, then 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.
    ACTION statement g. of TS 3.8.1.1 is added to address the 
required action in the event the new fuel oil properties do not meet 
the Diesel Fuel Oil Testing Program limits. A failure to meet the 
API gravity, kinematic viscosity, flash point or clarity limits is 
cause for rejecting the new fuel oil prior to the addition to the 
Diesel Fuel Oil Storage Tanks, but does not represent a failure to 
meet the Limiting Condition for Operation (LCO) of TS 3.8.1.1, since 
the new fuel oil has not been added to the storage tanks. Provided 
these new fuel oil properties are met subsequent to the addition of 
the new fuel oil to the storage tanks, 30 days is provided to 
complete the analyses of the other fuel oil properties specified in 
Table 1 of ASTM-D975-81, except sulfur which may be performed in 
accordance with ASTM-D1552-79 or ASTM-D2622-82. In the event the 
other new fuel oil properties specified in Table 1 of ASTM-D975-81 
are not met, ACTION statement g. of TS 3.8.1.1 provides an 
additional 30 days to meet the Diesel Fuel Oil Testing Program 
limits. This additional 30 day period is acceptable because the fuel 
oil properties of interest, even if they are not within limits, 
would not have an immediate effect on EDG operation.
    ACTION statement h. of TS 3.8.1.1 is added to address the 
required action in the event the stored fuel oil total particulates 
[do] not meet the Diesel Fuel Oil Testing Program limits. Fuel oil 
degradation during long term storage shows up as an increase in 
particulate, due mostly to oxidation. The presence of particulate 
does not mean the fuel oil will not burn properly in a diesel 
engine. The frequency for performing surveillance on stored fuel oil 
is based on stored fuel oil degradation trends which indicate that 
particulate concentration is unlikely to change significantly 
between surveillances.
    Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil 
Testing Program, FPL will need to determine if the proposed program 
change is at least as, if not more, effective, in detecting 
unsatisfactory fuel oil. Since the proposed changes do not involve a 
change in the design of any plant system or component, and since the 
proposed changes will need to evaluate the effect of any ASTM 
standard edition change on the level of EDG reliability, the change 
proposed will 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 proposed changes to the Technical Specifications will permit 
the Technical Specification required testing of Emergency Diesel 
Generator (EDG) fuel oil using more recent editions of the American 
Society for Testing and Materials (ASTM) standards listed in 
Technical Specification Surveillance Requirements 4.8.1.1.2e. and 
4.8.1.1.2f. Prior to changing the edition of the previously approved 
ASTM standard being used to evaluate the EDG fuel oil, the proposed 
edition standard will be evaluated pursuant to 10 CFR Sec. 50.59, 
``Changes, tests, and experiments.'' Title 10 CFR Sec. 50.59 permits 
a licensee to make changes in the procedures as described in the 
safety analysis report without prior Commission approval, provided 
that the proposed changes [do] not involve an unreviewed safety 
question. Title 10 CFR Sec. 50.59(a)(2) states that a proposed 
change involves an unreviewed safety question (iii) if the margin of 
safety as defined in the basis for any technical specification is 
reduced. Consequently, since any change to the edition of the ASTM 
standard to be used to evaluate EDG fuel oil acceptability must be 
evaluated relative to the more restrictive evaluation criterion of 
10 CFR Sec. 50.59, then operation of the facility in accordance with 
the proposed amendments would not involve a significant reduction in 
a margin of safety.
    ACTION statement g. of TS 3.8.1.1 is added to address the 
required action in the event the new fuel oil properties do not meet 
the Diesel Fuel Oil Testing Program limits. A failure to meet the 
API gravity, kinematic viscosity, flash point or clarity limits is 
cause for rejecting the new fuel oil prior to the addition to the 
Diesel Fuel Oil Storage Tanks, but does not represent a failure to 
meet the Limiting Condition for Operation (LCO) of TS 3.8.1.1, since 
the new fuel oil has not been added to the storage tanks. Provided 
these new fuel oil properties are met subsequent to the addition of 
the new fuel oil to the storage tanks, 30 days is provided to 
complete the analyses of the other fuel oil properties specified in 
Table 1 of ASTM-D975-81, except sulfur which may be performed in 
accordance with ASTM-D1552-79 or ASTM-D2622-82. In the event the 
other new fuel oil properties specified in Table 1 of ASTM-D975-81 
are not met, ACTION statement g. of TS 3.8.1.1 provides an 
additional 30 days to meet the Diesel Fuel Oil Testing Program 
limits. This additional 30 day period is acceptable because the fuel 
oil properties of interest, even if they are not within limits, 
would not have an immediate effect on EDG operation.
    ACTION statement h. of TS 3.8.1.1 is added to address the 
required action in the event the stored fuel oil total particulates 
[do] not meet the Diesel Fuel Oil Testing Program limits. Fuel oil 
degradation during long term storage shows up as an increase in 
particulate, due mostly to oxidation. The presence of particulate 
does not mean the fuel oil will not burn properly in a diesel 
engine. The frequency for performing surveillance on stored fuel oil 
is based on stored fuel oil degradation trends which indicate that 
particulate concentration is unlikely to change significantly 
between surveillances.
    Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil 
Testing Program, FPL will need to determine if the proposed program 
change is at least as, if not more, effective, in detecting 
unsatisfactory fuel oil. Since the proposed changes will require a 
safety evaluation to assure that the reliability of the EDGs using 
fuel oil tested in accordance with the different ASTM standard 
edition maintains the current margin of safety, the proposed 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 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: Victor M. McCree, Acting

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, 
Burke County, Georgia

    Date of amendment request: April 28, 1994
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3/4.8.1.1, ``AC Sources 
Operating,'' and the associated TS Bases for demonstrating the 
operability of the diesel generators (DGs), based upon three NRC 
guidelines:
    A. Generic Letter (GL) 93-05, ``Line-Item Technical Specifications 
Improvements to Reduce Surveillance Requirements for Testing During 
Power Operation.''
    1. Delete from TS action statement a the requirement to test the 
DGs in the event of an inoperable offsite circuit.
    2. Eliminate from TS action statement b the need to test the 
operable DG if the other DG became inoperable due to an inoperable 
support system or an independently testable component in addition to 
the existing provision excluding preplanned preventive maintenance or 
testing. Furthermore, ifthe operable DG must be tested, it would be 
tested within 8 hours (rather than 24 hours) unless the absence of any 
potential common mode failure for the remaining DG is demonstrated.
    3. Also eliminate from TS action statement c the need to test the 
operable DG if the other DG became inoperable due to an inoperable 
support system or an independently testable component in addition to 
the existing provision excluding preplanned preventive maintenance or 
testing. In addition, the operable DG would not have to be tested if 
the absence of any potential common mode failure for the remaining DG 
is demonstrated. A reference to TS action statement a would be deleted 
because of the proposed change to TS action statement a described 
above.
    4. Eliminate from TS action statement e the need to test the DGs 
when two offsite circuits are inoperable.
    5. Revise TS 4.8.1.1.2.g.2 to allow the DG to be gradually loaded, 
as opposed to a fast loading of 60 seconds or less to an indicated 
value of 6100-7000 kw. This change would extend gradual loading of DGs 
(that GL 93-05 recommends for routine monthly surveillance) to the 6-
month surveillance.
    B. Regulatory Guide (RG) 1.9, Revision 3, ``Selection, Design, 
Qualification, and Testing of Emergency Diesel Generator Units Used as 
Class 1E Onsite Electric Power Systems at Nuclear Power Plants,'' 
(insofar as this guide relates to reducing DG stress and wear due to 
testing and the elimination of certain reporting requirements).
    1. Incorporate into TS 4.8.1.1.2.a.4 the provision to perform 
routine monthly testing by gradually accelerating the DG to operating 
speed, rather than requiring the DG to attain rated voltage and 
frequency within 11.4 seconds. As a direct result of this proposed 
change, TS action statements b, c, and f would be revised to reference 
TS 4.8.1.1.2.g.1 instead of TS 4.8.1.1.2.a.4 in the event that an 
operable DG must be tested when the other DG is inoperable. This has 
the effect of requiring the operable DG to be fast-started for testing 
pursuant to the action statement.
    2. In TS 4.8.1.1.2.h.7, separate the 24-hour endurance run from the 
hot restart test. As a result, create new TS 4.8.1.1.2.h.8 to require 
the hot restart test. The DG would be operated for a minimum of 2 hours 
at a load of 6800-7000 kw, and the DG would be shut down. Within 5 
minutes of shutdown, the DG would be restarted and required to attain 
rated voltage and frequency within 11.4 seconds. Delete existing TS 
footnote  (which provides for reperforming the hot 
restart test without repeating the 24-hour endurance test) which is no 
longer required. Renumber existing TSs 4.8.1.1.2.h.8, .9, .10, .11, and 
.12 to accommodate the addition of new TS 4.8.1.1.2.h.8.
    3. Delete TS 4.8.1.1.3, ``Reports.'' (This is also in accordance 
with the Improved Technical Specifications, Revision 0, dated September 
28, 1992).
    4. In TS 4.8.1.2, ``A. C. Sources Shutdown,'' delete the reference 
to deleted TS 4.8.1.1.3.
    C. NUMARC 87-00, Revision 1, ``Guidelines and Technical Bases for 
NUMARC Initiatives Addressing Station Blackout at Light Water 
Reactors,'' (insofar as it relates to the test frequency for a problem 
DG). Specifically, TS Table 4.8-1, ``Diesel Generator Test Schedule,'' 
would be revised to incorporate the test schedule of Section D.2.4.4 of 
Appendix D to NUMARC 87-00, Revision 1. Under the proposed schedule, 
testing pursuant to TS 4.8.1.1.2.a would be conducted monthly provided 
the number of valid failures in the last 25 demands for a given DG is 
no more than 3. If the number of valid failures is 4 or more, testing 
would be conducted at least once per 7 days (but at intervals of no 
less than 24 hours) until 7 consecutive failure-free starts from 
standby conditions and load-run demands have been performed. Note that 
both NUMARC 87-00, Revision 1, and RG 1.9, Revision 3, use and define 
the terms ``start demand, start failure, load-run demand, and load-run 
failure'' rather than the old RG 1.108 terminology of valid tests. In 
fact, Section D.2.4.4 of Appendix D to NUMARC 87-00 refers to the last 
25 ``demands'' rather than tests. Therefore, the proposed change to TS 
Table 4.8-1 would count valid failures in terms of demands rather than 
valid tests. The criteria for determining the number of valid failures 
and demands would be in accordance with RG 1.9, Revision 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed changes affect the required actions in response to 
inoperable offsite and onsite ac sources, surveillance requirements 
for the emergency diesel generators, and reporting requirements for 
diesel generator failures. The proposed changes are based on the 
recommendations of Regulatory Guide 1.9, Revision 3, NUMARC 87-00, 
Revision 1, and Generic Letter 93-05. They are expected to result in 
improvements in diesel generator testing and failure reporting and 
reduce diesel generator aging due to excessive testing. As such, the 
proposed changes should result in improved diesel generator 
reliability, thereby providing additional assurance that the diesel 
generators will be capable of performing their safety function. 
Therefore, the proposed changes will not significantly increase the 
probability or consequences of any accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed changes affect the action and surveillance 
requirements for the onsite and offsite ac sources. Accordingly, the 
proposed changes do not involve any change to the configuration or 
method of operation of any plant equipment, and no new failure modes 
have been defined for any plant system or component nor has any new 
limiting failure been identified as a result of the proposed 
changes. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety. The proposed changes are based on existing 
regulatory guidance. Under the proposed changes, the emergency 
diesel generators will remain capable of performing their safety 
function, and the effects of aging on the diesel generators will be 
reduced by eliminating unnecessary testing. Therefore, the proposed 
changes do not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia 30830.
    Attorney for licensee: Mr. Arthur H. Domby, Esquire, Troutman 
Sanders, Nations Bank Plaza, Suite 5200, 600 Peachtree Street, NE., 
Atlanta, Georgia 30308-2210.
    NRC Project Director: Herbert N. Berkow

