[Federal Register Volume 60, Number 168 (Wednesday, August 30, 1995)]
[Notices]
[Pages 45172-45196]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-10830]



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NUCLEAR REGULATORY COMMISSION

Biweekly Notice

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

I. Background

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

    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from August 4, 1995, through August 18, 1995. The 
last biweekly notice was published on August 16, 1995 (60 FR 42597).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By September 29, 1995, the licensee may file a request for a 
hearing with respect to issuance of the amendment to the subject 
facility operating license and any person whose interest may be 
affected by this proceeding and who wishes to participate as a party in 
the proceeding must file a written request for a hearing and a petition 
for leave to intervene. Requests for a hearing and a petition for leave 
to intervene shall be filed in accordance with the Commission's ``Rules 
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. 
Interested persons should consult a current copy of 10 CFR 2.714 which 
is available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., 

[[Page 45173]]
Washington, DC and at the local public document room for the particular 
facility involved. If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or an Atomic 
Safety and Licensing Board, designated by the Commission or by the 
Chairman of the Atomic Safety and Licensing Board Panel, will rule on 
the request and/or petition; and the Secretary or the designated Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendments request: August 3, 1995
    Description of amendments request: The proposed amendment changes 
would add the analytical method supplement entitled ``Fuel Rod Maximum 
Allowable Gas Pressure,'' CEN-372-P-A, dated May 1990, and its 
associated Nuclear Regulatory Commission Safety Evaluation Report, 
dated April 10, 1990, to the list of analytical methods in TS 6.9.1.10 
used to determine the PVNGS core operating limits.
    Basis for proposed no significant hazards consideration 
determination: As requied by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not involve any change to the 
configuration or method of operation of any plant equipment that is 
used to mitigate the consequences of an accident. The proposed 
change adds an NRC approved methodology and its associated Safety 
Evaluation Report (SER), to the list of analytical methods used to 
determine the core operating limits. The use of this methodology 
ensures that the consequences of an accident remain within the 
limits established by existing analyses. They do not alter any of 
the assumptions or bounding conditions currently in the UFSAR.
    The U3C6 ECCS performance analysis included the analysis of the 
impact of the maximum calculated fuel rod gas pressures on the 
timing of cladding rupture and on the peak cladding temperature. 
This analysis concluded that the peak cladding temperature for Cycle 
6 remained below that of the analysis of record and that the peak 
cladding temperature continued to occur at 

[[Page 45174]]
low burnup, specifically the burnup corresponding to the maximum 
initial fuel stored energy.
    In addition to the LOCA analysis a DNB propagation analysis was 
performed to demonstrate that DNB propagation does not occur during 
postulated accidents that experience DNB when pressure in a fuel pin 
is higher than the system pressure. This analysis was performed 
using the fuel rod strain model described in CEN-372-P-A.
    Based on these analyses, there is no increase in the probability 
or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve any change to the 
configuration or method of operation of any plant equipment that is 
used to mitigate the consequences of an accident. Accordingly, no 
new failure modes have been defined for any plant system or 
component important to safety nor has any new limiting failure been 
identified as a result of the proposed change. The intent of the 
proposed change is to utilize a new analytical method to ensure that 
the consequences of any equipment malfunction remain within the 
limits of existing analyses resulting in no impact on radiological 
consequences.
    The impact of the maximum fuel rod gas pressures calculated for 
U3C6 was evaluated as part of the Cycle 6 ECCS performance analysis. 
Except for the highest burnup analyzed, the time of cladding rupture 
decreased as the initial fuel rod gas pressure increased with 
burnup. However, the peak cladding temperature occurred at the 
burnup with the maximum initial fuel stored energy. The analysis 
also determined that the ECCS performance analysis for U3C6 is 
bounded by that of the reference cycle analysis.
    An evaluation was conducted to ensure that fuel would not 
experience DNB propagation when the pressure in a fuel pin is higher 
than the system pressure. DNB was shown not to propagate by 
demonstrating that the degree of cladding deformation is no more 
than the limit defined by the fuel rod maximum pressure Topical 
Report (CEN-372-P-A).
    Therefore, it can be concluded that the proposed change to 
Section 6.9.1.10 does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change adds an NRC approved Topical Report 
(methodology) and its associated SER, to the list of analytical 
methods used to determine core operating limits. The use of the new 
methodology ensures that safety margins are maintained within the 
results of existing calculations. Since the core operating limits 
will continue to be established by an NRC approved methodology and 
will provide adequate core protection, the proposed amendment does 
not involve a significant reduction in the margin of safety.
    Analyses were conducted to determine the impact of higher fuel 
rod pressure on ECCS performance and DNB propagation. The results of 
the analyses show that the effects of higher fuel rod pressure are 
bounded by previous results.
    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involve no significant hazards consideration. Local 
Public Document Room location: Phoenix Public Library, 1221 N. Central 
Avenue, Phoenix, Arizona 85004
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999
    NRC Project Director: William H. Bateman

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

    Date of amendments request: July 13, 1995
    Description of amendments request: The proposed amendments would 
revise the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, 
Technical Specifications (TSs) Section 5.2.1, ``Fuel Assemblies.'' The 
current TSs only allow fuel that is clad with either zircaloy or ZIRLO. 
The proposed change would allow the use of cladding material other than 
zircaloy or ZIRLO with an approved exemption. Thus, the proposed change 
will eliminate the need for future amendments to allow the use of 
different cladding material for which the Commission has issued an 
exemption.
    Basis for proposed no significant hazards consideration 
determination: As requied by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    Calvert Cliffs Technical Specification 5.2.1, Fuel Assemblies, 
states that fuel rods are clad with either zircaloy or ZIRLO. This 
reflects the requirements of 10 CFR 50.44, 50.46, and 10 CFR [Part] 
50, Appendix K, which also restrict fuel rod cladding materials to 
zircaloy or ZIRLO. Baltimore Gas and Electric Company proposes to 
insert fuel assemblies into Calvert Cliffs Unit 1 which have some 
fuel rods clad in zirconium alloys that do not meet the definition 
of zircaloy or ZIRLO for testing purposes and has applied for an 
exemption to the regulations to allow that change. The proposed 
change to the Calvert Cliffs Technical Specifications will allow the 
use of cladding materials that are not zircaloy or ZIRLO with an 
approved exemption in accordance with 10 CFR 50.12.
    The proposed change to the Unit 1 and Unit 2 Technical 
Specifications will allow the use of fuel rod cladding materials 
other than zircaloy or ZIRLO as long as those materials have been 
approved by an exemption to the regulations. To obtain approval of 
new cladding materials, 10 CFR 50.12 requires that the applicant 
show that the proposed exemption is authorized by law, is consistent 
with the common defense and security, will not present an undue risk 
to the public health and safety; and is accompanied by special 
circumstances.
    Under the proposed change, any fuel rod cladding materials that 
are not zircaloy or ZIRLO must still be approved by the Nuclear 
Regulatory Commission (NRC) prior to use under 10 CFR 50.12. This 
change to the Technical Specifications allows the NRC to approve the 
use of cladding materials that are not either zircaloy or ZIRLO 
under 10 CFR 50.12 and not require an additional approval under 10 
CFR 50.90. As such, the proposed change eliminates a duplicative 
regulatory requirement and would have no effect on the probability 
or consequences of an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The proposed change eliminates a duplicated approval requirement 
and would have no effect on the possibility of a new or different 
type of accident. The proposed change to the Technical 
Specifications would allow the NRC to approve the use of fuel rod 
cladding materials that are not either zircaloy or ZIRLO under 10 
CFR 50.12 and not require an additional approval under 10 CFR 50.90.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The proposed change eliminates a duplicated approval requirement 
and will have no effect on the margin of safety. The proposed change 
to the Technical Specifications would allow the NRC to approve the 
use of fuel rod cladding materials that are not either zircaloy or 
ZIRLO under 10 CFR 50.12, and not require an additional approval 
under 10 CFR 50.90.
    Therefore, the proposed change does not involve a significant 
reduction in a margin 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 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and 

[[Page 45175]]
Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Ledyard B. Marsh

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

    Date of application for amendment requests: April 24, 1995
    Description of amendment requests: The licensee proposes to amend 
Section 6 of the Technical Specifications of all ComEd stations to make 
the following changes: (1) delete the ``Review, Investigative and Audit 
Functions'' sections, in their entirety, and relocate these 
requirements to appropriate sections of the ComEd Quality Assurance 
Topical Report, (2) change titles to reflect the reorganization of 
ComEd's Nuclear Operations Division, and (3) miscellaneous 
administrative and editorial changes.
    Basis for proposed no significant hazards consideration 
determination: As requied 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.
    (1) The proposed relocation of the ``Review, Investigative and 
Audit Functions'' sections of Technical Specifications to the QA 
Topical Report does not affect any accident initiators or 
precursors, and does not change or alter the design assumptions for 
the systems and components used to mitigate the consequences of an 
accident.
    The relocation of these sections is consistent with the 
recommended changes specified in the October 25, 1993 letter from W. 
T. Russell (USNRC) to the Chairpersons of the Owner Groups' 
Technical Specifications Committees, entitled, ``Content of Standard 
Technical Specifications, Section 5.0, Administrative Controls''.
    Relocating these requirements to the QA Topical Report will 
continue to ensure that proposed future changes to these 
requirements will receive proper regulatory oversight. NRC review of 
the Quality Assurance Program is governed by 10CFR50.54. 
10CFR50.54(a)(3) states: ``Changes to the quality assurance program 
description that do not reduce the commitments must be submitted to 
the NRC in accordance with the requirements of 50.71. Changes to the 
quality assurance program description that do reduce the commitments 
must be submitted to NRC and receive NRC approval prior to 
implementation, ...'' Based on these 10CFR50.54 requirements, 
appropriate licensee and regulatory control of the requirements in 
the subject relocated Technical Specification sections will be 
maintained.
    (2) The proposed title and organizational 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.
    Commonwealth Edison's organizational changes allow for increased 
senior management attention and oversight of station activities. 
Position titles and associated responsibilities have changed to 
increase the company's efficiency in the management of its nuclear 
stations. These administrative changes do not reduce any 
requirements or commitments. The proposed changes enhance the 
administrative controls necessary to ensure safe plant operation.
    (3) Other proposed administrative/editorial changes simply make 
corrections or provide needed clarification prompted by the 
reorganization. These changes provide consistency with station 
procedures, programs, other Technical Specifications, and Standard 
Technical Specifications. They are administrative in nature and do 
not impact any accident previously evaluated in the UFSAR.
    In conclusion, none of the proposed changes involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    B. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    (1) The proposed relocation of the ``Review, Investigative and 
Audit Functions'' sections of Technical Specifications to the QA 
Topical Report does 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 that could give rise to a new or 
different kind of accident that was previously evaluated.
    The proposed changes are considered to be administrative or 
programmatic in nature and do not affect equipment or components 
that could initiate an accident. All administrative commitments 
being relocated to the QA Topical Report will continue to receive 
appropriate regulatory oversight pursuant to 10CFR50.54.
    (2) The proposed title and organization 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 
that will enhance the performance of organizations responsible for 
the safe operation of the plant to respond to plant transients or 
emergencies. All responsibilities described in Technical 
Specifications for management activities will continue to be 
performed by qualified individuals.
    (3) All other proposed changes are administrative in nature and 
do not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    In conclusion, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    C. The proposed changes do not involve a significant reduction 
in a margin of safety.
    (1) The proposed changes are administrative or programmatic in 
nature and do not affect the margin of safety for any safety 
parameters and setpoints addressed in Technical Specifications. The 
assumptions, initial conditions and methodologies used in the 
accident analyses remain unchanged, therefore, accident analyses 
results are not impacted.
    Placing these requirements in QA Topical Report will continue to 
ensure that proposed future changes to these requirements will 
receive proper regulatory oversight pursuant to 10CFR50.54.
    (2) The proposed title and organizational changes are 
administrative in nature and 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 analyses results are not impacted.
    (3) All other proposed changes are administrative in nature and 
have no impact on the margin of safety for any Technical 
Specification.
    In conclusion, 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: for Braidwood, the Wilmington 
Public Library, 201 S. Kankakee Street, Wilmington, Illinois 60481; for 
Byron, the Byron Public Library District, 109 N. Franklin, P.O. Box 
434, Byron, Illinois 61010; for Dresden, Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450; for LaSalle, 
Jacobs Memorial Library, Illinois Valley Community College, Oglesby, 
Illinois 61348; for Quad Cities, Dixon Public Library, 221 Hennepin 
Avenue, Dixon, Illinois 61021; for Zion, Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085