GPU Nuclear Corporation, Docket No. 50-320, Three Mile Island 
Nuclear Station, Unit No. 2, (TMI-2), Dauphin County, Pennsylvania

    Date of amendment request: October 9, 1991.
    Description of amendment request: Facility Operating License No. 
DPR-73, a possession only license for the TMI-2 facility, held by 
General Public Utilities Nuclear Corporation (GPU Nuclear), expires 
November 4, 2009. The proposed amendment would extend the expiration 
date of
    Facility Operating License No. DPR-73 for TMI-2 to April 19, 2014. 
No other changes to the license or the Technical Specifications are 
proposed.
    The TMI-2 facility is currently in long term storage. GPU Nuclear, 
the licensee, has named this storage period Post Defueling Monitored 
Storage or PDMS. The licensee plans to maintain TMI-2 in PDMS until 
Three Mile Island Nuclear Station Unit No. 1 (TMI-1) permanently ceases 
operation, at which time both TMI-1 and TMI-2 will be decommissioned 
simultaneously. The TMI-1 Operating License expires on April 19, 2014.
    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:
    10 CFR 50.92 provides the criteria which the Commission uses to 
perform a No Significant Hazards Consideration. 10 CFR 50.92 states 
that an amendment to a facility license involves No Significant 
Hazards 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.
    The proposed modification of the expiration date of the TMI-2 
License does not involve any physical changes to the facility. All 
that is involved is an extension of the time TMI-2 would be in a 
monitored storage condition. Based on this, GPU Nuclear concludes 
that the proposed change does not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated. Accident 
evaluations for PDMS are provided in the PDMS Safety Analysis Report 
(SAR) and the PDMS Final Programmatic Environmental Impact 
Statement, Supplement 3 (PEIS) dated August 1989. These documents 
evaluated monitored storage of TMI-2 for extended periods of time 
and provide for surveillances to ensure monitored storage conditions 
are appropriately maintained. The PDMS PEIS specifically evaluated 
monitored storage until 2014. No evaluated accident has a 
probability or consequence that are increased significantly during 
the period 2009 to 2014 over the period before 2009. Therefore, it 
can be concluded that this change does 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 accident previously evaluated. As previously stated, the 
act of modifying the expiration date of the TMI-2 License does not 
involve any physical changes to the facility and therefore, the 
possibility of a new or different kind of accident is not created.
    3. Involve a significant reduction in the margin of safety 
during PDMS. The surveillances identified in the PDMS SAR will be 
performed to ensure that the facility is maintained in the condition 
defined by the SAR. These conditions and surveillances will continue 
to apply during the extended license. Therefore, there will not be a 
reduction in the margin of safety.
    Based on the above analysis, it is concluded that the proposed 
changes involve No Significant Hazards Consideration as defined by 
10 CFR 50.92.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601 Harrisburg, Pennsylvania 17105
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW, Washington, D.C. 20037
    NRC Project Director: Seymour H. Weiss

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

    Date of amendment request: June 30, 1994
    Description of amendment request: The proposed amendment would 
clarify the requirement for the audit of conformance to Technical 
Specifications, delete the requirement for Safety Committee oversight 
of the Emergency Plan and Security Plan and allow designation by the 
Plant Superintendent signature authority for procedure approval.
    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 amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. No physical changes will result from this amendment. 
These changes revise audit requirements and procedure approval 
requirements. The subject audits will still be performed to provide 
assurance of conformance to the requirements, and the procedures 
will still receive adequate technical reviews by the cognizant 
departments while relieving the Plant Superintendent-Nuclear of the 
administrative burden of signing each procedure revision.
    2) The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. No physical changes will result from this amendment. 
These changes revise audit requirements and procedure approval 
requirements. The subject audits will still be performed to provide 
assurance of conformance to the requirements, and the procedures 
will still receive adequate technical reviews by the cognizant 
departments while relieving the Plant Superintendent-Nuclear of the 
administrative burden of signing each procedure revision.
    3) The proposed amendment does not involve a significant 
reduction in a margin of safety. No physical changes will result 
from this amendment. These changes revise audit requirements and 
procedure approval requirements. The subject audits will still be 
performed to provide assurance of conformance to the requirements, 
and the procedures will still receive adequate technical reviews by 
the cognizant departments while relieving the Plant Superintendent-
Nuclear of the administrative burden of signing each procedure 
revision.
    The NRC staff has reviewed the licensee's analysis and, based on 
thisreview, 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: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401
    Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea, 
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC 
20036
    NRC Project Director: John N. Hannon

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

    Date of amendment request: June 30, 1994
    Description of amendment request: The proposed amendment would add 
Operability Requirements, Limiting Conditions for Operations (LCO) and 
Surveillance Requirements for the Control Building Chillers.
    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 because the requested revisions do not affect 
any FSAR analysis involving these systems.
    The proposed revision only adds LIMITING CONDITIONS for 
OPERATION (LCO) and Surveillance Requirements (SR) for the Control 
Building Chillers. These additions will provide assurance that the 
affected systems will be OPERABLE when required.
    2) The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated because there is no equipment or design change associated 
with this change. The proposed amendment only adds LCOs and SRs for 
the Control Building Chillers.
    3) The proposed amendment will not involve any reduction in a 
margin of safety. The safety function of the Control Building 
Chillers is to remove the design basis heat load under all normal 
and emergency conditions. The addition of LCOs and SRs for the 
Control Building Chillers ensures they will be OPERABLE when 
required.
    The NRC staff has reviewed the licensee's analysis and, based on 
thisreview, 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: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401
    Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea, 
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC 
20036
    NRC Project Director: John N. Hannon

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

    Date of amendment request: June 30, 1994
    Description of amendment request: The proposed amendment would 
modify the surveillance testing of the Emergency Service Water (ESW) 
system by deleting the flow rate test and the requirement to test the 
pumps each week when river water temperature exceeds 80 deg.F and by 
adding a surveillance regarding the Cedar River (Ultimate Heat Sink) 
water temperature.
    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. No physical changes will result from this 
amendment. The ESW system will still maintain its ability to support 
various safety related equipment which is designed to mitigate the 
consequences of certain accidents and transients. These safety 
related systems play no part in the probability of these accidents 
or transients occurring. Since the ESW system will continue to fully 
support the cooling requirements of the safety related equipment 
which mitigate the consequences of certain accidents and transients, 
this amendment will not affect the consequences of these accidents 
and transients. The re-analysis of the component heat loads assumed 
worst case conditions and involved conservative assumptions. Our 
continuing program for monitoring heat exchanger performance, which 
was established in response to Generic Letter 89-13, ``Service Water 
System Problems Affecting Safety-Related Equipment,'' will continue 
to verify that the individual components are capable of performing 
their design function. Therefore, the proposed amendment does not 
involve a change in the probability or consequences of an accident 
previously evaluated.
    2) The proposed license amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated. The safety function of the ESW system is 
unchanged. The revised flow requirements for the system have been 
established using conservative assumptions and worst case heat loads 
and are appropriately documented in the FSAR and plant procedures. 
This amendment will result in no physical changes to the ESW system 
and therefore, will not affect its ability to continue to provide 
reliable cooling water. Consequently, the proposed license amendment 
does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3) The proposed amendment will not reduce the margin of safety. 
The re-analysis of the ESW flow rate requirements and component heat 
loads was performed using conservative assumptions and maximum 
component heat loads. The actual operation of the ESW system will 
not be changed. Any degradation of ESW pump performance would be 
detected by the IST program which requires quarterly testing of 
these pumps and monitoring of the pump's differential pressure and 
flow. Deleting the requirement to perform the surveillance each week 
when river water temperature exceeds 80 deg.F will not reduce the 
margin of safety because even at a river water temperature of 
95 deg.F, the required ESW flow to supply all the branches is well 
below the normal system flow rate of approximately 1100 gpm. 
Deleting the weekly surveillance will eliminate unnecessary testing 
of the ESW pumps, thereby reducing wear on the pumps. Adding a 
surveillance requirement for river water temperature will formalize 
the recording of water temperature every hour to assure acceptable 
ESW performance.
    The NRC staff has reviewed the licensee's analysis and, based on 
thisreview, 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: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401
    Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea, 
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC 
20036
    NRC Project Director: John N. Hannon

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

    Date of amendment request: June 30, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.2.7.1, ``Primary Coolant System 
Pressure Isolation Valves.'' Specifically, TS Table 3.2.7.1, ``Primary 
Coolant System Pressure Isolation Valves,'' would be revised to add 
Shutdown Cooling System (SCS) check valves 38-165, 166, 167, 168, 169, 
170, 171, and 172 each with a maximum allowable leakage rate of less 
than or equal to 0.375 gpm. The proposed amendment would add the check 
valves in lieu of replacing the SCS isolation valves with ones that are 
10 CFR Part 50, Appendix J, Type C air testable. The added check valves 
would provide high pressure/low pressure interfaces between the high 
pressure Reactor Coolant System and the low pressure Core Spray System. 
The addition of the check valves will allow utilization of the Core 
Spray System as a seal water system for sealing the Shutdown Cooling 
isolation valves as permitted by Section III.C.3 of 10 CFR Part 50, 
Appendix J.
    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 operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change requires the addition of Primary Coolant 
System pressure isolation valves for the prevention of an 
intersystem LOCA [loss-of-coolant accident]. The proposed addition 
does not affect operation of either the Shutdown Cooling or Core 
Spray Systems. These changes do not alter any accident initiators or 
precursors and therefore does not affect the probability of a 
previously evaluated accident.
    Testing these valves in accordance with Specification 3.2.7.1 
provides assurance that the Core Spray System will not be damaged by 
an overpressurization event which could lead to potential loss of 
integrity of the system and subsequent release of radioactivity. 
Thus, the addition of the valves would not increase the consequences 
of any accident. Therefore, the operation of Nine Mile Point Unit 1, 
in accordance with the proposed amendment, will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed addition of Primary Coolant System pressure 
isolation valves, although a physical change, does not alter the 
initial conditions used for any design basis accident. The check 
valves provide the high pressure/low pressure isolation between the 
Reactor Coolant and Core Spray Systems. These valves will be subject 
to leak rate testing in accordance with Specification 3.2.7.1. This 
ensures that an intersystem LOCA is prevented. The proposed change 
has no effect on operation of either the Shutdown Cooling or Core 
Spray Systems. Therefore, the design capabilities of these systems 
are not challenged in a manner previously assessed so as to create 
the possibility of a new or different kind of accident. Accordingly, 
operation of Nine Mile Point Unit 1, in accordance with the proposed 
amendment, will not create the possibility of a new or different 
kind of accident from any accident previously analyzed.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The proposed change which requires the addition of Primary 
Coolant System pressure isolation valves, ensures proper isolation 
of a high pressure/low pressure interface between the Reactor 
Coolant and Core Spray Systems. The pressure isolation valves will 
be leak tested in accordance with Specification 3.2.7.1. This 
provides assurance that the Core Spray System will not be damaged by 
an overpressurization event and will not result in loss of integrity 
of the system. Thus, the results of any event previously analyzed 
remains unchanged. Therefore, the operation of Nine Mile Point Unit 
1, in accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Michael L. Boyle

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

    Date of amendment request: April 18, 1994
    Description of amendment request: The proposed change will revise 
the current surveillance frequency which verifies area temperature 
limits at least once per 12 hours. The revised surveillance requirement 
will verify area temperature limits at least once per 7 days when the 
data-logger alarm is operable, or at least once per 12 hours when the 
data-logger alarm is inoperable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    ...The proposed change does not involve an SHC [significant 
hazards consideration] because the change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change reduces the frequency at which area 
temperature monitoring must be verified when the temperature data-
logger alarm function is operable. For conditions where the 
temperature data-logger alarm function is inoperable, the frequency 
at which normal ambient temperature is verified remains unchanged. 
In addition, the proposed change does not affect any system, 
equipment, or component credited in any previous accident 
evaluation, any environmental qualification or post-accident 
profiles. Therefore, the proposed change 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.
    The proposed change does not alter or affect the design, 
function, failure mode, or operation of the plant. There is no 
change to the way in which the plant is operated and, therefore, no 
increase in the probability of plant operation with any area 
temperature outside of its limits. Therefore, this change does not 
create the possibility of a new or different kind of accident from 
those previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change does not challenge or affect the performance 
of any of the protective boundaries, revise temperature limits in 
the technical specifications, or perform any modifications that 
would increase the likelihood of technical specification temperature 
limits being exceeded. The proposed change requires the data-logger 
alarm function to be operable in order to relax the surveillance 
frequency. This alarm function provides continuous monitoring that 
would detect temperature excursions prior to the current 
surveillance which does not credit operability of the data-logger 
alarm function. Also, the proposed change does not increase the 
interval for which temperatures could exceed technical specification 
limits. Therefore, the proposed change does not cause a reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.
    Attorney for licensee: Gerald Garfield, Esquire, Day, Berry & 
Howard, City Place, Hartford, Connecticut 06103-3499.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: June 30, 1994
    Description of amendment request: The proposed change revises the 
Technical Specifications to change the trip setpoint for the 4kV bus 
undervoltage relay (for the grid degraded voltage) from its current 
value of [greater than or equal to] 3710 volts to its new setting of 
[greater than or equal to] 3730 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 change does not involve an SHC [significant 
hazards consideration] because the change does not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change involves the modification of the 
undervoltage relay setpoint from 3710V to 3730V. The protection 
provided by this system in unaffected and is still in accordance 
with the guidance provided in NRC Branch Technical Position PSB-1. 
This refinement increases the Technical Specification minimum trip 
setpoint for the degraded voltage relays on the 4kV safety buses. It 
does not detrimentally affect the safe operation of the plant, nor 
does this proposed modification increase the probability or 
consequences of an accident previously evaluated. The actual trip 
setpoints of the subject relays do not require any changes and are 
currently conservatively set at 3745V. The allowable value of 
[greater than or equal to] 3706V remains unchanged. This slightly 
higher than required setting was chosen by NNECO to provide added 
margin should an undervoltage condition be present. This higher 
setting will not cause more actuations of the ESF [engineered safety 
feature] systems.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The undervoltage protection system is provided to address the 
concerns identified in NRC Branch Technical Position PSB-1 by 
providing a scheme to detect the loss of offsite power at the class 
1E buses, and a second level of undervoltage protection to protect 
class 1E equipment. The change in the setpoint will not affect the 
ability of this circuitry to detect a loss of offsite power or to 
respond to an undervoltage condition.
    Since the equipment will operate as previously described in the 
FSAR [Final Safety Analysis Report], and there are no physical plant 
modifications required (the current setting at the undervoltage 
relay is 3745V), the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    These relays do not cause a loss of offsite power, nor do they 
cause a degraded voltage condition. These relays react to conditions 
that have been placed upon the Plant. In the event that a degraded 
voltage condition exists on the 4kV safety buses, alarms in the 
control room alert the operators of this condition. In addition, the 
Connecticut Valley Electric Exchange (CONVEX), the system dispatch 
center for generation and VAR/voltage control, is aware of the 
minimum voltage requirements for the three nuclear plants at the 
Millstone station and has a minimum target switchyard voltage of 
345kV. Under normal operation conditions the switchyard voltage 
would have to degrade below 328kV before one of the 4kV safety buses 
would start to enter the degraded voltage level and trip the 
degraded voltage circuit. These administrative controls help 
preclude a degraded voltage condition on the 4kV safety buses prior 
to actuation of the degraded voltage protection circuits.
    The proposed change of the 4kV degraded voltage minimum trip 
setpoint to 3730V from 3710V will not result in any physical relay 
setting change. The existing trip setting for the 4kV degraded 
voltage relays have been conservatively set at 3745V, while the 
existing allowable value remains unchanged at 3706V.
    The response times or actuation logic of the degraded voltage 
protection circuit remains unaffected, therefore revising the trip 
setpoint value in the Technical Specifications 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: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.
    Attorney for licensee: Gerald Garfield, Esquire, Day, Berry & 
Howard, City Place, Hartford, Connecticut 06103-3499.
    NRC Project Director: John F. Stolz

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: January 29, 1993, as revised June 15, 
1994.
    Description of amendment requests: The proposed amendments would 
change core exit thermocouple action statements.
    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.
    The purpose of the post accident monitoring equipment is to 
display unit variables that provide information required by the 
control room operators during accident situations and as such help 
limit the consequences of an accident. The proposed changes, which 
will allow continued plant operation with less than four core exit 
thermocouples per core quadrant, have no impact on the probability 
of an accident because they are only used in response to accident 
situations.
    Continued plant operation with the core exit thermocouple system 
in the degraded condition as allowed by the proposed core exit 
thermocouple action statements would not affect the operators 
ability to monitor for inadequate core cooling following an 
accident. At least two core exit thermocouples would be operable per 
core quadrant, a minimum of four thermocouples would be available in 
the center region of the core and at least one thermocouple would be 
available in each quadrant of the outside core region. The smaller 
size of the Prairie Island core, and therefore higher density of 
thermocouples per unit of core area, provides additional assurance 
that core exit temperatures can be adequately monitored with a 
reduced number of core exit thermocouples.
    Alternate means of monitoring for inadequate core cooling would 
also be available. These include the reactor vessel water level 
indication system, the subcooling margin monitors and wide range 
reactor coolant system temperature.
    The combination of the remaining operable core exit 
thermocouples and the alternate monitoring capability will ensure 
that the operators ability to identify inadequate core cooling in a 
timely manner and take appropriate corrective action will not be 
impaired, and therefore; the proposed changes will have no 
significant impact on the consequences of an accident.
    The core exit thermocouples perform no active role in the 
mitigation of an accident. Their inoperability will not affect the 
operability of any engineered safety features equipment or that 
equipments ability to mitigate the consequences of an accident.
    Therefore, for the reasons discussed above, the proposed changes 
will not significantly affect 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 
analyzed.
    There are no new failure modes or mechanisms associated with the 
proposed changes. The proposed changes do not involve any 
modification of plant equipment or any changes in operational 
limits. The proposed changes only modify the requirements for 
instrumentation used to monitor plant parameters during an accident. 
The core exit thermocouples are passive monitoring devices, their 
failure or inoperability cannot result in a plant accident of any 
kind.
    Therefore, for the reasons discussed above, the proposed changes 
do not create the possibility of a new or different kind of accident 
from any previously evaluated, and the accident analyses presented 
in the Updated Safety Analysis Report will remain bounding.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    Continued plant operation with the core exit thermocouple system 
in the degraded condition as allowed by the proposed core exit 
thermocouple action statements would not affect the operators 
ability to monitor for inadequate core cooling following an 
accident. At least two core exit thermocouples would be operable per 
core quadrant, a minimum of four thermocouples would be available in 
the center region of the core and at least one thermocouple would be 
available in each quadrant of the outside core region. The smaller 
size of the Prairie Island core, and therefore higher density of 
thermocouples per unit of core area, provides additional assurance 
that core exit temperatures can be adequately monitored with a 
reduced number of core exit thermocouples.
    Alternative means of monitoring for inadequate core cooling 
would also be available. These include the reactor vessel water 
level indication system, the subcooling margin monitors and wide 
range reactor coolant system temperature.
    The combination of the remaining operable core exit 
thermocouples and the alternate monitoring capability will ensure 
that the operators ability to identify inadequate core cooling in a 
timely manner and take appropriate corrective action will no be 
impaired.
    Therefore, for the reasons discussed above, the proposed changes 
will not result in any reduction in the plant's 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: 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: L. B. Marsh