[[Page 45176]]

    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, Illinois Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, 
Will County, Illinois Docket Nos. 50-373 and 50-374, LaSalle County 
Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: June 8, 1995
    Description of amendment request: The proposed amendments would 
revise Technical Specifications Section 3/4.8, Electrical Power 
Systems, and the associated Bases for LaSalle County, Byron, and 
Braidwood Stations. The proposed changes revise surveillance and 
administrative requirements associated with emergency diesel generators 
(EDGs) in accordance with the guidance of NRC Generic Letter 94-01, 
``Removal of Accelerated Testing and Special Reporting Requirements for 
Emergency Diesel Generators,'' Generic Letter 93-05, ``Line-Item 
Technical Specifications Improvements to Reduce Surveillance 
Requirements for Testing During Power Operation,'' and Regulatory Guide 
(RG) 1.9, ``Selection, Design, Qualification, and Testing of Emergency 
Diesel Generator Units Used as Class 1E Onsite Electric Power Systems 
at Nuclear Power Plants.'' The proposed changes include: (1) 
eliminating increased testing requirements for EDGs, (2) eliminating 
special reporting requirements for EDGs, (3) eliminating the semi-
annual fast load test and replacing it with a requirement to load EDGs 
semi-annually in accordance with the vendor recommendations for all 
test purposes other than the refueling outage Loss of Offsite Power 
(LOOP) tests, (4) de-coupling the 24-hour endurance run and the LOOP/
loss-of-coolant (LOCA) (LOOP only for LaSalle) sequencing requirements 
for the hot start test, (5) removing RG 1.108 references to testing 
requirements, (6) eliminating testing requirements when an EDG becomes 
inoperable due to an inoperable support system, an independently 
testable component, or preplanned maintenance or testing, or if there 
is not a potential common mode failure for the remaining diesel 
generator, (7) deleting the requirement for inspecting the EDGs in 
accordance with procedures prepared in conjunction with its 
manufacturer's recommendations, and (8) making editorial changes.
    Basis for proposed no significant hazards consideration 
determination: As requied 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 do not affect accident initiators or 
precursors and do not alter the design assumptions affecting the 
ability of the EDGs to mitigate the consequences of an accident.
    Deleting the special reporting requirements from the Technical 
Specifications is administrative. ComEd will continue to notify the 
Commission of significant EDG failures in accordance with 10 CFR 
50.72 and 50.73 criteria.
    Excessive testing requirements have proven to be a contributor 
to increased equipment degradation. Removing inappropriate and 
redundant requirements increases EDG reliability and enhances the 
ability of EDGs to mitigate the consequences of an accident. 
Implementing ComEd's alternative to the maintenance rule for the 
EDGs provides additional assurance that high EDG performance will be 
maintained.
    EDG equipment degradation will be reduced by eliminating the 
semi-annual fast load test for EDGs in accordance with the vendor 
recommendations for test purposes other than the refueling outage 
Loss of Offsite Power (LOOP) tests. This improves EDG reliability 
and availability and further enhances their ability to mitigate the 
consequences of an accident. The LOOP test would still be performed 
to provide assurance that the EDG is capable of responding to a LOOP 
as assumed in the accident analyses.
    De-coupling the 24 hour endurance test and the LOOP/LOCA (for 
LaSalle, LOOP) sequencing test requirements for the hot start test 
has no effect on accident mitigation. Demonstrating diesel generator 
hot restart capability without loading the engine does not 
invalidate or reduce the effectiveness of the hot restart test. The 
hot restart test can be conducted in any plant condition since its 
performance at power will have no adverse effect on plant 
operations.
    The proposed editorial changes are administrative in nature. 
They improve readability and provide consistency with current 
industry guidance.
    Therefore, the proposed changes do not involve an 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:
    The proposed changes do not alter the ability of the EDGs to 
perform their intended function to mitigate the consequences of an 
initiating event within the acceptance limits assumed in plant 
safety analyses. The proposed changes have no impact on component or 
system interactions, or the plant design basis.
    Instrumentation setpoints, starting, sequencing, and loading 
functions associated with EDGs are not affected by the proposed 
changes. Furthermore, combining the alternate EDG system maintenance 
rule implementation program with the proposed amendment will enhance 
both the availability and the performance of the EDGS.
    Therefore, there is not a potential for creating the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3) Involve a significant reduction in a margin of safety:
    The proposed changes do not increase the probability or 
consequences of an accident, and there is no impact on equipment 
design or operation. The proposed changes do not affect the results 
of accident and transient analyses. Plant and system response to an 
initiating event will remain in compliance within the assumptions of 
safety analyses. There is no associated change to the type, amount, 
or control of radioactive effluents, nor is there an associated 
increase in individual or cumulative occupational radiation 
exposure. There is no effect upon the capabilities of the associated 
systems to perform their intended functions within the allowed 
response times assumed in safety analyses.
    The proposed changes are compatible with plant operating 
experience and are consistent with the guidance provided in NUREG-
1366, Generic Letters 93-05 and 94-01, and Regulatory Guide 1.9. In 
two instances ComEd's proposed changes deviate from these guidance 
documents. However, the changes are consistent with the intent of 
the documents or other NRC guidance documents. Eliminating excessive 
testing requirements can improve safety by reducing challenges to 
plant systems and reducing equipment wear and degradation. While the 
proposed changes affect surveillance intervals; there are no changes 
to the methods used to perform the surveillances.
    EDG reliability and availability will be improved by the 
proposed changes. The surveillances will continue to demonstrate the 
ability of the EDGs to perform their intended function of providing 
electrical power to the emergency safety systems needed to mitigate 
design basis transients. No margin of safety is reduced.
    Guidance has been provided in ``Final Procedures and Standards 
on No Significant Hazards Considerations,'' Final Rule, 51 FR 7744, 
for the application of standards to license change requests for 
determination of the existence of significant hazards 
considerations. This document provides examples of amendments which 
are and are not considered likely to involve significant hazards 
considerations. These proposed amendments most closely fit the 
example of a change which may either result in some increase to the 
probability or consequences of a previously analyzed accident or may 
reduce in some way a safety margin, but where the results of the 
change are clearly within all acceptance criteria with respect to 
the system or component specified in the standard review plan.
    This proposed amendment does not involve a significant 
relaxation of the criteria used to establish safety limits, a 
significant 

[[Page 45177]]
relaxation of the bases for the limiting safety system settings, or a 
significant relaxation of the bases for the limiting conditions for 
operations. The proposed change does not reduce the margin of safety 
as defined in the basis for any Technical Specification.
    Therefore, based on the guidance provided in the Federal 
Register and the criteria established in 10 CFR 50.92(c), ComEd has 
concluded that the proposed change does not constitute 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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481; for LaSalle, Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

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

    Date of application for amendment requests:  August 30, 1994, as 
supplemented August 4, 1995.
    Description of amendment requests: As a result of findings by a 
Diagnostic Evaluation Team inspection performed by the NRC staff at the 
Dresden Nuclear Power Station in 1987, Commonwealth Edison Company 
(ComEd, the licensee) made a decision that both the Dresden Nuclear 
Power Station and sister site Quad Cities Nuclear Power Station needed 
attention focused on the existing custom Technical Specifications (TS) 
used.
    The licensee made the decision to initiate a Technical 
Specification Upgrade Program (TSUP) for both Dresden and Quad Cities. 
The licensee evaluated the current TS for both Dresden and Quad Cities 
against the Standard Technical Specifications (STS) contained in NUREG-
0123, ``Standard Technical Specifications General Electric Plants BWR/
4.'' The licensee's evaluation identified numerous potential 
improvements such as clarifying requirements, changing TS to make them 
more understandable and to eliminate interpretation, and deleting 
requirements that are no longer considered current with industry 
practice. As a result of the evaluation, ComEd has elected to upgrade 
both the Dresden and Quad Cities TS to the STS contained in NUREG-0123.
    The TSUP for Dresden and Quad Cities is not a complete adaption of 
the STS. The TSUP focuses on (1) integrating additional information 
such as equipment operability requirements during shutdown conditions, 
(2) clarifying requirements such as limiting conditions for operation 
and action statements utilizing STS terminology, (3) deleting 
superseded requirements and modifications to the TS based on the 
licensee's responses to Generic Letters (GL), and (4) relocating 
specific items to more appropriate TS locations.
    The August 30, 1994, and August 4, 1995, applications proposed to 
upgrade only Section 3/4.2 (Instrumentation) of the Dresden and Quad 
Cities TS.
    Basis for proposed no significant hazards consideration 
determination: As requied 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 
because:
    In general, the proposed amendment represents the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis. 
Implementation of these changes will provide increased reliability 
of equipment assumed to operate in the current safety analysis, or 
provide continued assurance that specified parameters remain within 
their acceptance limits, and as such, will not significantly 
increase the probability or consequences of a previously evaluated 
accident.
    Some of the proposed changes to the current Technical 
Specifications (CTS) represent minor curtailments of the current 
requirements which are based on generic guidance or previously 
approved provisions for other stations. The proposed amendment for 
Dresden and Quad Cities Station's Technical Specification Section 3/
4.2 are based on BWR-STS (NUREG-0123, Revision 4 ``Standard 
Technical Specifications General Electric Plants BWR/4) guidance or 
NRC accepted changes at later operating BWR plants. Any deviations 
from BWR-STS and CTS requirements do not significantly increase the 
probability or consequences of any previously evaluated accident for 
Dresden and Quad Cities Station. These proposed changes are 
consistent with the current safety analyses and have been previously 
determined to represent sufficient requirements for the assurance 
and reliability of equipment assumed to operate in the safety 
analysis, or provide continued assurance that specified parameters 
remain within their acceptance limits. As such, these changes will 
not significantly increase the probability or consequences of a 
previously evaluated accident.
    The associated systems that make up the Instrumentation Systems 
are not assumed in any safety analysis to initiate any accident 
sequence for both Dresden and Quad Cities Stations; therefore, the 
probability of any accident previously evaluated is not increased by 
the proposed amendment. In addition, the proposed surveillance 
requirements for the proposed amendments to these systems are 
generally more prescriptive than the current requirements specified 
within the Technical Specifications. These more prescriptive 
surveillance requirements increase the probability that the 
Instrumentation Systems will perform their intended functions. 
Therefore, the proposed TS will improve the reliability and 
availability of all affected systems and reduce the consequences of 
any accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated because:
    In general, the proposed amendment represents the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis. Others 
represent minor curtailments of the current requirements which are 
based on generic guidance or previously approved provisions for 
other stations. These changes do not involve revisions to the design 
of the station, other than technically valid trip setpoint changes. 
Some of the changes may involve revision in the operation of the 
station; however, these changes provide additional restrictions 
which are in accordance with the current safety analyses, or are to 
provide for additional testing or surveillances which will not 
introduce new failure mechanisms beyond those already considered in 
the current safety analyses. Therefore, these changes will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    The proposed amendment for Dresden and Quad Cities Station's 
Technical Specification Section 3/4.2 is based on BWR-STS guidelines 
or NRC accepted changes at later operating BWR plants. The proposed 
amendment has been reviewed for acceptability at the Dresden and 
Quad Cities Nuclear Power Stations considering similarity of system 
or component design versus the BWR-STS or later operating BWRs. Any 
deviations from BWR-STS or CTS requirements do not create the 
possibility of a new or different kind of accident than previously 
evaluated for Dresden and Quad Cities Stations. No new modes of 
operation are introduced by the proposed changes. Various 
surveillance requirements are changed to reflect improvements in 
technique, frequency of performance or operating experience at later 
plants. Proposed changes to action statements in many places add 
requirements that are not in the present technical specifications or 
adopt 