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

    Date of amendment request: June 30, 1994
    Description of amendment request: This amendment would relocate 
selected recirculation and control rod block instrumentation setpoints 
from Technical Specifications (TS) Table 3.3.6-2, and Section 3/4.4.1 
to the Core Operating Limits Report (COLR), thereby revising TS Section 
6.9.1.9 to document relocation of these items into the COLR.
    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 TS change proposed is the relocation of the recirculation 
pump Motor-Generator (MG) set mechanical and electrical stop and 
control rod block recirculation flow upscale trip setpoint values to 
the COLR. No physical plant equipment change is proposed. The TS 
requirements for the setpoints and the associated surveillance 
requirements remain unchanged. Only the location of the setpoint 
values will be changed. The subject setpoint values will become 
cycle depend[e]nt and will be determined by NRC approved methods, as 
are the balance of setpoints and thermal limits found in the COLR. 
However, the subject setpoint values are not modified as part of 
this TS change.
    Therefore, the proposed TS change does not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The TS changes proposed are the relocation of the recirculation 
pump MG set mechanical and electrical stop and control rod block 
recirculation flow upscale trip setpoint values to the COLR. No 
physical plant equipment change is part of the proposed TS changes. 
The TS LCOs and surveillance requirements remain unchanged. The only 
change proposed is the relocation of the subject setpoint values as 
noted above. These setpoint values have been determined in 
accordance with previously NRC approved methods and assure 
sufficient operating margins in accordance with existing core design 
methodology. Therefore, the proposed TS changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The following TS BASES were reviewed for potential reduction in 
the margin of safety:
    3/4.2 Power Distribution Limits
    3/4.3.6 Control Rod Block Instrumentation
    3/4.4.1 Recirculation System
    The margin of safety, as defined in the TS BASES, will not be 
reduced. The proposed TS changes do not affect existing accident 
analyses or design assumptions, nor do they impact any safety limits 
of the plant, since they are administrative in nature.
    Therefore, the proposed TS changes do not involve a reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
Pennsylvania 19101
    NRC Project Director: Charles L. Miller

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

    Date of amendment request: June 23, 1994
    Description of amendment request: This amendment will change the 
Technical Specification 4.0.5 for each unit to reflect NRC's policy 
with respect to relief requests for the inservice inspection programs. 
Specifically, the change would clarify the fact that relief requests 
for impracticable testing or surveillance requirements can be 
implemented prior to the Commission approval of such requests.
    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 are administrative in nature in that the 
changes eliminate any possibility of misinterpretation of the ASME 
Code requirements that allow for a utility to submit relief requests 
to the Commission within one year and allows for the implementation 
of these request[s] prior to Commission review and approval. The 
relief requests are based on and provide for alternative testing 
based on industry practice that provides an equivalent level of 
quality and safety as the Code requirement. The Commission will 
still provide acceptance of the relief requests in writing. 
Therefore, it can be concluded that the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Create the possibility or a new or different kind of accident 
from any accident previously evaluated.
    No new failure modes have been defined for any plant system or 
component important to safety nor has any new limiting failure been 
identified as a result of the proposed changes. Therefore, it can be 
concluded that the proposed changes do not create the possibility of 
a new or different kind of accident from those previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes are administrative in nature and do not 
adversely impact the plant's ability to meet applicable regulatory 
requirements related to inservice testing or inspection. The 
proposed changes eliminate any possible misinterpretation of the 
Code requirements regarding relief requests and do not reduce the 
protection of public health and safety. Therefore, it can be 
concluded that the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: Charles L. Miller
    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: June 13, 1994
    Description of amendment request: The proposed amendments would 
permit an out-of-service component to be returned to service under 
administrative controls for the purpose of determining operability. The 
proposed change is consistent with the method utilized in the new 
Westinghouse Standard Technical Specifications (NUREG-1431). In 
addition, the proposed amendment corrects a typographical error in the 
header information on Page 3/4 0-2 of the current 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. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    a. Header Information
    This editorial change corrects a typographical error only. As 
such, existing accident analyses are unaffected.
    b. Specification 3.0.6
    The proposed change merely clarifies the intent of Specification 
3.0.2. As such, existing accident analyses are unaffected.
    2. Will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
     a.Header Information
    This editorial change corrects a typographical error only. As 
such, it does not alter the function of any plant equipment, involve 
any design changes, nor does it create any new operating modes or 
accident scenarios.
    b. Specification 3.0.6
    The proposed change merely clarifies the intent of Specification 
3.0.2. As such, it does not alter the function of any plant 
equipment, involve any design changes, nor does it create any new 
operating modes or accident scenarios.
    3. Will not involve a significant reduction in a margin of 
safety.
    a. Header Information
    This editorial change corrects a typographical error only. As 
such, the present margins of safety are unaffected.
    b. Specification 3.0.6
    The proposed change merely clarifies the intent of Specification 
3.0.2. As such, the present margins of safety are unaffected.
    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

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: June 17, 1994 as supplemented July 13, 
1994
    Description of amendment request: The proposed amendment would 
change the requirement to perform the Channel Functional Test of the 
Power Operated Relief Valve (PORV) position indication from quarterly 
to every 18 months and to exempt the PORV Block Valve position 
indication from performance of the channel Function Test if the PORV 
Block Valve is shut as required to isolate a PORV that cannot be 
manually cycled.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    A change from quarterly (Q) to at least every 18 months (R) may 
appear to be non-conservative, at first; however, by extending the 
surveillance requirement to cycle the PORV during non-power 
conditions, the change eliminates the potential risk and 
consequences of having the valve sticking open at power or not fully 
closing (leaking). Therefore, by extending the surveillance the 
probability and consequences of any previously analyzed accident is 
reduced, since the testing would now be conducted in a non-power 
condition, and the margin to safety is increased. Consequently, a 
net safety gain is realized by eliminating or minimizing these 
risks.
    The added note for PORV block valve is included for consistency 
and alignment between the surveillance requirement under this T/S 
(Table 4.3-11) with that of T/S surveillances 4.4.3.2 and 4.4.5.2 
for Units 1 and 2 respectively.
    Therefore, the proposed amendment does not involve a physical or 
procedural change to any structure, component, or system that 
significantly affects accident/malfunction probabilities or 
consequences previously evaluated in the UFSAR.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed changes do not introduce any design or physical 
configuration changes to the facility which could create new 
accident scenarios.
    3. Does not involve a significant reduction in a margin of 
safety.
    As stated in response to question number 1 above, the proposed 
changes do not eliminate the required T/S surveillance requirements. 
The first change eliminates the need to cycle the PORV valves 
through one complete cycle of full travel at power. The second 
change allows for not having to perform a surveillance on a valve 
that it is being used as an isolation point. The valve has been 
closed to comply with requirements of another T/S action statement. 
Therefore, the probability and consequences of any previously 
analyzed accident is reduced, thus increasing the margin to safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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

Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear 
Plant, Unit 3, Limestone County, Alabama

    Date of amendment request: March 29,1994 (TS 340)
    Description of amendment request: The proposed amendment to the 
Unit 3 Technical Specifications adds a limiting condition for operation 
and a surveillance requirement for a load shedding logic being added by 
a design change to Unit 3. The load shedding logic is being added to 
ensure that the maximum capacity of the Unit 3 Emergency Diesel 
Generators is not exceeded during a postulated loss of offsite power 
event concurrent with a design basis accident.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This proposed change establishes a surveillance testing 
requirement and limiting condition for operation for the Unit 3 480-
volt load shedding logic system. This Technical Specification change 
will not introduce any new failure mode and will not alter any 
assumptions previously made in evaluating the consequences of an 
accident. Accordingly, this change does not affect any design 
limiting safety system settings or operating parameters. 
Furthermore, the change does not modify or add any accident 
initiating events or parameters. Therefore, these proposed changes 
do not involve an 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.
    This proposed change establishes a limiting condition for 
operation and a surveillance requirement for the Unit 3 480-volt 
load shedding logic system. The addition of a limiting condition for 
operation and surveillance requirement will not adversely affect the 
operation of Unit 3 or the manner in which it is operated. 
Furthermore, the change does not create a failure mode that can lead 
to an accident of a different type than previously evaluated. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The addition of a limiting condition for operation and 
surveillance requirement will not reduce the margin of safety. The 
testing of the 480-volt load shedding logic on an 18-month interval 
is consistent with BWR/4 (NUREG-1433) Standard Technical 
Specifications. These are based on the guidance set forth in NRC 
Regulatory Guide 1.108, ``Periodic Testing of Diesel Generator Units 
Used as Onsite Electric Power Systems at Nuclear Power Plants.'' The 
addition of a limiting condition for operation establishes a minimum 
acceptable level of performance for the 480-volt load shedding logic 
system. Thus, the ability of the Emergency Diesel Generators to 
supply power during a loss of offsite power coincident with a design 
basis accident is assured.
    Furthermore, no reductions in the requirements or setpoints of 
the equipment supplied by the Emergency Diesel Generators are made 
which could result in a reduction in the margin of safety. 
Therefore, this proposed change does not involve a reduction in the 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: February 23, 1994
    Brief description of amendments: The proposed changes would revise 
the Comanche Peak Steam Electric Station (CPSES), Units 1 and 2, 
Technical Specifications (TS) to (1) allow a one-hour allowed outage 
time (AOT) following discovery of a closed cold leg injection 
accumulator discharge isolation in Modes 1, 2, or 3; (2) eliminate the 
redundant requirement to reverify accumulator boron concentration 
following fill from the refueling water storage tank (RWST); (3) 
relocate the accumulator water level and pressure channel analog 
channel operational test (ACOT) and channel calibration from the CPSES 
Technical Specifications to an administratively controlled program; (4) 
change the accumulator limits to analysis values rather than indicated 
values; and (5) reduce the inspection frequency following containment 
entries.
    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 a previously evaluated accident.
    The current requirement to immediately open a cold leg 
accumulator discharge isolation valve (or shut down the unit) upon 
discovery that the valve is closed is modified by the requested 
change to provide a one hour allowed outage time (AOT) prior to 
requiring a unit shutdown. This change is consistent with NUREG-
1431. The currently required action is more restrictive than that 
required by CPSES Technical Specification 3.0.3, that specifies the 
action required if an LCO [Limiting Condition for Operation] and its 
associated action requirements are not met and which provides a one 
hour AOT prior to taking steps to place the plant in Mode 3 within 
the following 6 hours. Following this requested change, the required 
actions for an accumulator declared inoperable due to a closed 
discharge isolation valve will be identical to those actions 
required for inoperability for other reasons, with the exception of 
the accumulator boron concentration being out of specification that 
has an AOT of 72 hours. Changing the AOT from ``immediate'' to one 
hour does not affect the probability of an accident. The only 
previously evaluated accident that is potentially impacted is the 
Loss of Coolant Accident (LOCA). With all valves open and thus all 
accumulators available, a potential LOCA is bounded by the existing 
accident analyses. With one accumulator discharge isolation valve 
closed and thus one accumulator not available, the consequences of a 
LOCA could be more severe; however, this requested amendment does 
not create this scenario. In other words, although the change in AOT 
may slightly increase the probability that, were a LOCA to occur, an 
accumulator would not be available (see the response to [number] 3 
below), it does not involve a significant increase in the 
consequences of an accident previously evaluated.
    The requirement to test the accumulator boron concentration 
following a 101 gallon or greater solution volume increase is 
modified by the requested change to exclude volume additions from 
the Refueling Water Storage Tank (RWST). Since the RWST boron 
concentration must be confirmed to satisfy the limits for the 
accumulators, there is no impact on the probability or consequences 
of any accident.
    The relocation of the accumulator water level and pressure 
channel ACOT and Channel Calibration from CPSES Technical 
Specifications to an administratively controlled program is 
essentially an administrative change. Because proper tests will 
still be performed, there is no impact on the probability or 
consequences of any accident.
    The requested change to reduce the containment debris 
inspections from ``at the completion of every entry'' to ``[a]t 
least once daily'' will require fewer inspections and is consistent 
with SR [Surveillance Requirement] 4.6.1.3 for the containment air 
locks. The accident of concern is a LOCA and these inspections have 
no impact on the probability of a LOCA. Performing fewer inspections 
would slightly increase the possibility that, should a LOCA occur, 
there could be debris in containment which could be transported to 
and partially clog the containment sump. However, inspecting at 
least daily if containment entries have been made is adequate and is 
justified by the reduced total radiation exposure for plant 
personnel. The inspections conducted at least daily assures that 
there is not a significant increase in the consequence of any 
accident.
    The requested changes do not modify the existing LCOs for 
Technical Specifications 3.5.1 and 3.5.2 with the exception of the 
replacement of ``indicated'' values with analysis values in LCO 
3.5.1, consistent with the relocation of the SRs for accumulator 
instrumentation. The requested changes are consistent with NUREG-
1431 and GL 93-05, and, as such, have already been generically 
assessed by the NRC. It is concluded that the requested changes do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated 
accident.
    The only requested change that modifies current operation of the 
plant is the requested one hour alowed outage time for action 
following discovery of a closed cold leg accumulator discharge 
isolation valve. The requested one hour completion time to open the 
valve continues to ensure that prompt action will be taken to return 
the inoperable accumulator to an operable status, minimizing the 
potential for exposure of the plant to a LOCA under this condition. 
In addition, as LCO 3.5.1a will continue to require that the 
accumulator discharge isolation valve be open with power removed 
from the valve operator, the probability of the discharge isolation 
valve being closed in Modes 1, 2, or 3 will remain low. This change 
in current operation does not create the possibility of a new or 
different kind of accident.
    The requested slight reduction in the containment inspection 
frequencies specified in SR 4.5.2 only serves to reduce the number 
of unnecessary inspections. It does not make substantial changes to 
the inspection requirements, nor does it change the method of 
performing these requirements. Thus, the requested change does not 
create the possibility of a new or different kind of accident.
    No significant changes to the limiting conditions for operation 
of the accumulators or the emergency core cooling system are 
requested as part of this amendment request. The requested changes 
do not involve any physical changes to the plant. The requested 
changes are consistent with NUREG-1431 and GL 93-05, and, as such, 
have already been generically assessed by the NRC. Thus, the 
requested changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The only requested change that modifies current operation of the 
plant is the requested one hour allowed outage time for action 
following discovery of a closed cold leg accumulator discharge 
isolation valve. As noted in the response to [number] 1 above, this 
requested change in AOT does not significantly affect the 
probability or consequences of an accident, but does increase the 
possibility that, should a LOCA occur, one of the accumulators may 
not be available to help mitigate the consequences of the accident. 
However, the requested one hour completion time to open the valve 
continues to ensure that prompt action will be taken to return the 
inoperable accumulator to an operable status, minimizing the 
potential for exposure of the plant to a LOCA under this condition. 
In addition, as LCO 3.5.1a will continue to require that the 
accumulator discharge isolation valve be open with power removed 
from the valve operator, the probability of the discharge isolation 
valve being closed in Modes 1, 2, or 3 will remain low. Considering 
the controls above and the fact that the requested action statement 
is consistent with TS 3.0.3, it is concluded that the requested 
change does not involve a significant reduction in the margin of 
safety.
    The requested slight reduction in the containment inspection 
frequencies specified in SR 4.5.2 only serves to reduce the number 
of unnecessary inspections conducted and reduce the personnel 
exposure associated with the inspections. As adequate inspections 
will continue to be conducted, this requested change does not 
involve a significant reduction in a margin of safety.
    No significant changes to the limiting conditions for operation 
of the accumulators or the emergency core cooling systems are 
requested as part of this amendment request. The requested changes 
are consistent with NUREG-1366, NUREG-1431 and GL 93-05, and, as 
such, have already been generically assessed by the NRC. Thus, it is 
concluded that the requested 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: University of Texas at 
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O. 
Box 19497, Arlington, Texas 76019
    Attorney for licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: April 22, 1994
    Brief description of amendments: The proposed amendments would 
revise the technical specifications by changing the frequency of 
auxiliary feedwater pump operational testing from monthly to quarterly.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences or a previously evaluated 
accident.
    Because the Auxiliary Feedwater System pumps are provided to 
mitigate certain accidents, altering the test frequency of the pumps 
will not impact the probability of an accident. The Auxiliary 
Feedwater System pumps will continue to be tested quarterly on a 
staggered basis to the same standards applied to safety-related 
pumps as defined by ASME Section XI. Satisfactory completion of the 
testing in accordance with the Code is used as verification that 
safety-related pumps will be available to perform their intended 
function. Quarterly testing of the Auxiliary Feedwater System pumps 
on a staggered basis, therefore, will continue to assure that the 
Auxiliary Feedwater System will be capable of performing its 
intended function. It is thus concluded that the requested change 
will not involve a significant increase in the probability of 
occurrence or consequences of an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Changing the surveillance test frequency of the Auxiliary 
Feedwater Pumps does not involve any physical modification of the 
plant or result in a change in a method of operation. Therefore, it 
is concluded that the requested change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    Changing the surveillance testing frequency of the Auxiliary 
Feedwater System pumps does not affect any safety limits or any 
limiting safety system settings. System operating parameters are 
unaffected. The availability of equipment required to mitigate or 
assess the consequence of an accident is not reduced; in fact the 
availability is increased because the system is rendered inoperable 
on a quarterly basis to perform pump testing, rather than a monthly 
basis. Further, vibration testing being the most effective early 
indication of gradual pump degradation continues to be performed on 
the same frequency. Quarterly testing of the Auxiliary Feedwater 
pumps on a staggered basis in accordance with the criteria specified 
in the ASME Section XI code provides adequate assurance that the 
Auxiliary Feedwater System pumps are capable of performing their 
intended function. Thus, its [sic] is concluded that the requested 
change does not involved [sic] a significant reduction in a margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O. 
Box 19497, Arlington, Texas 76019
    Attorney for licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: April 25, 1994
    Brief description of amendments: The proposed amendments would 
revise the technical specifications to reduce the number of fast starts 
currently required by surveillance requirements for the emergency 
diesel generators (EDGs).
    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 a previously evaluated accident.
    This change reduces the number of ``fast starts'' required on 
the EDGs and allows the EDGs to be tested using ``slow starts.'' 
Reducing the number of ``fast starts'' (required to start in 10 
seconds or less) will reduce the wear on the EDGs primarily by 
minimizing coking of fuel in the cylinder and preventing premature 
wearing of the turbocharger thrust bearing. This increases engine 
reliability and availability. A ``slow start'' may require that the 
EDG be taken out of service to perform the test if the EDG start 
time is not 10 seconds or less. TU Electric feels that testing the 
``fast start'' capability of the EDG every 184 days will maintain 
its present level of reliability. The period of time in which the 
EDG is actually inoperable due to testing (i.e., may not start and 
be ready to load in 10 seconds) is quite short. Overall the 
reliability and the availability of the EDG will be increased.
    The impact of the EDGs on the postulated accidents is directly 
related to their reliability and availability. Therefore, the 
proposed reduction in the number of ``fast starts'' does not involve 
a significant increase in the probability or consequences of any 
previously evaluated accident.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated 
accident.
    The revised testing allowed by this Technical Specification 
change does not create a new or different kind of accident. The EDGs 
are primarily accident mitigation components. The potential failure 
of EDGs have already been assessed in the CPSES design.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The only aspect of this change that could adversely affect the 
margin of safety is the potential impact on the start time of the 
engine. The start time of the engine is not expected to exceed the 
assumption in the accident analyses except possibly in the short 
period of time required to perform the ``slow start'' test. If the 
EDG does not start in 10 seconds or less under these conditions, the 
EDG is declared inoperable, as allowed by the Technical 
Specifications, to perform the ``slow start'' test. Because these 
periods of inoperability are only implemented as allowed by the 
Technical Specifications, there is no impact on the margin of 
safety. The margin of safety established by the assumed EDG 
availability will be enhanced by the increased reliability and 
availability of the EDGs.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O. 
Box 19497, Arlington, Texas 76019
    Attorney for licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: April 25, 1994
    Brief description of amendments: The proposed amendments would 
revise the technical specifications by updating the unit staff 
qualification requirements to Regulatory Guide 1.8, Revision 2, 
``Qualification and Training of Personnel for Nuclear Power Plants,'' 
and by relocating administrative control of training from the technical 
specifications to the Final Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of a previously evaluated accident.
    This proposed amendment involves a consolidation of previous 
unit staff qualification requirements into one source document, and 
a relocation of training requirements to a more appropriate license 
basis document. The qualification requirements of the unit staff 
remains the same as the existing requirements, and the relocation of 
the training requirements does not change the scope of the program 
as it now exists. The relocated training program requirements retain 
adequate administrative and regulatory controls to ensure the plant 
is not placed in an unanalyzed condition.
    These changes are administrative in nature. They remain within 
the assumptions of the current accident analysis. As a result, they 
do not increase the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated 
accident.
    These changes are administrative in nature. They merely 
consolidate qualification requirements and relocate training 
requirements. The relocated program requirements retain adequate 
administrative and regulatory controls to ensure the plant is not 
introduced to an unreviewed safety question.
    These changes are administrative in nature. They do not 
introduce any new initiating events. As a result, they do not create 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed changes do not involve a signficant reduction in 
a margin of safety.
    These changes are administrative in nature and have no impact on 
actual plant protection or safety actuation systems, or the assumed 
actions performed in accordance with normal, abnormal, or emergency 
operating procedures. There are adequate regulatory and plant 
configuration controls existing to ensure there is no impact on the 
plant 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 Texas at 
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O. 
Box 19497, Arlington, Texas 76019
    Attorney for licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: April 21, 1994
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 
by adding requirements for the steam exclusion system. The new TS 
addresses the steam exclusion system (TS 3.15) and the surveillance on 
the steam exclusion system (TS 4.15). This system was not previously 
addressed by 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 which is presented below:
    A) TS 3.15 Steam Exclusion System (New)
    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The intent of this new specification is to specify the 
operability requirements for the Steam Exclusion System and to 
demonstrate the acceptability of removing the Steam Exclusion System 
from service for short periods of time.
    The proposed change will not significantly increase the 
probability of an accident previously evaluated. The accident under 
consideration is a high energy line break outside of containment. 
Allowing a steam exclusion boundary to be inoperable for a short 
period of time has no effect on the probability of occurrence of a 
high energy line break outside of containment.
    The proposed change will not significantly increase the 
consequences of an accident previously evaluated. Again, the 
accident under consideration is a high energy line break outside of 
containment. Calculations conclude that the core damage frequency 
for a high energy line break outside of containment with a non-
redundant steam exclusion boundary open is 2.57E-8 per 12-hour 
period. Further conservative assumptions of one non-redundant steam 
exclusion boundary being open 12 hours per day, 5 days per week, 52 
weeks per year results in a core damage frequency of 6.68E-6 per 
year. This analysis was conservatively calculated taking minimal 
credit for mitigating the accident, and is considered to be an 
acceptable level of risk on an annual basis. A safety factor of five 
was applied to NUREG/CR-4550 data to determine the initiating event 
frequency of a high energy line break. This calculation supports the 
conclusion that this addition to the Technical Specifications will 
not result in a significant increase in the probability or 
consequences of an high energy line break outside of containment.
    Furthermore, calculations conclude that the core damage 
frequency for a high energy line break outside of containment with 
one of two redundant steam exclusion boundaries open is 4.62E-10 per 
72-hour period. Further conservative assumptions of one redundant 
steam exclusion damper being open 24 hours per day, 5 days per week, 
52 weeks per year results in a core damage frequency of 4.00E-8 per 
year. This analysis was conservatively calculated taking minimal 
credit for mitigating the accident, and is also considered to be an 
acceptable level of risk on an annual basis. Again, a safety factor 
of five was applied to NUREG/CR-4550 data to determine the 
initiating event frequency of a high energy line break. This 
calculation also supports the conclusion that this addition to the 
Technical Specifications will not result in a significant increase 
in the probability or consequences of an high energy line break 
outside of containment.
    Specific requirements for the Steam Exclusion System do not 
currently exist in the Technical Specifications. Addition of TS 3.15 
is an enhancement to the Kewaunee Technical Specifications, and 
providing this information for the plant staff and operators will 
not significantly increase the probability or consequences of an 
accident previously evaluated, nor will it adversely affect the 
health and safety of the public.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed amendment does not alter the plant configuration, 
operating setpoints or overall plant performance. Therefore, it 
cannot create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3) Involve a significant reduction in the margin of safety.
    Addition of the specification is an enhancement to the Technical 
Specifications and does not alter input to the safety analysis. 
Furthermore, the supporting analysis demonstrates an acceptable 
level of risk for removing components from service for limited 
periods of time. Therefore, it will not involve a significant 
reduction in the margin of safety.
    Additionally, the proposed change is similar to example 
C.2.e(ii) in 51 FR 7751. Example C.2.e(ii) states that changes that 
constitute an additional limitation, restriction or control not 
presently included in the TS's are not likely to involve a 
significant hazard.
    B) S 4.15 Steam Exclusion System (New)
    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change will not significantly increase the 
probability or consequences of an accident previously evaluated. The 
accident under consideration is a high energy line break outside of 
containment. The performance of periodic surveillance requirements, 
testing which verifies that components in the Steam Exclusion System 
are operating properly, cannot significantly increase the 
probability or consequences of a high energy line break, nor will it 
adversely affect the health and safety of the public.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed amendment does not alter the plant configuration, 
operating setpoints or overall plant performance. Therefore, it 
cannot create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3) Involve a significant reduction in the margin of safety.
    Addition of the specification is an enhancement to the Kewaunee 
Technical Specifications and does not alter input to the safety 
analysis. Therefore, it will not involve a significant reduction in 
the margin of safety.
    Additionally, the proposed change is similar to example 
C.2.e(ii) in 51 FR 7751. Example C.2.e(ii) states that changes that 
constitute an additional limitation, restriction or control not 
presently included in the TS's are not likely to involve a 
significant hazard.
    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