[[Page 45178]]
requirements that have been used at other operating BWRs with designs 
similar to Dresden and Quad Cities. The proposed changes maintain at 
least the present level of operability. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The associated systems that make up the Instrumentation Systems 
are not assumed in any safety analysis to initiate any accident 
sequence for Dresden or Quad Cities Stations. In addition, the 
proposed surveillance requirements for affected systems associated 
with the Instrumentation Systems are generally more prescriptive 
than the current requirements specified within the Technical 
Specifications; therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Involve a significant reduction in the margin of safety 
because:
    In general, the proposed amendment represents the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis. Others 
represent minor curtailments of the current requirements which are 
based on generic guidance or previously approved provisions for 
other stations. Some of the later individual items may introduce 
minor reductions in the margin of safety when compared to the 
current requirements. However, other individual changes are the 
adoption of new requirements which will provide significant 
enhancement of the reliability of the equipment assumed to operate 
in the safety analysis, or provide enhanced assurance that specified 
parameters remain within their acceptance limits. These enhancements 
compensate for the individual minor reductions, such that taken 
together, the proposed changes will not significantly reduce the 
margin of safety.
    The proposed amendment to Technical Specification Section 3/4.2 
implements present requirements in accordance with the guidelines 
set forth in the BWR-STS. Any deviations from BWR-STS and CTS 
requirements do not significantly reduce the margin of safety for 
Dresden and Quad Cities Stations. The proposed changes are intended 
to improve readability, usability, and the understanding of 
technical specification requirements while maintaining acceptable 
levels of safe operation. The proposed changes have been evaluated 
and found to be acceptable for use at Dresden and Quad Cities based 
on system design, safety analysis requirements and operational 
performance. Since the proposed changes are based on NRC accepted 
provisions at other operating plants that are applicable at Dresden 
and Quad Cities and maintain necessary levels of system or component 
readability, the proposed changes do not involve a significant 
reduction in the margin of safety.
    The proposed amendment for Dresden and Quad Cities Stations will 
not reduce the availability of systems associated with the 
Instrumentation Systems when required to mitigate accident 
conditions; therefore, the proposed changes do not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: for Dresden, Morris Public 
Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities, 
Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021
    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. 50-295 and 50-304, Zion 
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

    Date of amendment request: March 8, 1995, as supplemented June 1, 
1995
    Description of amendment request: The proposed amendments would 
revise the secondary undervoltage setpoint.
    Basis for proposed no significant hazards consideration 
determination: As requied 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 an increase in the 
probability of occurrence or consequences of any accident previously 
evaluated.
    The proposed amendment does not change the fundamental function 
or capability of the Secondary Undervoltage protection as described 
in UFSAR section 8.3. Inadvertent or spurious operation of the 
Secondary Undervoltage protection function will initiate loading of 
the safe shutdown loads on the diesel generators and is not assumed 
to initiate an accident. The proposed Secondary Undervoltage 
setpoints are low enough to prevent spurious actuations given the 
expected off site grid voltages.
    This change does not affect the initiators or precursors of any 
accident previously evaluated. This change will not increase the 
likelihood that a transient initiating event will occur because 
transients are initiated by equipment malfunction and/or 
catastrophic system failure. The change in setpoints for the 
Secondary Undervoltage protection system does involve some changes 
to existing plant equipment (such as transformer tap changes and 
Circulating Water pump excitation circuit changes). However, all 
changes to existing plant equipment have been or will be evaluated 
in accordance with the requirements of 10CFR50.59 prior to 
installation, to determine that no unreviewed safety questions exist 
with regard to the plant changes.
    Since any design changes have been or will be determined to be 
acceptable per 10CFR50.59 prior to installation and no new plant 
equipment will be installed, the probability of occurrence of 
accidents previously evaluated will not increase.
    With Zion Station's new Auxiliary Power System configuration and 
the proposed Secondary Undervoltage setpoints, the probability of a 
Loss of Off-Site Power (LOOP) is actually reduced since the original 
Auxiliary Power System configuration and Secondary Undervoltage 
setpoints required a higher grid voltage to ensure that safety 
related loads would be powered from Off-Site power sources during a 
design basis accident.
    The consequences of accidents previously evaluated are not 
increased. The proposed change does not affect the required level of 
availability or systems required to mitigate the accidents 
considered in the Analyses. Administrative controls will be in place 
to ensure that the installed setpoints are low enough to ensure that 
the Emergency Diesel Generators are not unnecessarily challenged. 
The proposed changes will increase the level of confidence that the 
ESF equipment will be capable of starting and operating during a 
design basis accident with degraded off-site grid voltage. The 
increase in the level of confidence is the result of the more 
rigorous methodology used to determine limited ESF bus voltages, 
given the minimum expected off-site AC voltage. Based on the 
previous discussion, it is determined that there will be no 
significant increase in the consequences of any accident previously 
evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any previously analyzed.
    The proposed Secondary Undervoltage setpoint change does not 
change the design of the Secondary Undervoltage protection system or 
its function to protect against degraded offsite power. Actuation of 
the Secondary Undervoltage protection system will initiate a 
sequence of events that will start the Emergency Diesel Generator 
(EDG) for the associated ESF bus, strip all loads from the bus, open 
all feed breakers to the bus, close the Emergency feed breaker (thus 
energizing the bus from the EDG), and initiate sequenced starting of 
the Safe Shutdown equipment supplied by the bus, including a Service 
Water pump, Component Cooling Water pump, Auxiliary Feedwater pump, 
and Reactor Containment Fan Cooler(s), as applicable.
    The proposed change does not involve the addition of any new or 
different types of equipment, nor does it involve the operation of 
equipment required for safe operation of the facility in a manner 
different from those addressed in the Final Safety Analysis Report. 
No safety related equipment or function will be altered as a result 
of this proposed change. Because no new failure modes are 
introduced, the proposed amendment does not create a new or 
different kind of accident from any previously analyzed in the 
UFSAR.
    Based on the above discussion, the proposed amendment does not 
create a new or different kind of accident from any previously 
analyzed in the UFSAR.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.

[[Page 45179]]

    The proposed amendment will allow the Secondary Undervoltage 
setpoint to be conservatively established based on new engineering 
calculations which consider the lowest expected offsite grid voltage 
and operation of all required ESF equipment under design basis 
accident loading conditions.
    The proposed Secondary Undervoltage setpoints will provide 
increased confidence that adequate bus voltage will be available to 
support starting and operation of all required ESF loads. The 
proposed setpoint includes worst case instrument error to ensure 
that the lowest possible voltage will not be lower than the degraded 
voltage analytical limits. Additionally, the proposed setpoints are 
low enough to prevent spurious actuations due to expected 
fluctuations in the grid voltage. The new setpoints are based on a 
minimum expected grid voltage of 343 kV, with added margin. The 
proposed changes will provide an increase in the level of protection 
that currently exists and will ensure the margin of safety is 
adequately maintained.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam 
Neck Plant, Middlesex County, Connecticut

    Date of amendment request: August 3, 1995
    Description of amendment request: The proposed amendment will add 
an one-time footnote to Technical Specification (TS) Section 3/4.7.12, 
``Ultimate Heat Sink,'' to increase the allowed outage time from 6 
hours to 18 hours for the months of August and September. In addition, 
also for the months of August and September, the maximum service water 
limit will be elevated from 90 deg.F to 95 deg.F.
    Basis for proposed no significant hazards consideration 
determination: As requied 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 analyzed.
    The proposed addition of a 12 hour time period to monitor the 
ultimate heat sink temperature to the Technical Specification 
Limiting Condition for Operation action statements does not involve 
an increase in the probability of an accident previously evaluated. 
The probability of an accident previously evaluated is not increased 
by a short-term increase in the ultimate heat sink temperature. An 
evaluation has been performed that safe shutdown will be achieved 
and maintained for a loss of normal AC power event with the 
additional consideration of a single failure with service water 
inlet temperatures as high as 95 deg.F. In addition, an evaluation 
of the credible FSAR Chapter 15 events with AC power available and 
no isolation of non-essential service water loads has been performed 
that demonstrates that safe shutdown will be achieved and 
maintained. There has been no significant increase in the 
consequences of these events previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed technical specification change does not create the 
possibility of a new or different kind of accident previously 
analyzed. The addition of a 12 hour time period to monitor the 
ultimate heat sink temperature increases the amount of time that is 
allowed for the plant to be in Hot Standby from 6 to 18 hours should 
the ultimate heat sink temperature increase above 90 deg.F. This 
extension of the time allowed for the plant to be in Hot Standby 
does not change the plant configuration. As such, the change does 
not create the possibility of a new or different kind of accident 
previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    The proposed technical specification change does not involve a 
significant reduction in the margin of safety. The addition of a 12 
hour time period to monitor the ultimate heat sink temperature 
increases the time required for the plant to be in Hot Standby from 
6 to 18 hours should the ultimate heat sink temperature exceed 
90 deg.F. An evaluation has been performed to demonstrate that the 
risk significance associated with the increased action time is very 
low. In addition, safe shutdown capability has been demonstrated for 
service water inlet temperatures as high as 95 deg.F.
    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: Russell Library, 123 Broad 
Street, Middletown, CT 06457
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: May 5, 1995
    Description of amendment request: The proposed amendment would 
change the surveillance frequency of radiation area, and effluent and 
process monitors from monthly to quarterly; and the required frequency 
for minimum exercise of control element assemblies also from monthly to 
quarterly.
    Basis for proposed no significant hazards consideration 
determination: As requied by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration. The 
NRC staff has reviewed the licensee's analysis against the standards of 
10 CFR 50.92(c). The staff's review is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. Extending surveillance test intervals as proposed will 
reduce the probability of inadvertent reactor scrams and ensuing 
challenges to safety systems. This is accomplished by reducing the 
occasions and thus the total time that the subject systems are 
removed from their ``normal'' configuration and placed into the 
required ``test'' configuration. In addition, the probability of 
test-induced failures, or failures caused by human error, is 
likewise decreased. Thus, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Extending surveillance test intervals as proposed will not 
require installation of any new or different equipment, and will not 
alter or otherwise modify existing plant equipment. Thus, the 
proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Independent research has found that equipment failures and 
personnel errors during several types of surveillance tests caused a 
significant number of reactor scrams and attendant unnecessary 
challenges to safety equipment. The results of this research have 
been corroborated by the licensee's plant specific operating 
experience. The licensee concludes that the reduced test intervals 
proposed in this amendment remain sufficient to ensure known 
phenomena, such as instrument setpoint drift and random hidden 
failures, remain within the assumptions of the safety analysis. 
Thus, the proposed change does not involve a significant reduction 
in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) 