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

    Date of amendment request: May 26, 1994
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 
Sections 2.3, 3.6, and 4.6 by correcting minor typographical errors and 
format inconsistencies. These changes are being proposed as a part of 
the licensee's ongoing effort to revise each section of the KNPP TS to 
achieve a consistent format and to convert the entire document to Word 
Perfect. In addition, changes to the basis for TS Sections 2.3, 3.6, 
and 4.6 have also been proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    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.
    2) create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3) involve a significant reduction in the margin of safety.These 
proposed changes involve the conversion of the TS to the Word 
Perfect format now being used at WPSC. Minor typographical errors 
and format inconsistencies were corrected. These proposed changes 
are administrative in nature; accordingly, these proposed changes do 
not involve a significant hazards consideration. Additionally, the 
proposed changes are similar to example C.2.e.(i) in 51 FR 7751. 
Example C.2.e.(i) states that changes which are purely 
administrative in nature; i.e., to achieve consistency throughout 
the Technical Specifications, correct an error, or a change in 
nomenclature, are not likely to involve a significant hazard.
    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.

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 application for amendment: January 4, 1994
    Brief description of amendment request: The proposed amendment 
would implement recommended changes from Generic Letter (GL) 93-05, 
``Line-Item Technical Specification Improvements to Reduce Surveillance 
Requirements for Testing During Power Operation.'' Specifically, the 
licensee proposed to change their Technical Specifications 
corresponding to the following GL 93-05 line numbers: 4.1.2, 5.8, 5.14, 
6.1, 7.5, 8.1, 9.1, 12, and 14. Date of individual notice in Federal 
Register: July 22, 1994 (59 FR 37513)
    Expiration of individual notice: August 22, 1994
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004

Arizona Public Service Company, et al., Docket No. STN 50-529, Palo 
Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona

    Date of application for amendment: July 1, 1994
    Brief description of amendment request: The proposed amendment 
would modify Technical Specification (TS) Figure 3.2-1, ``Reactor 
Coolant Cold Leg Temperature vs. Core Power Level.'' Specifically, the 
minimum cold leg temperature for core power levels between 90 percent 
and 100 percent would be changed to 552 deg.F (which is a reduction of 
10 deg.F from the previous TS requirement). This TS change permits 
reactor operation at full power with a lower reactor coolant 
temperature to minimize potential steam generator tube degradation.
    Date of individual notice in Federal Register: July 13, 1994 (59 FR 
35767)
    Expiration of individual notice: August 12, 1994
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004

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

    Date of amendment request: June 20, 1994
    Description of amendment request: The proposed amendment would 
revise the Braidwood, Unit 1, Technical Specifications (TSs) to remove 
the condition limiting operation of the facility to 100 days during the 
present fuel cycle when Thot is greater than 500 deg.F and to 
restore the reactor coolant dose equivalent Iodine-131 limit to 1 
microcurie per gram of coolant from the present value of 0.35. Both the 
limit on permissible operational time and the reduction in the 
permissible level of Iodine-131 were incorporated into the TSs by 
Amendment No. 50 issued to
    Facility Operating License No. NPF-72 for Braidwood Station, Unit 
1, on May 7, 1994.
    Date of publication of individual notice in Federal Register:  July 
11, 1994 (59 FR 35389)
    Expiration of individual notice: August 10, 1994
    Local Public Document Room location: Wilmington Township Public 
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.

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

    Date of amendment request: June 16, 1994
    Description of amendment request: The proposed amendments would 
consist primarily of an administrative change to the Zion Station's 
Technical Specifications (TSs) to reflect an exemption to 10 CFR Part 
50, Appendix J, Section III.D.3.
    Date of publication of individual notice in Federal Register: June 
30, 1994 (59 FR 33798)
    Expiration of individual notice: August 1, 1994
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085

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

    Date of amendment request: June 24, 1994
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to change the Administrative 
Controls section to require an individual who serves as the Operations 
Manager to either hold a Millstone Unit 2 senior reactor operator (SRO) 
license or have an SRO license at another pressurized water reactor. If 
the Operations Manager does not hold a Millstone Unit 2 SRO license, 
then an individual serving as the Assistant Operations Manager would be 
required to possess an SRO license at Millstone Unit 2. Date of 
publication individual notice in Federal Register: July 7, 1994 (59 FR 
34872).
    Expiration of individual notice: August 8, 1994
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.

Southern Nuclear Operating Company, Inc., Docket No. 50-348,Joseph 
M. Farley Nuclear Plant, Unit 1

    Date of amendment request: June 17, 1994
    Brief description of amendment request: The amendment changes the 
Technical Specifications to revise the nuclear enthalpy rise hot 
channel factor (F delta H) from equal to or less than 1.65 [1 plus 
0.3(1-P)] to equal to or less than l.70 [1 plus 0.3(1-P)] where P is a 
fraction of rated power. The amendment also revises the action 
statement to reflect guidance contained in the improved standard 
technical specifications.
    Date of publication of individual notice in Federal Register: June 
22, 1994 (59 FR 32249)
    Expiration of individual notice: July 22, 1994
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302

Notice Of Issuance Of Amendments To Facility Opersting 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.

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

    Date of application for amendment: August 5, 1993, as supplemented 
by letter dated January 19, 1994.
    Brief description of amendment: The amendment extends the 
surveillance interval of selected channel functional test for the 
reactor protection system (RPS) and the engineered safety feature 
actuation system (ESFAS) instrumentation from once per month to 
quarterly. In addition, the amendment will modify Technical 
Specification (TS) 2.2.1, Table 2.2-1 to change selected reactor trip 
setpoints and allowable values. This amendment supersedes an amendment 
request dated December 28, 1992, published in the Federal Register on 
February 3, 1993 (58 FR 6994)
    Date of issuance: July 15, 1994
    Effective date: July 15, 1994
    Amendment Nos.: 78, 64, and 50
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: 
Amendment revised the Technical Specifications.
    Date of initial notice in Federal Register: September 15, 1993 (58 
FR 48378) The additonal information contained in the January 19, 1994, 
letter was clarifying in nature, was within the scope of the initial 
notice, and did not affect the NRC staff's proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
July 15, 1994.No significant hazards consideration comments received: 
No.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004

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

    Date of application for amendment: March 25, 1994
    Brief description of amendment: The amendment revises TS 3/4.8.4.2, 
by eliminating the term ``motor starter'' and replacing it with a more 
accurate description of the MOV bypass configuration.
    Date of issuance: July 12, 1994
    Effective date: July 12, 1994
    Amendment No.: 48
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1994 (59 FR 
22002) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 12, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson 
SteamElectric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: May 20, 1994
    Brief description of amendment: The amendment revises the Technical 
Specification requirement for the Manager - Operations to hold a Senior 
Reactor Operators (SRO) license at HBR.The revision allows the Manager 
- Operations position to be filled by an individual who holds or has 
held an SRO license at either HBR or a similar plant. The amendment 
also requires the Manager - Shift Operations to hold an SRO license for 
HBR.
    Date of issuance: July 15, 1994
    Effective date: July 15, 1994
    Amendment No.: 148
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29625) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 15, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Hartsville Memorial Library, 
Home and Fifth Avenues, Hartsville, South Carolina 29550

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

    Date of application for amendment:  March 12, 1993
    Brief description of amendment: The amendment revises Technical 
Specification Table 3.3.7.1-1 Action 72, to clarify the actions to be 
taken if the control room ventilation radiation monitor becomes 
inoperable.
    Date of issuance: July 18, 1994
    Effective date: as of its date of issuance to be implemented within 
90 days.
    Amendment No. 64
    Facility Operating License No. NPF-58. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1994 (59 FR 
22013) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 18, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

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: March 11, 1994
    Brief description of amendments: The amendments add new hydraulic 
snubbers on the main steamlines to the Technical Specifications (TS) 
for Quad Cities, Units 1 and 2, and also change the snubber visual 
inspection interval and corrective actions in TS Section 3.6.1 and 
4.6.1 to the format and content of the BWRs STSs, as revised by Generic 
Letter 90-09, ``Alternative Requirements for Snubber Visual Inspection 
Intervals and Corrective Actions,'' dated December 11, 1990.
    Date of issuance: July 13, 1994
    Effective date: July 13, 1994
    Amendment Nos.: 149 and 145
    Facility Operating License Nos. DPR-29 and DPR-30. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: 59 FR 17595 (April 13, 
1994) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 13, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South CarolinaDate of 
application for amendments: January 25, 1993, as supplemented May 
12, 1993

    Brief description of amendments: The amendments revise the 
Technical Specifications to allow longer surveillance test intervals 
and allowed outage times for the reactor protection system and the 
engineered safety features actuation system.
    Date of issuance: July 18, 1994
    Effective date: July 18, 1994
    Amendment Nos.: 122 and 116
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 4, 1993 (58 FR 
41501) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 18, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

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

    Date of application for amendments: February 25, 1994
    Brief description of amendments: The amendments revise Technical 
Specification (TS) Tables 3.3-10 and 4.3-7 to add four instruments as 
part of the accident monitoring instrumentation, and delete five 
instruments from the TS Tables that are not part of the accident 
monitoring instrumentation.
    Date of issuance: July 18, 1994
    Effective date: July 18, 1994
    Amendment Nos.: 144 and 126
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17597) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 18, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223

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, 
FloridaDate of application for amendments: February 22, 1994

    Brief description of amendments: These amendments relocate the 
instrument response time limits for the Reactor Protective System and 
the Engineered Safety Features Actuation System from the Technical 
Specifications to be Updated Safety Analysis Report for both units.

    Date of issuance: July 12, 1994
    Effective date: July 12, 1994
    Amendment Nos.: 128 and 67
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17598) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 12, 1994No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

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

    Date of application for amendment: April 11, 1994
    Brief description of amendment: The amendment revises the plant 
Technical Specifications (TS) to relocate the detailed inspection 
criteria, methods and frequencies of the containment tendon 
surveillance program to the Final Safety Analysis Report (FSAR) and to 
provide a direct reference to the tendon surveillance program in the 
TS, and make certain editorial changes.

    Date of issuance: July 14, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 187
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29627). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 14, 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.

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

    Date of amendment request: March 21, 1994
    Brief description of amendments: The amendments revise Technical 
Specifications 3.1.2.3 ``Reactivity Control Systems Charging Pumps - 
Shutdown'' and 3.1.2.1 ``Boration Systems Flow Paths - Shutdown.'' The 
amendments allow energizing of an inoperable centrifugal charging pump 
in preparation for switching of the centrifugal charging pumps, 
provided the pump discharge is isolated from the reactor coolant 
system. The amendment allows for continued flow to the reactor coolant 
pump seals.
    Date of issuance: July 12, 1994
    Effective date: July 12, 1994, to be implemented within 31 days of 
issuance.
    Amendment Nos.:  Unit 1 - Amendment No. 62; Unit 2 - Amendment No. 
51
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17602) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 12, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488

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: April 18, 1994
    Brief description of amendment: Technical Specification Sections 3/
4.6.6.3, ``Standby Gas Treatment System,'' and 3/4.7.2, ``Control Room 
Ventilation System,'' require periodic testing of the charcoal filter 
beds to demonstrate their continuing effectiveness in removing 
radioiodine. The specifications, which referenced the testing 
methodology of ASTM D3803-79, have been updated to reference the 1989 
version of the standard.
    Date of issuance: July 22, 1994
    Effective date: July 22, 1994
    Amendment No.: 91
    Facility Operating License No. NPF-62. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1994 (59 FR 
24750) The June 16, 1994, submittal consisted of revisions/
clarifications which did not change the staff's initial proposed no 
significant hazards considerations determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 22, 1994.No significant hazards 
consideration comments received: No
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.

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: November 17, 1993.
    Brief description of amendments: The amendments modify the 
Technical Specifications to allow a portion of the Waste Gas Holdup 
System Explosive Monitoring System to be inoperable for 160 days on a 
one-time basis so that the Waste Gas Oxygen Analyzer can be replaced. 
These amendments also make an editorial change to the Automatic Gas 
Analyzer tag numbers.
    Date of issuance: July 7, 1994
    Effective date: July 7, 1994
    Amendment Nos.: 179 & 163
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1994 (59 FR 
4938) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 7, 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.