[[Page 45180]]
are satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, 329 Bath Road, Brunswick, ME 04011
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: July 24, 1995.
    Description of amendment request: The proposed amendment would 
delete Table 3.4-1, ``Reactor Coolant System Pressure Isolation 
Valves'' from the Seabrook Station, Unit No. 1 Technical Specification 
section 3.4.6.2. Reference to Table 3.4-1 also would be deleted from 
Limiting Condition for Operation 3.4.6.2 f and from Surveillance 
Requirement 4.4.6.2.2. The information contained in Table 3.4-1 would 
be relocated to the Technical Requirements Manual.
    Basis for proposed no significant hazards consideration 
determination: As requied by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration. The 
NRC staff has reviewed the licensee's analysis against the standards of 
10 CFR 50.92(c). The NRC staff's review is presented below.
    A. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
(10 CFR 50.92(c)(1)) because they do not in any way alter the 
operability or surveillance requirements for pressure isolation 
valves. The proposed changes merely delete a listing of valves which 
are designated as pressure isolation valves in accordance with the 
definition provided in 10 CFR Part 50. Therefore, neither the 
probability nor consequences of previously evaluated accidents are 
affected.
    B. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
(10 CFR 50.92(c)(2)) because they do not affect in any way the 
manner by which the facility is operated or make any changes in 
structures, systems, or components which could affect the 
operational characteristics of the facility.
    C. The proposed changes do not involve a significant reduction 
in a margin of safety (10 CFR 50.92(c)(3)) because the proposed 
changes do not affect the operability requirements or surveillance 
testing of any pressure isolation valve and do not affect in any way 
the manner by which the facility is operated or involve equipment or 
features which affect the operational characteristics of the 
facility.
    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: Exeter Public Library, 
Founders Park, Exeter, NH 03833
    Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast 
Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: July 28, 1995
    Description of amendment request: The proposed amendment adds 
Technical Specifications (TS) to Section 3.10, Refueling and Spent Fuel 
Handling. Specifically, the proposed TS (with applicability, action, 
and surveillance requirements) will require that: (1) the reactor be 
subcritical for at least 100 hours before the start of reactor 
refueling operations, (2) the spent fuel pool bulk temperature be 
maintained less than or equal to 140 deg.F, and (3) two trains of 
shutdown cooling be operable during reactor refueling operations. In 
support of the request, NNECO proposes to: (1) use the ORIGEN2 code to 
more accurately predict decay heat loads from the spent fuel, (2) use 
the ONEPOOL code to credit the effect of evaporative cooling on the 
spent fuel pool bulk temperature, and (3) take credit for both trains 
of shutdown cooling to assist the spent fuel pool cooling system during 
refueling outages. In addition, the proposed amendment modifies the 
table of contents and associated Bases section to reflect the changes.
    Basis for proposed no significant hazards consideration 
determination: As requied by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    NNECO has reviewed the proposed changes in accordance with 
10CFR50.92 and concluded that the changes do not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
compromised. The proposed changes do not involve an SHC because the 
changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed license amendment will allow NNECO to use the 
shutdown cooling system (SCS) to assist the spent fuel pool cooling 
(SFPC) system to cool the spent fuel pool during refueling outages. 
This amendment request does not affect: the number of spent fuel 
bundles allowed in the spent fuel pool, spent fuel pool criticality 
analysis, structural analysis of the spent fuel pool, or 
radiological release scenarios.
    The proposed license amendment also allows NNECO to use ORIGEN2 
and ONEPOOL codes. The ORIGEN2 code more accurately predicts decay 
heat loads from the spent fuel in the spent fuel pool. The ONEPOOL 
code credits the effect of evaporative cooling on the spent fuel 
pool bulk temperature. The use of these codes will improve the 
accuracy of predicting spent fuel pool bulk temperatures during 
normal and abnormal refueling scenarios.
    The use of the SCS to assist the SFPC system to cool the spent 
fuel pool will allow the movement of spent fuel to begin 100 hours 
after reactor shutdown. The existing accident analysis for a dropped 
spent fuel bundle during refueling bounds this situation as the 
analysis assumed a decay time of 24 hours.
    The three new proposed technical specifications will provide 
sufficient controls on the movement of spent fuel into the spent 
fuel pool, bulk temperature of the spent fuel pool and operability 
of the shutdown cooling system to operate within analysis 
assumptions during refueling operations at Millstone Unit No. 1.
    Therefore, based on the above, the use of the SCS to assist the 
SFPC system to cool the spent fuel pool during refueling outages, 
the use of the ORIGEN2 code, the use of the ONEPOOL code, and the 
addition of three technical specifications will not involve a 
significant increase in the probability or consequences of an 
accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed license amendment to use the SCS to assist the SFPC 
system to cool the spent fuel pool will allow SCS train B to cool 
the spent fuel pool in a method similar to train A.
    The proposed license amendment to use ORIGEN2 and ONEPOOL codes 
to predict spent fuel pool bulk temperatures will increase the 
accuracy of analyzing normal and abnormal refueling scenarios.
    The three new proposed technical specifications will 
sufficiently control refueling operations to support analyzed 
accident scenarios.
    Therefore, the use of the SCS to assist the SFPC system to cool 
the spent fuel pool, the use of the ORIGEN2 code, the use of ONEPOOL 
code and the addition of three technical specifications do not 
create the possibility of a new or different kind of accident from 
any previously analyzed.
    3. Involve a significant reduction in the margin of safety.
    The proposed license amendment to use the SCS to assist the SFPC 
system to cool the spent fuel pool will allow the crediting of the 
SCS and SFPC system to remove heat from 

[[Page 45181]]
the spent fuel pool during normal refueling scenarios. The analysis 
demonstrates that this cooling configuration will maintain the spent 
fuel pool bulk temperature below the pool design limit of 140 deg.F 
with a postulated single active failure.
    The addition of the train B SCS cross-tie does not adversely 
affect the existing design basis of the SCS to remove sensible and 
decay heat from the reactor water, cool it from 280 deg.F to 
125 deg.F within 24 hours, and to maintain the reactor water at 
125 deg.F.
    The proposed license amendment to use ORIGEN2 and ONEPOOL codes 
will improve the accuracy of predicting spent fuel pool bulk 
temperatures during normal and abnormal refueling scenarios.
    The thermal hydraulic analysis most limiting time to boil 
calculation of 5.4 hours for loss of all forced cooling to the spent 
fuel pool is consistent with assumed operator response times for 
similar scenarios.
    The three new proposed technical specifications will ensure that 
the margin of safety established by engineering analysis of 
refueling operations is maintained.
    Therefore, based on the above, the use of the SCS to assist the 
SFPC system to cool the spent fuel pool, the use of the ORIGEN2 
code, the use of the ONEPOOL code, and the addition of three 
technical specifications does not involve a significant reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270
    NRC Project Director: Phillip F. McKee

Northeast Nuclear Energy Company, et al., Docket Nos. 50-245, 50-
336 and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2, 
and 3, New London, Connecticut

    Date of amendment request: August 4, 1995
    Description of amendment request: The proposed license amendments 
will modify the Administrative Controls Section (Section 6) of the 
Millstone Unit Nos. 1, 2, and 3 Technical Specifications to allow the 
Plant Operations Review Committee (PORC) and Site Operations Review 
Committee (SORC) to direct its efforts in the review of more critical 
safety matters which affect day-to-day operation. This will be 
accomplished by the establishment of a Station Qualified Reviewer 
Program (SQRP) and the reassignment of certain procedure approvals to 
designated managers in lieu of approval by PORC/SORC.
    Basis for proposed no significant hazards consideration 
determination: As requied by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration 
(SHC), which is presented below:
    ...These proposed changes do not involve an SHC because the 
changes do not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    These changes are administrative in nature. They do not involve 
any modifications to plant systems and do not alter the method of 
operation of any plant equipment. The change involves the 
establishment of a SQRP for the review of plant procedures, programs 
or changes thereto that do not involve a 10CFR50.59 evaluation.
    Implementing a SQRP will not result in a degradation of the 
current level of procedure review. PORC/SORC will retain the 
responsibility for reviewing any document for which a 10CFR50.59 
evaluation is required. Personnel selected to be SQRs [Station 
Qualified Reviewers] will possess the technical experience and 
expertise to provide a thorough technical review as required by 
plant procedures. These personnel, and the managers authorized to 
approve these procedures, will be designated in writing by the Unit 
Director or the Senior Vice President - Millstone Station. 
Procedures or classes of procedures that can be reviewed per the 
SQRP will be specified in writing by the Unit Director or the Senior 
Vice President - Millstone Station. Procedures will receive an 
appropriate cross-disciplinary review when necessary.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed technical specification changes do not change the 
design or function of any plant structure, system, or component, nor 
do they introduce any new failure modes. As stated above, the 
implementation of a SQRP will not degrade the quality of plant 
procedures.
    There are no modifications to plant structures, systems, or 
components associated with these proposed changes, and the operation 
of plant equipment and systems remain unchanged. Since the changes 
proposed in this license amendment request do not revise existing 
plant structures, systems, or components, do not change the manner 
in which the plant is operated and, do not change the manner in 
which the plant will respond to any design basis accidents, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The changes proposed in this proposed license amendment request 
do not affect the ability of any system to perform its safety-
related function. As described above, these proposed changes are 
administrative in nature. They do not change any plant operating 
parameters or design features and do not reduce the level of 
effectiveness of any existing administrative controls. The proposed 
change will not result in changes to the bases for any technical 
specification. The establishment of the SQRP will continue to 
provide for the adequate review of procedures. In addition, another 
direct benefit of this program is that the amount of material 
presented to PORC/SORC will decrease. The reduction in the amount of 
material presented to PORC/SORC for review will allow the PORC/SORC 
to focus on safety significant issues. Since none of the assumptions 
in the technical specifications bases are affected by the changes 
presented in this license amendment request, the margin of safety 
which exists in the current technical specifications is not reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270
    NRC Project Director: Phillip F. McKee

Northern States Power Company, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota

    Date of amendment request: June 22, 1995
    Description of amendment request: The proposed changes modify the 
facility requirements for thermal-hydraulic instability avoidance and 
protection to address concerns over reactor fuel performance during 
instability events. Changes are proposed to the Technical 
Specifications to utilize the flow biased Average Power Range Monitor 
high neutron flux scram and a power-flow map exclusion region 
consistent with one of the NRC approved BWR Owners' Group solutions. In 
addition, a change to correct an error in the Average Planar Linear 
Heat Generation Rate during single loop operation is also proposed.
    Basis for proposed no significant hazards consideration 
determination: As requied by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards 

[[Page 45182]]
consideration, which is presented below:
    a. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The implementation of BWR Owner's Group long term stability 
solution Option 1-D at Monticello does not modify the assumptions 
contained in the existing accident analysis. The use of an exclusion 
region and the operator actions required to avoid and minimize 
operation inside the region do not increase the possibility of an 
accident. Conditions of operation outside of the exclusion region 
are within the analytical envelope of the existing safety analysis. 
The operator action requirement to exit the exclusion region upon 
entry minimizes the probability of an oscillation occurring. The 
actions to drive control rods and/or to increase recirculation flow 
to exit the region are maneuvers within the envelope of normal plant 
evolutions. The flow based scram has been analyzed and will provide 
automatic fuel protection in the event of a core wide instability. 
Thus, each proposed operating requirement provides defense in depth 
for protection from an instability event while maintaining the 
existing assumptions of the accident analysis. The proposed change 
to the method by which the MAPLHGR [maximum average planar linear 
heat-generation rate] is obtained for single loop operation is 
consistent with the analysis performed for the Average Power Range 
Monitor/Rod Block Monitor Technical Specifications (ARTS) program. 
The analysis performed in support of the ARTS program demonstrated 
that the limits established assure compliance with fuel limits. 
Therefore, this amendment will not cause a significant increase in 
the probability or consequences of an accident previously evaluated 
for the Monticello plant.
    b. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    As stated above, the proposed operating requirements either 
mandate operation within the envelope of existing plant operating 
conditions or force specific operating maneuvers within those 
carried out in normal operation. Since operation of the plant with 
all of the proposed requirements is within the existing operating 
basis, an unanalyzed accident will not be created through 
implementation of the proposed change. Therefore, the proposed 
amendment will not create the possibility of a new or different kind 
of accident.
    c. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    Each of the proposed requirements for the plant thermal-
hydraulic stability provides a means for fuel protection. The 
combination of avoiding possible unstable conditions and the 
automatic flow biased reactor scram provides an in-depth means for 
fuel protection. Therefore, the individual or combination of means 
to avoid and suppress an instability supplements the margin of 
safety. The operating limits established for the single loop 
operation MAPLHGR provide an acceptable margin of safety as 
demonstrated in NEDC-30492, ``Average Power Range Monitor, Rod Block 
Monitor and Technical Specification Improvement (ARTS) Program for 
Monticello Nuclear Generating Plant-April 1984.'' The proposed 
amendment will 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: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: John N. Hannon

Northern States Power Company, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota

    Date of amendment request: July 5, 1995
    Description of amendment request: The proposed amendment, part of 
the Monticello Surveillance Test Interval/Allowed Outage Time (STI/AOT) 
Program, extends the surveillance test intervals and allowable out-of-
service times for selected instrumentation. The proposed changes are 
intended to minimize unnecessary testing and remove excessively 
restrictive out-of-service times that could potentially degrade overall 
plant safety and availability.
    Basis for proposed no significant hazards consideration 
determination: As requied 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 amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The maximum failure frequency change is for the ECCS Actuation 
Instrumentation as identified by General Electric topical report 
NEDC-30936P-A, and Monticello specific report RE-006. These reports 
concluded core damage frequency changed by less than 4% when STIs 
were increased to once per 3 months, AOTs for surveillance were 
increased to 6 hours, and AOTs for repair were increased to 24 
hours. Since this small increase was within the guideline of 
acceptability stated in NEDC-30936P-A, and Monticello only proposes 
to increase the repair AOT to 12 hours rather than 24 hours, this 
amendment will not cause a significant increase in the probability 
or consequences of an accident previously evaluated for the 
Monticello plant (see RE-006).
    The drift analysis determined the associated instrumentation 
would not be adversely effected with the longer calibration 
intervals. Pertinent process parameters including instrument drift 
will still be within acceptance criteria with the longer 
surveillance intervals.
    The recirculation flow meters and flow instrumentation are not 
used in any safety or accident analysis. Therefore, no analysis 
would be changed by increasing the calibration interval to once per 
cycle.
    b. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    These changes only affect the instrument STI and AOT times. No 
changes are being made to the functions of the instrumentation. 
Therefore, the proposed amendment will not create the possibility of 
a new or different kind of accident.
    c. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    These changes will improve the performance of equipment and are 
intended to reduce the potential for equipment failures due to 
unnecessary testing. The safety limits and the limiting safety 
system setpoints will not be affected by these changes. No safety 
margins are affected, therefore, the drift will remain within the 
margins of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment request: August 4, 1995
    Description of amendment request: This proposed amendment would 
revise the Technical Specifications (TS) for the requirements for the 
containment radiation high signal (CRHS) and the safety injection and 
refueling water (SIRW) tank low signal (STLS) contained in TS 2.15, 
Tables 2-3 and 2-4. Specification 3.1, Table 3-2 will also be revised 
to include administrative changes to the CRHS surveillance 

[[Page 45183]]
methods to be consistent with the applicable surveillance functions. 
The Basis for Specification 2.15 is being revised to clarify that the 
number of installed channels for CRHS is two. The term ``SOURCE CHECK'' 
is being deleted from the Definitions section.
    Basis for proposed no significant hazards consideration 
determination: As requied 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 Omaha Public Power District (OPPD) proposes to revise 
Technical Specification (TS) 2.15, Table 2-3 by revising the 
requirement for placing the Safety Injection Refueling Water (SIRW) 
tank low level channel(s) in the tripped condition to placing them 
in the bypassed condition. Due to the derived signal, if a channel 
was in the tripped condition and a single failure occurred, (that 
being one channel of STLS on either A or B circuits), a premature 
SIRW tank low signal (STLS) would be generated. During a design 
basis accident (DBA) with a valid Containment Pressure High Signal 
(CPHS) or Pressurizer Pressure Low Signal (PPLS), this single 
failure would prevent the contents of the SIRW tank from being 
injected into the reactor coolant system. The resulting logic of 
placing the SIRW tank low level channels in BYPASS rather than TRIP 
would not cause a premature switchover of the high pressure safety 
injection pumps to the containment sump and it would not prevent the 
switchover when needed.
    OPPD also proposes to revise TS 2.15, Table 2-4, by reducing the 
number of minimum operable Containment Radiation High Signal (CRHS) 
channels from two to one. This proposed change revises the 
requirements of TS 2.15 to coincide with changes to the TS and 
Offsite Dose Calculation Manual (ODCM) that were implemented by TS 
Amendment 152. The Engineered Safety Feature (ESF) actuation system 
supervisory A and B safeguard initiation channels will not be 
affected by this proposed TS change. The minimum level of engineered 
safeguards performance acceptable for the DBA, (i.e., minimum 
safeguards) will continue to be maintained in accordance with IEEE 
279 - 1971, ``Criteria for Protection Systems for Nuclear Power 
Generating Stations.''
    Included in this change are administrative revisions to TS 3.1, 
Table 3-2, for replacing the current surveillance methods for 
checking and testing the CRHS instrumentation with the defined terms 
``CHANNEL CHECK'' and ``CHANNEL FUNCTIONAL TEST,'' respectively. 
These proposed revisions are administrative in nature and reflect 
TS-defined terminology for the instrumentation surveillance methods 
utilized to ensure that the CRHS instrumentation is operable. A 
channel check requires a qualitative determination of acceptable 
operability by observation of channel behavior during normal plant 
operation. A channel functional test requires the injection of a 
simulated signal into the channel to verify that it is operable, 
including any alarm and/or trip initiating actions. Other proposed 
administrative changes include deleting the term ``SOURCE CHECK'' 
from the TS Definitions section as source check will no longer be 
used in the FCS TS and adding verbiage to the TS 2.15 Basis for 
clarifying that the number of installed channels for CRHS is two.
    Therefore, the proposed change, as described above, would not 
increase 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.
    There will be no physical alterations to the plant 
configuration, changes to setpoint values, or changes to the 
implementation of setpoints or limits as a result of the proposed 
changes to TS 2.15, Tables 2-3 and 2-4. The proposed revisions to TS 
3.1, Table 3-2 are administrative changes to make the TS more 
accurately reflect defined terminology and the methods utilized to 
ensure that the CRHS instrumentation is operable. The proposed TS 
revisions do not require any changes to the present methods of 
verifying CRHS instrumentation operability. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    There are no changes to the equipment or plant operations as a 
result of the changes being made to the number of minimum operable 
CRHS channels. The proposed changes to the STLS will require that 
the inoperable channel be placed in BYPASS rather than TRIP. This 
action would ensure that a single failure would not cause a 
premature safety injection switchover to the containment sump and 
would not prevent switchover when needed. Therefore, this proposed 
change does not reduce a margin of safety.
    The proposed revisions to TS 3.1, Table 3-2 are administrative 
changes to make the TS more accurately reflect defined terminology 
and the methods utilized to ensure that the CRHS instrumentation is 
operable. The proposed TS revisions do not require any changes to 
the present methods of verifying CRHS instrumentation operability. 
The proposed changes to the Definitions and TS 2.15 Basis sections 
are administrative in nature. Therefore, these 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:  W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William H. Bateman

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 22, 1995
    Description of amendment request: The amendments would revise the 
Technical Specifications 3.4.1.4 and 3.9.8.2 by deleting footnotes and 
associated information regarding Service Water header operation and its 
support function for Residual Heat Removal operation. These footnotes 
and associated information had been placed in the Technical 
Specifications because of the concern about Service Water system piping 
integrity in the mid-1980's.
    Basis for proposed no significant hazards consideration 
determination: As requied 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. Do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Even though one service water loop will be out for maintenance, 
both loops of residual heat removal (RHR) will be kept operable, 
consistent with the requirements of STS (NUREG 1431). A minimum of 
two RHR, two component cooling (CC), and two service water (SW) 
pumps, powered from two different vital busses, will be kept 
operable.
    Only one component cooling heat exchanger will be operable since 
only one service water loop is operable. The CC heat exchangers for 
both Units 1 and 2 have a very high reliability. The primary heat 
transfer surfaces of the heat exchangers are made of titanium; no 
material problems have been experienced in ten years of service.
    The remaining active components that, through misoperation, 
could potentially defeat RHR capability are, (1) the motor operated 
valves in RHR or SW that could develop a ``hot short'' and 
subsequently close and (2) the air operated temperature/ flow 
control valves of the CC heat exchangers. Additional actions will be 
taken to effectively eliminate the possibility of these single point 
valves from failing and defeating RHR capability. The motor operator 
breakers will be tagged open during MODES 5 and 6, except for 
flooding the cavity, when the RHR suction valves must be closed. The 
CC Heat Exchanger air operated temperature/flow control valves fail 
open, or as is, on loss of air which is the safe position. Operators 
will monitor critical temperatures; this equipment is accessible if 
any corrective action is required. Thus, with one service water 
header out of service, the intent of the 