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

    Date of application for amendments: August 6, 1993
    Brief description of amendments: This amendment request changes the 
technical specifications to provide for a separate action if the 
Accumulator cannot meet the requirements of the Limiting Condition for 
Operation due to boron concentration. The allowed outage time to 
restore boron concentration is changed from 1 hour to 72 hours.
    Date of issuance: July 20, 1994
    Effective date: July 20, 1994
    Amendment Nos. 152 and 132
    Facility Operating License Nos. DPR-70 and DPR-75. These amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 15, 1993 (58 
FR 48389) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 20, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

Public Service Electric & Gas Company, Docket No. 50-311, Salem 
Nuclear Generating Station, Unit No. 2, Salem County, New Jersey

    Date of application for amendment: June 11, 1993 and supplemented 
July 19, August 3, and September 16, 1993.
    Brief description of amendment: The amendment reduces the boron 
concentration in the boric acid tank from 12 percent by weight to 
between 3.75 and 4 percent by weight. The reduced boron concentration 
results in eliminating the need for heat tracing in the boric acid tank 
piping system.
    Date of issuance: July 20, 1994
    Effective date: July 20, 1994
    Amendment No. 133
    Facility Operating License No. DPR-75: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 18, 1993 (58 FR 
43932) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 20, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, 
AlabamaDate of application for amendments: December 23, 1993 (TS 
346)

    Brief description of amendment: The amendments revise the Technical 
Specification surveillance requirements regarding the visual inspection 
of snubbers, consistent with the guidance in Generic Letter 90-09, 
``Alternative Requirements for Snubber Visual Inspection Intervals and 
Corrective Actions.''
    Date of issuance: July 5, 1994
    Effective date: July 5, 1994
    Amendment Nos.:  210, 225 and 183
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: This 
amendment revised the Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27067) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 5, 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: May 16, 1994 (TS 93-18)
    Brief description of amendments: The amendments change the 
Electrical Power Systems surveillance requirements wording to reflect 
the use of the new common station service transformers with auto load 
tap changers as the normal power supply for the 6.9 KV unit boards.
    Date of issuance: July 11, 1994
    Effective date: July 11, 1994
    Amendment Nos.: 184 and 176
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29637) The Commission's related evaluation of the amendments are 
contained in a Safety Evaluation dated July 11, 1994. No significant 
hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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

    Date of application for amendment: March 18, 1994, as supplemented 
on June 20, 1994.
    Brief description of amendment: The amendment revises TS 2.1.2 
(Reactor Core), TS 2.2.1 (Reactor Protection System Setpoints), Bases 
2.1.1 and 2.1.2 (Reactor Core), Bases 2.2.1 (Reactor Protection System 
Instrumentation Setpoints), TS 3.2.2 and 3.2.3 (Power Distribution 
Limits), Bases 3/4 (Power Distribution Limits), and TS 6.9.1.7 
(Administrative Controls, Core Operating Limits Report). This amendment 
removes cycle-specific limits from TS and relocates them in the Core 
Operating Limits Report.
    Date of issuance: July 22, 1994
    Effective date: July 22, 1994
    Amendment No. 189
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1994 (59 FR 
22014) The June 20, 1994, submittal provided supplemental 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 July 22, 1994.No 
significant hazards consideration comments received: No
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

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

    Date of application for amendment: February 21, 1992, supplemented 
by letters dated April 16, 1992, and March 29, 1994
    Brief description of amendment: The amendment revises the Technical 
Specifications Appendix B, Environmental Protection Plan (Non-
radiological), by removing Sections 2.3 and 4.3, ``Cultural 
Resources.'' Union Electric has developed and maintains a management 
plan for the protection of cultural resources on the Callaway Plant 
site. The amendment request summarizes the plan that provides the 
status and disposition of each portion of the current Appendix B 
sections related to cultural resources.
    Date of issuance: July 13, 1994
    Effective date: to be implemented within 30 days of issuance
    Amendment No.: 90
    Facility Operating License No. NPF-30. Amendment revised the 
Technical Specifications Appendix B, Environmental Protection Plan.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17607) The additional information contained in the April 16, 1992, and 
March 29, 1994, letters was supplemental in nature, is within the scope 
of the initial notice, and did not affect the NRC staff's proposed no 
significant hazards consideration. The Commission's related evaluation 
of the amendment is contained in a Safety Evaluation dated July 13, 
1994.No significant hazards consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

Virginia Ele1ctric and Power Company, et al., Docket Nos. 50-338 
and 50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa 
County, Virginia

    Date of application for amendments: March 1, 1994, as supplemented 
by letter dated June 16, 1994
    Brief description of amendments: The amendments modify the 
requirement for operability testing of an EDG when the alternate safety 
buses' EDG is inoperable. Also, the requirement for operability testing 
of the EDGs when one or both offsite AC sources are inoperable is 
deleted. Finally, the amendments eliminate fast loading of EDGs except 
for the Loss of Offsite Power test and separate the hot restart test 
from the 24-hour loaded test run of the EDGs. The changes are 
consistent with NRC Generic Letter 93-05 dated September 27, 1993.
    Date of issuance: July 18, 1994
    Effective date: July 18, 1994
    Amendment Nos.: 184 and 165
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14899) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 18, 1994No significant 
hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

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

    Date of application for amendment: December 6, 1993, supplemented 
by letter dated May 6, 1994
    Brief description of amendment: The amendment revised Table 
4.3.7.5-1 of Technical Specification (TS) 3/4.3.7.5, ``Accident 
Monitoring Instrumentation,'' to include a note that requires that 
safety/relief valve (SRV) position indicator surveillance testing be 
performed within 12 hours after steam pressure and flow are adequate to 
do the testing. The amendment also revised TS 3/4.4.2, ``Safety/Relief 
Valves,'' and TS 3/4.5.1, ``Emergency Core Cooling Systems,'' to 
require that the main steam system and automatic depressurization 
system SRVs be surveilled within 12 hours after steam pressure and flow 
are adequate to do the testing. Additionally, Bases Section 3/4.4.2 was 
revised to reflect the changes in the TS.
    Date of issuance: July 8, 1994
    Effective date: July 8, 1994
    Amendment No.: 128
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 13, 1994 (59 FR 
25131) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 8, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

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

    Date of application for amendment: December 7, 1993
    Brief description of amendment: The amendment revises Kewaunee 
Nuclear Power Plant (KNPP) Technical Specification (TS) 5.3.a.1 to 
provide flexibility in the repair of fuel assemblies containing damaged 
and leaking fuel rods by reconstituting the assemblies, provided that 
an NRC-approved methodology is used. This change is consistent with 
guidance provided in Supplement 1 to Generic Letter (GL) 90-02, 
``Alternative Requirements for Fuel Assemblies in the Design Features 
Section of Technical Specifications,'' dated July 31, 1992. In 
addition, administrative changes to KNPP TS Section 5 have been made.
    Date of issuance: July 15, 1994
    Effective date: to be implemented within 30 days
    Amendment No.: 109
    Facility Operating License No. DPR-43. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1994 (59 FR 
4951) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 15, 1994.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.

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 September 2, 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).

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

    Date of application for amendment: July 14, 1994
    Brief description of amendment: The amendment revises Technical 
Specification 3.6.1.4, main steam isolation valve (MSIV) leakage 
control system (LCS), to add a footnote to the APPLICABILITY statement. 
The footnote states, ``The provisions of Specification 3.0.4 are not 
applicable from the effective date of this amendment until the 
completion of Operating Cycle 5.''
    Date of issuance: July 15, 1994
    Effective date: July 15, 1994
    Amendment No. 63
    Facility Operating License No. NPF-58. This amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: No.The Commission's related 
evaluation of the amendment, finding of emergencycircumstances, and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated July 15, 1994.
    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, DC 20037.
    NRC Project Director: John N. Hannon

Patriots Point Development Authority and U.S. Maritime 
Administration, Docket No. 50-238, N.S. Savannah

    Date of application for amendment: May 19, 1994, as supplemented on 
May 24 and 27, 1994, and June 3, 1994.
    Brief description of amendment: The amendment (1) deleted the 
Patriots Point Development Authority (PPDA) as a co-licensee, (2) 
allowed relocation of the N.S. Savannah to the James River Reserve 
Fleet, a U.S. Maritime Administration (MARAD) facility, (3) changed the 
performance of radiological health physics coverage, surveillance and 
response to the U.S. Army Center for Public Works, Humphries 
Engineering Center, (4) changed the composition of the Review and Audit 
Committee to be consistent with the deletion of PPDA as a co-licensee, 
and (5) discontinued public access to the facility and made other minor 
changes to the TS.
    Date of issuance: June 29, 1994
    Effective date: June 29, 1994
    Amendment No.: 12Amended Facility License No. NS-1: Amendment 
revised the Technical Specifications and license.Public comments 
requested as to proposed no significant hazards consideration: 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 no significant hazards consideration be 
provided to the staff by the close of business on June 2, 1994. The 
notice was published in The Virginian-Pilot/The Ledger-Star, Norfolk, 
Virginia on Sunday, May 29, 1994, The Daily Press, Newport News, 
Virginia on Friday, May 27, 1994, and The Post and Courier, Charleston, 
South Carolina, on Friday, May 27, 1994. No comments have been 
received.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, consultation with the States of South Carolina 
and Virginia and final no significant hazards consideration 
determination are contained in a Safety Evaluation dated June 29, 1994.
    Attorney for licensees: M. J. McMorrow, Office of the Chief Council 
(MAR 220), U.S. Maritime Administration, Room 7228, 400 Seventh Street, 
SW, Washington, D.C. 20590.
    Local Public Document Room location: N/A

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

    Date of application for amendment: July 8, 1994
    Brief description of amendment: The amendment modified the 
technical specifications to permit post-maintenance testing of control 
rod scram insertion times to be performed at lower reactor coolant 
pressures than currently allowed.
    Date of issuance: July 14, 1994
    Effective date: July 14, 1994
    Amendment No.: 129
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.Public comments on proposed no significant 
hazards consideration comments received: 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 July 14, 1994.
    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, NW., Washington, DC 20005-3502
    NRC Project Director: Theodore R. Quay
    Dated at Rockville, Maryland, this 27th day of July 1994.
    For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II,Office of Nuclear Reactor 
Regulation.
[FR Doc. 94-18741 Filed 8-2-94; 8:45 am]
BILLING CODE 7590-01-F