[[Page 45184]]
technical specifications as defined in the bases section (to have a 
single failure proof RHR system) is met with the proposed system 
configuration. Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Do not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The catastrophic failure of a moderate energy Class 3 piping 
system is not a credible event, based on the upgraded reliability of 
the system, the redundancy of active components, the elimination of 
single failure points, and on the industry and regulatory positions 
established for this type of system. Since SW is a Class 3 moderate 
energy system, the only postulated passive failure mode is a leakage 
crack. In accordance with Generic Letter (GL) 91-18 and GL 90-05, a 
leak in the SW system, following acceptable evaluation, does not 
constitute a failure that causes the loss of capability to perform 
it's intended safety function. A moderate energy Class 3 piping leak 
does not cause the system to be declared inoperable. Therefore, the 
proposed changes do not create the possibility of a new or different 
type of accident from any previously evaluated.
    3. Do not involve a significant reduction in a margin of safety.
    RHR redundancy is maintained; no credible single failure point 
exists that could cause a nonrecoverable loss of SW. 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: Salem Free Public library, 112 
West Broadway, Salem, New Jersey 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. 
Ginna Nuclear Power Plant, Wayne County, New York

    Date of amendment request: September 15, 1992, as supplemented 
April 20, 1993, April 26, 1995, and July 27, 1995.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) 3.1.1.4, 3.1.1.6, and 4.3.4, and 
add a Basis to address Generic Letter (GL) 90-06. GL 90-06 represents 
the technical resolution of Generic Issue (GI) 70, ``Power Operated 
Relief Valve and Block Valve Reliability,'' and GI 94, ``Additional Low 
Temperature Overpressure Protection for Light Water Reactors.'' The 
resolution of these issues proposes new requirements and TS changes 
that enhance the reliability of power-operated relief valves (PORVs) 
and block valves along with TS changes that will provide additional 
low-temperature overpressure protection (LTOP).
    Basis for proposed no significant hazards consideration 
determination: As requied by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    There is no significant increase in the probability or 
consequences of an accident previously evaluated because the 
accident conditions and assumptions are not significantly affected 
by the proposed change.
    The proposed change to action statement 3.1.1.4a(i) [proposed to 
be renumbered to 3.1.1.6c] to include the removal of power from a 
closed block valve will provide additional assurance to preclude any 
inadvertent opening of the block valve at a time in which the PORV 
may not be operable to assure RCS [reactor coolant system] 
integrity.
    The provision of the generic letter requires, with one or both 
PORV(s) inoperable to initiate shutdown actions if PORV operability 
is not restored within 72 hours or 1 hour respectively. RG&E 
[Rochester Gas and Electric Corporation] does not address these 
shutdown actions, but rather will concentrate on re-establishing 
valve operability. If the block valve(s) and power are not removed 
within 1 hour shutdown provisions must be initiated. [***].
    Proposed action statement 3.1.1.4a(ii) [proposed to be 
renumbered to 3.1.1.6d] includes a provision to place the block 
valves associated PORV(s) switch in manual control due to an 
inoperable block valve(s). This requirement precludes the automatic 
opening for an overpressure event to avoid the potential for a 
stuck-open PORV at a time that the block valve is open and 
inoperable. [***].
    The proposed change of maintaining power to closed block valves 
could potentially increase the probability of an inadvertent opening 
of a block valve. The safety impact is, however, not significant 
since the proposed changes are only applicable if the PORV is 
inoperable due to excessive seat leakage (proposed action 3.1.1.6b). 
[***].
    Proposed action statement 3.1.1.6b establishes reactor coolant 
pressure boundary integrity for a PORV that has excessive seat 
leakage and is therefore considered operable to perform its intended 
safety function. [***].
    Proposed Surveillance Requirement 4.3.4.3 addresses operability 
of the Nitrogen System by demonstration of the PORVs at least once 
per 18 months by operating the PORVs through a complete cycle of 
full travel. [***].
    Based on the above efforts, the proposed amendment does not 
involve a significant increase in the probability or consequences of 
any accident previously evaluated.
    The possibility of a new or different kind of accident from any 
previously evaluated is not created. In matters related to nuclear 
safety, all accidents continue to bound previous analyses. The 
proposed changes do not add or modify any equipment design nor do 
the proposed changes involve any significant operational changes to 
any plant systems.
    The proposed amendment does not involve a significant reduction 
in the margin of safety as defined in the basis for any technical 
specification because the results of the accident analyses which are 
documented in the UFSAR [Updated Final Safety Analysis Report] 
continue to bound operation under the proposed changes so that there 
is no safety margin reduction. [***].
    Therefore, the proposed changes do not involve a significant 
reduction in 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: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610
    Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005
    NRC Project Director: Ledyard B. Marsh

Sacramento Municipal Utility District (SMUD), Docket No. 50-312, 
Rancho Seco Nuclear Station, Sacramento County, California

    Date of amendment request: June 20, 1995 and as amended August 14, 
1995
    Description of amendment request: The proposed amendment (PA-191) 
would permit SMUD to change the Fuel Storage Building load handling 
limits to allow placing the shield plugs on the dry shielded cannisters 
in order to permit transfer of spent fuel assemblies from the spent 
fuel pool (SFP) to the Rancho Seco Independent Spent Fuel Storage 
Installation.
    Basis for proposed no significant hazards consideration 
determination: As requied by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    PA-191 will not create a significant increase in the probability 
or consequences of an accident previously evaluated in the Safety 
Analysis Report (SAR), because dropping the dry shielded canister 
(DSC) top shield plug over a DSC loaded with 24 spent fuel 
assemblies is not considered a credible event. Also, the gantry 
crane is designed such 

[[Page 45185]]
that it can only handle loads over the SFP cask pit area and can not 
move a load over the SFP fuel storage racks.
    PA-191 will not create the possibility of a new or different 
type of accident than previously evaluated in the SAR, because the 
proposed Permanently Defueled Technical Specification heavy load 
handling exceptions do not create a new credible accident scenario. 
Dropping the DSC top shield plug and damaging spent fuel assemblies 
is not considered a credible event.
    PA-191 will not involve a significant reduction in the margin of 
safety, because the proposed heavy load handling exceptions do not 
create a credible accident scenario.
    The NRC staff has reviewed the licensee's analyses of June 20, 1995 
and August 14, 1995. The August 14 submittal enhanced these analyses by 
providing design details regarding the significant safety factors built 
into the crane and other lifting hardware. 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:  Central Library, Government 
Documents 828 I Street, Sacramento, CA 95814
    Attorney for licensee: Dana Appling, Esq. Sacramento Municipal 
Utility District, P. O. Box 15830, Sacramento, CA 95852-1830
    NRC Project Director: Seymour H. Weiss

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

    Date of amendment requests: July 17, 1995
    Description of amendment requests: The licensee proposes to revise 
surveillance requirements associated with Technical Specifications 3/
4.3.1, ``Reactor Protective Instrumentation,'' and 3/4.3.2, 
``Engineered Safety Feature Actuation System Instrumentation.'' The 
surveillance interval is to be increased to 120 days for performance of 
channel functional tests for certain reactor protective system and 
engineered safety feature actuation system instrumentation. The 
proposed change also revises Bases 3/4.3.1, ``Reactor Protective and 
Engineered Safety Features Actuation System Instrumentation,'' to 
reflect the new interval.
    Basis for proposed no significant hazards consideration 
determination: As requied by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change would extend the current sequential Channel 
Functional Test (CFT) surveillance interval for Plant Protective 
System (PPS) instrumentation and Nuclear Instrumentation (NI). This 
change does not involve any changes to plant equipment or operation. 
The proposed change actually maintains or decreases the PPS system 
unavailability. PPS uncertainty and setpoint modifications will 
account for the new surveillance interval. Therefore, the proposed 
change will not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This amendment request does not involve any change to plant 
equipment or operation. The PPS system is used for monitoring and 
mitigation of evaluated accidents. Increasing the availability of 
the PPS system, as proposed in this amendment request, will not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    This amendment does not change the manner in which safety 
limits, limiting safety settings, or limiting conditions for 
operation are determined. This amendment request will increase 
Reactor Protective System and Engineered Safety Features Actuation 
System availability. Therefore, this 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 requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770
    NRC Project Director: William H. Bateman

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

    Date of amendment request: August 7, 1995 (TS 95-12)
    Description of amendment request: The proposed change would correct 
various errors of an editorial nature that have been identified in the 
technical specifications and remove the provisions that have exceeded 
their allowed time interval for implementation or the required 
conditions no longer exist.
    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:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed revisions do not change the TS requirements, plant 
setpoints or functions, or plant operating practices. These changes 
provide clarifications to the existing TSs by correcting editorial 
errors and removing provisions that no longer apply in the 
specifications. The probability or consequences of an accident will 
not be increased by providing the proposed verbiage corrections that 
are editorial and nonintent.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    No plant functions or compliance activities associated with the 
TS requirements have been affected by the proposed editorial 
changes. Therefore, the possibility of a new or different kind of 
accident is not created.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes will not alter TS setpoint values or 
functions. The proposed corrections will enhance the application of 
TS requirements and will support the margin of safety provided by 
the TSs. Therefore, the margin of safety will not be reduced by the 
proposed revisions.
    The NRC 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: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

[[Page 45186]]


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

    Date of amendment request: August 7, 1995 (TS 95-17)
    Description of amendment request: The proposed change would 
relocate the heat flux hot channel factor penalty of two percent from 
Surveillance Requirement 4.2.2.2.e.1 to the Core Operating Limits 
Report and add a reference to the factor to Specification 6.9.1.14.5. 
Also, Specification 6.9.1.14.a.2 would be revised to reference Revision 
1A of Westinghouse Commercial Atomic Power (WCAP) 10216-P-A, 
``Relaxation of Constant Axial Offset Control - FQ Surveillance 
Technical Specifications,'' dated February 1994.
    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:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change involves only the manner in which the 
penalty factors for FQ(Z) would be specified (i.e, a burnup-
dependent factor specified in the Core Operating Limits Report 
[COLR] versus a constant factor specified in the TS). This is simply 
used to account for the fact that FQ(Z) may increase between 
surveillance intervals. These penalty factors are not assumed in any 
of the initiating events for the accident analyses. Therefore, the 
proposed change will have no effect on the probability of any 
accidents previously evaluated. The penalty factors specified in the 
COLR will be calculated using NRC-approved methodology and will 
therefore continue to provide an equivalent level of protection as 
the existing TS requirement. Therefore, the proposed change will not 
affect the consequences of any accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed change does not involve a physical alteration to 
the plant (no new or different kind of equipment will be installed) 
or alter the manner in which the plant would be operated. Thus, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change will continue to ensure that potential 
increases in FQ(Z) over a surveillance interval will be 
properly accounted for. The penalty factors will be calculated using 
NRC-approved methodology. Therefore, the proposed change will not 
involve a reduction in margin of safety.
    The NRC 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: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: August 7, 1995 (TS 95-18)
    Description of amendment request: The proposed change would revise 
the titles of various administrative positions found in Section 6.0 of 
the Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c).
    Operation of Sequoyah Nuclear Plant (SQN) in accordance with the 
proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes only involve the administrative titles of 
management positions in TVA [Tennessee Valley Authority]. Plant 
equipment and operating practices are not affected by the proposed 
administrative changes. Therefore, there is no increase in the 
probability or consequences of an accident.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    Plant features are not impacted by the proposed revision; 
therefore, this revision can not create the possibility of a new or 
different accident.
    3. Involve a significant reduction in a margin of safety.
    Plant setpoints and features that establish and maintain the 
margin of safety for SQN are not involved in the proposed 
administrative TS change. Therefore, the margin of safety is not 
reduced by the proposed change.
    The NRC 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 Romm location: Chattanooga-Hamilton County Library,1101 
Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: August 7, 1995 (TS 95-03)
    Description of amendment request: The proposed change would modify 
Technical Specifications (TS) 3/4.1.3, ``Movable Control Assemblies,'' 
and Bases 3/4.1.3. The proposed change addresses operation with a rod 
urgent failure condition (the control rods are out-of-service because 
of failures external to the individual rod drive mechanisms; i.e., 
programming circuitry, but the control rods remain operable), including 
limited operation with one control or shutdown bank inserted up to 18 
steps below its insertion point. In addition, the surveillance interval 
for rod movement verifications would be increased from 31 days to 92 
days.
    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:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c).
    Operation of Sequoyah Nuclear Plant (SQN) in accordance with the 
proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Allowing for continued operation during diagnosis and repair as 
a result of electronic or electrical malfunctions of the rod control 
system is acceptable, since the design safety function of the 
control rods (reactor trip will remain unaffected during the 
diagnosis and repair period. During the extended 

[[Page 45187]]
troubleshooting and repair period, the requirements for control rod 
alignment, insertion limits (except for a small allowed deviation 
for one bank) and shutdown margin will be maintained. The small 
deviation from the control rod insertion limits allowed for one 
bank, for up to 72 hours, will not adversely impact the current TS 
requirements for normal operation core power distributions. The 
proposed changes do not affect the ability of the control rods to 
perform their intended safety function (rods remain trippable) when 
a safety system setting is reached. No new or unique accident 
precursors be introduced by the proposed changes. Therefore, the 
probability and consequences of accidents related to or dependent on 
control rod operation will remain unaffected.
    The proposed change will result in a small increase in the 
probability, that at any given time, a control or shutdown bank will 
be inserted slightly below (i.e., up to 18 steps) its insertion 
limit. However, by design, the control and shutdown banks will 
continue to meet the safety analysis criterion for steady state and 
American Nuclear Society (ANS) Condition II (moderate frequency) 
transients. The allowed insertion is not a malfunction of equipment 
important to safety in this case; therefore, the probability of such 
a malfunction is not increased. Limiting the allowed time for 
operation with the rod control system out-of-service, but with the 
rods trippable and with a control or shutdown bank below the 
insertion limit, eliminates the need for consideration of this 
condition coincident with any of the low frequency (ANS Condition 
III or IV) design basis accidents.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    There are no new failure mechanisms associated with plant 
operation for an extended period to perform diagnosis and repair on 
the rod control system. Limited periods of operation with immovable, 
but trippable control rods, does not involve any modification to the 
operational limits or physical design of the involved systems. There 
are no new accident precursors created because of the allowed 
diagnosis and repair period.
    3. Involve a significant reduction in a margin of safety.
    The results of the current accident analyses are not impacted by 
the change. In addition, the margin of safety as defined in the 
basis of the TS has not been reduced because current core design 
limits continue to be met for the accidents of concern. Therefore, 
the margin of safety is not impacted.
    The NRC 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: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: June 23, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Surveillance Requirements 
4.1.3.1.2, 4.4.6.2.2.b, 4.4.3.2, 4.6.2.1.d, 4.6.4.2, and Table 4.3-3 in 
accordance with guidance provided in NRC Generic Letter (GL) 93-05, 
``Line Item Technical Specification Improvements to Reduce Surveillance 
Requirements for Testing During Power Operations.'' Additionally, the 
proposed amendment would revise TS 4.1.1.1.1, 4.1.1.2, 3/4.1.3.1 and 
associated Bases to implement portions of the Standard Technical 
Specifications - Westinghouse Plants, NUREG-1431.
    Basis for proposed no significant hazards consideration 
determination: As requied 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 Technical Specification changes do not involve a 
significant hazards consideration per 10 CFR 50.92 because operation 
of Callaway Plant with the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    All changes are in accordance with the recommendations of NRC 
Generic Letter 93-05, Line-Item Technical Specifications 
Improvements to Reduce Surveillance Requirements for Testing During 
Power Operation or NUREG 1431, Standard Technical Specifications - 
Westinghouse Plants. None of the changes affects accident initiators 
and each has been evaluated against Callaway Plant operating 
experience.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The proposed Technical Specification changes do not modify any 
equipment nor create any potential accident initiators. The changes 
per GL 93-05 involve Technical Specification surveillance 
frequencies and do not alter the methodology nor associated 
acceptance criteria. The changes per NUREG-1431 do not create any 
accident initiators and are consistent with Callaway design and 
operation.
    3. Involve a significant reduction in a margin of safety.
    The surveillance frequency changes were recommended via GL 93-05 
and are compatible with Callaway Plant experience. The changes per 
NUREG-1431 do not impact the margin of safety. The Shutdown margin 
requirements and associated safety margins are unaffected by these 
changes.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location:  Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: June 26, 1995
    Description of amendment request: The proposed amendment would 
revise the allowed outage time for component cooling water motor 
operated containment isolation valves, remove the list of containment 
isolation valves, and allow containment penetration check valves to be 
used as isolation devices.
    Basis for proposed no significant hazards consideration 
determination: As requied 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 revision to TS 3/4.6 to remove the listing of 
containment isolation valves, revise the ACTION Statement for the 
CCW MOVs, and credit penetration check valves as isolation devices 
does not involve a significant hazards consideration because 
operation of Callaway Plant with this change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes simplify the TS, meet the regulatory 
requirements for control of containment isolation and are consistent 
with the guidelines of GL 91-08. The information contained in Table 
3.6-1 has not been changed, but only relocated to a different 
controlling document. This is an administrative change which should 
result in improved plant practices and have no impact on plant 
operations. Addition of the footnote to allow up to 12 hours for 
valve testing does not affect the severity of any accident 
previously evaluated. The proposed revision to the TS will not 
adversely impact plant safety since the second barrier of the two 
required is still available to provide isolation between the 
containment atmosphere or the reactor coolant system and the outside 
atmosphere.

[[Page 45188]]

    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    There are no design changes being made that would create a new 
type of accident or malfunction and the method and manner of plant 
operation remain unchanged. Addition of the footnote to allow up to 
12 hours for valve testing does not affect the severity of any 
accident previously evaluated. The additional time provides 
assurance that the inoperable valve is in proper working order prior 
to returning it to OPERABLE condition.
    3. Involve a significant reduction in a margin of safety.
    There are no changes being made to the safety limits or safety 
system settings that would adversely impact plant safety. 
Containment isolation will still be maintained as provided by the 
second isolation valve to ensure that the release of radioactive 
material to the environment will be consistent with the assumptions 
used in the analyses for a LOCA. This will assure that containment 
integrity is maintained.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location:  Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: July 25, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3/4.8.1 and its associated Bases to 
improve overall emergency diesel generator reliability and 
availibility.
    Basis for proposed no significant hazards consideration 
determination: As requied by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    The proposed changes do not involve a significant hazards 
consideration because operation of Callaway Plant with these changes 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    These proposed changes do not involve a change in the 
operational limits or physical design of the emergency power system. 
Emergency diesel generator operability and reliability will continue 
to be assured while minimizing the number of required emergency 
diesel generator starts. Also, emergency diesel generator 
reliability will be enhanced by minimizing severe test conditions 
which can lead to premature failures.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    These proposed changes do not involve a change in the 
operational limits or physical design of the emergency power system. 
The performance capability of the emergency diesel generator will 
not be affected. Emergency diesel generator reliability and 
availability will be improved by the implementation of the proposed 
changes. There is no actual impact on any accident anaiysis.
    3. Involve a significant reduction in the margin of safety.
    These proposed changes do not involve a change in the 
operational limits or physical design of the emergency power system. 
The performance capability of the emergency diesel generator will 
not be affected. Emergency diesel generator reliability and 
availability will be improved by the implementation of the proposed 
changes. No margin of safety is reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: November 29, 1994
    Description of amendment request: The proposed change would revise 
and update the NA-1&2 Environmental Protection Plan (EPP) to reflect 
current obligations to the Commonwealth of Virginia, revise portions of 
the transmission corridor rights-of-way erosion control program for 
clarification and to be consistent with the state regulations, 
eliminate inconsistencies, and delete obsolete material. Specifically, 
references to National Pollutant Discharge Elimination System (NPDES) 
permits are changed to reflect the correct permit title, Virginia 
Pollutant Discharge Elimination System (VPDES). Vegetation and aquatic 
biota studies referred to in the EPP were satisfactorily completed on 
or before June 24, 1986. The discussion of the detailed subject matter 
in these studies is removed because it is extraneous information. A 
reference to 10 CFR 51.5(b)(2) (which does not exist) is corrected to 
10 CFR 51.60(b)(2). The explicit reporting requirements for unusual or 
important environmental events are replaced with the reporting 
requirement which the NRC has required pursuant to 10 CFR 50.72 
(b)(2)(vi). Therefore, the reporting inconsistency (EPP requires report 
to NRC within 24 hours, whereas the 10 CFR 50.72 requires a four hour 
report to the NRC) is resolved. The description of the audit program to 
be utilized for auditing the EPP is replaced by referring to the Audit 
Program established in accordance with 10 CFR 50, Appendix B. Another 
inconsistency is eliminated by revising the two year records retention 
requirement for erosion control inspection field logs to five years. 
This makes the requirement consistent with EPP Section 5.2, Records 
Retention. References to the State Water Control Board are updated to 
that agency's successor, the Department of Environmental Quality. 
Additionally, the licensee's obligation to comply with Virginia 
regulations concerning erosion and sediment control within the 
transmission corridor rights-of-way are recognized to eliminate 
redundancy with previous EPP commitments. The Virginia Soil and Water 
Conservation Board is recognized as the regulatory authority concerning 
erosion within the transmission corridor rights-of-way. The Virginia 
Soil and Water Conservation Board reviews and approves erosion and 
sediment control specifications submitted by utilities on an annual 
basis.
    Basis for proposed no significant hazards consideration 
determination: As requied by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Specifically, operation of the North Anna Power Station in 
accordance with the EPP changes will not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated. The likelihood 
that an accident will occur is neither increased or decreased by the 
proposed changes to the EPP. Sufficient controls are established to 
ensure that environmental controls impacting safety-related 
structures, systems, and components are maintained current and 
accurate. The only potentially credible accident which might be 
affected is the Loss of Offsite Power (if erosion were severe 

[[Page 45189]]
enough to undermine the bases of a transmission tower). Each of the 
three 500 KV transmission lines connected to North Anna Power 
Station can supply sufficient power to the site. This limits the 
effect that one transmission tower has on safe operation of the 
nuclear facility. However, the erosion noted to date has not been 
severe enough to make such an accident credible. Additionally, each 
of the 500 KV transmission lines are inspected for material 
condition annually. Although the intent of this inspection is not 
soil erosion (the annual erosion inspections are currently conducted 
by another group who specializes in land management), evidence of 
severe erosion would be noted and addressed as appropriate. 
Therefore, this EPP change will not impact the function or method of 
operation of plant equipment. Thus, a significant increase in the 
probability of a previously analyzed accident does not result due to 
this change. Nuclear station systems, equipment, or components are 
not affected by the proposed changes. Thus, the consequences of a 
malfunction of equipment important to safety previously evaluated in 
the UFSAR [Updated Final Safety Analysis Report] are not increased 
by this change.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated. The proposed 
changes do not involve changes to the physical plant or operations. 
... the proposed EPP changes do not contribute to accident 
initiation and therefore do not produce a new accident scenario or 
produce a new type of equipment malfunction. Also, this EPP change 
does not alter any existing accident scenarios. The proposed changes 
do not affect nuclear plant equipment or its operation, and thus do 
not create the possibility of a new or different kind of accident. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident.
    (3) Involve a significant reduction in a margin of safety. The 
EPP does not have a formal basis description other than the 
discussion in the FES-OL [Final Environmental Statement-Operating 
License]. The FES-OL discusses the non-radiological impacts of 
facility construction and operation on the environment. The 
discussion indicates that the environment will be managed to a 
stabilized condition during the operations phase, and a program will 
be implemented to maintain the environment in a stabilized 
condition. This intent is not altered by the proposed changes to the 
EPP. The proposed changes do not affect nuclear plant equipment or 
its operation, and thus do not involve any reduction in the margin 
of safety.
    Therefore, use of the proposed EPP would not involve any 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219
    NRC Project Director: David B. Matthews

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

    Date of amendment request: July 26, 1995
    Description of amendment request: The proposed changes would revise 
the Technical Specifications (TS) for the North Anna Power Station, 
Units No. 1 and No. 2 (NA-1&2). Specifically, the proposed changes 
would increase the pressurizer safety valve lift setpoint tolerance as 
well as reduce the pressurizer high pressure reactor trip setpoint and 
allowable value.
    The licensee has prepared a safety evaluation which justifies 
increasing the current TS pressurizer safety valve (PSV) at-power 
(Modes 1-3) lift setpoint tolerance from plus or minus 1% as-found and 
plus or minus 1% as-left to +2%/-3% average as-found with no single 
valve outside plus or minus 3% as-found and plus or minus 1% per valve 
as-left. The as-found value is based on testing, the results of which 
are expressed as an error (i.e., positive or negative percentage 
deviation from the nominal lift setpoint). The errors of the tested 
valves are summed and the result divided by the number of valves 
tested. This result is compared to the acceptable range of +2% to -3%. 
No single valve is allowed to be outside of the plus or minus 3% 
tolerance.
    The safety evaluation also supports an increase to the Hot Shutdown 
(Mode-4) required PSV lift setpoint tolerance from plus or minus 1% as-
found and plus or minus 1% as-left to plus or minus 3% per valve as-
found and plus or minus 1% per valve as-left. These proposed changes 
will provide greater operational flexibility in meeting periodic test 
requirements established by the safety analyses.
    A concurrent reduction in the pressurizer high pressure reactor 
trip setpoint and allowable value of TS Table 2.2-1 are also proposed. 
These changes ensure that the analysis results for the loss of external 
load accident continue to meet the acceptance criteria with the higher 
PSV tolerance.
    The Loss of Load, Locked Rotor, and Rod Withdrawal event analyses 
demonstrate that increasing the at-power PSV lift setpoint tolerance to 
+2%/-3% average as-found with no single valve outside plus or minus 3% 
as-found and plus or minus 1% per valve as-left does not result in a 
transient pressure in excess of the overpressure safety limit. Further, 
the increased setpoint tolerance does not adversely impact the DNBR 
[departure from nucleate boiling ratio] results of any North Anna UFSAR 
[Updated Final Safety Analysis Report] Chapter 15 transient analysis. 
Mode 4 overpressure protection is adequate with one PSV with a 
tolerance of plus or minus 3%.
    Finally, the increased PSV setpoint tolerances and reduction of the 
high pressurizer pressure reactor trip setpoint do not present any 
operational considerations which would significantly impact the 
performance of the plant during normal operation or during postulated 
accident conditions. In summary, each pertinent safety criterion was 
evaluated for the proposed TS changes, and all were found to be 
acceptable.
    Basis for proposed no significant hazards consideration 
determination: As requied by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Specifically, operation of North Anna Power Station in 
accordance with the proposed Technical Specifications changes will 
not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    Affected safety related parameters were analyzed for a change to 
North Anna 1 and 2 Technical Specifications 3.4.2 and 3.4.3 and 
Table 2.2-1 item 10. It was determined that the overpressure safety 
limits would not be exceeded in the most limiting overpressure 
transients (Loss of Load, Locked Rotor, and Rod Withdrawal events) 
with the as-found pressurizer safety valve lift setpoint tolerance 
increased to an average of +2%/-3%, no single valve outside of [plus 
or minus] 3%, and the 25 psi reduction in the Pressurizer High 
Pressure Reactor Trip setpoint. The DNBR results of transients 
impacted by the proposed setpoint tolerance increase meet the 
acceptance criterion after accounting for the impact of the proposed 
changes. The increased setpoint tolerance will not result in an 
inadvertent opening of the pressurizer safety valves. Mode 4 
overpressure protection is adequate with one PSV with a tolerance of 
[plus or minus] 3%.
    2. Create the possibility of a new or different kind of accident 
from any accident previously identified.
    The proposed change to North Anna 1 and 2 Technical 
Specifications 3.4.2 and 3.4.3 and Table 2.2-1 item 10 does not 
involve any changes which would introduce any new or unique 
operational modes or accident precursors. Only the allowable 
tolerance about the existing PSV lift setpoint will be changed, 
along with a reduction in the 

[[Page 45190]]
pressurizer high pressure reactor trip setpoint.
    3. Involve a significant reduction in a margin of safety.
    It was determined that the most limiting overpressure transients 
do not result in maximum pressures in excess of the overpressure 
safety limits. The DNBR results of transients impacted by the 
proposed setpoint tolerance increase meet the acceptance criterion 
after accounting for the impact of the proposed changes. Therefore, 
the margin of safety is unchanged by the proposed increase in the 
safety valve setpoint tolerances.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219
    NRC Project Director: David B. Matthews

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

    Date of amendment request: July 26, 1995
    Description of amendment request: The proposed change would revise 
the Technical Specifications (TS) for the North Anna Power Station, 
Units No. 1 and No. 2 (NA-1&2). Specifically, the change would clarify 
the TS to allow switching of charging and low-head safety injection 
pumps during unit shutdown conditions. The proposed changes would also 
allow additional methods of rendering these same pumps incapable of 
injecting into the reactor coolant system (RCS) when required for low-
temperature conditions. NA-1&2 is equipped with three charging pumps. 
These charging pumps provide inventory control, normal boration to the 
RCS, and flow to the reactor coolant pump seals. They also act as the 
high-head safety injection pumps during accident conditions. During 
certain shutdown conditions, it is necessary to render two of the three 
charging pumps inoperable to maintain the low-temperature overpressure 
protection (LTOP) design bases assumptions. This provides assurance 
that a mass addition pressure transient can be relieved by the 
operation of a single pressurizer power-operated relief valve (PORV). 
Low-temperature overpressure protection for each NA-1&2 unit is 
provided by two pressurizer PORVs.
    During shutdown conditions, periodic surveillance testing of the 
charging pumps is required by the NA-1&2 TS. Also during shutdown 
conditions, it may be desirable to switch from one charging pump to 
another to allow for other activities such as maintenance or testing.
    The current NA-1&2 TS associated with charging pumps during 
shutdown conditions are very restrictive and do not allow sufficient 
latitude for surveillance testing or pump switching. The current NA-1&2 
TS specifically state in the surveillance requirements that the method 
used to render a charging pump inoperable is to place the pump control 
switch in the pull-to-lock position. This requirement would not allow 
for surveillance or post-maintenance testing of the inoperable charging 
pumps since this switch is used to start those pumps.
    Therefore, the licensee proposes to modify NA-1&2 TS to allow more 
than one charging pump to be operable and capable of injecting into the 
RCS for pump switching operations. Additionally, the methods used to 
render charging pumps inoperable will be expanded to allow for post-
maintenance and surveillance testing.
    Basis for proposed no significant hazards consideration 
determination: As requied by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Specifically, operation of North Anna Power Station in 
accordance with the proposed Technical Specifications changes will 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Allowing more than one charging pump to be operable and capable 
of injecting into the RCS during RCS low temperature operation for 
pump switching for post-maintenance and surveillance testing does 
not increase the probability of occurrence or the consequences of 
any previously analyzed accident. Pump switching operations will be 
under the direct administrative control of a licensed operator and 
will only be for a short duration of time. Any situation that could 
result in an excessive RCS mass addition would be immediately 
recognized by the operator and remedial action would be taken to 
prevent challenges to RCS integrity. Using methods such as opening 
the charging pump power supply breaker or closing the charging pump 
discharge valve(s) to render a charging pump inoperable will ensure 
that these pumps will not be capable of injecting water into the 
RCS. These alternate methods are as effective as placing the control 
switches in the pull-to-lock position.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Allowing more than one charging pump to be operable and capable 
of injecting into the RCS during low-temperature operation for pump 
switching for post-maintenance and surveillance testing does not 
involve any physical modifications of the plant nor result in a 
change in a method of operation. Licensed operator control of 
charging pump switching operations will continue to ensure that the 
RCS will not be challenged by excessive mass addition events. Using 
methods other than placing charging pump control switches in the 
pull-to-lock position to render the pump inoperable will still 
ensure that only one pump will be capable of injecting into the RCS 
during low temperature operations. Therefore, a new or different 
type of accident is not made possible.
    3. Involve a significant reduction in a margin of safety.
    Allowing more than one charging pump to be operable and capable 
of injecting into the RCS during RCS low temperature operation for 
pump switching for post-maintenance and surveillance testing does 
not affect any safety limits or limiting safety system settings. The 
alternate methods of rendering pumps inoperable provide the same 
level of assurance that the pump is incapable of flowing into the 
RCS as placing the pump control switch in the pull-to-lock position. 
System operating parameters remain unaffected. The availability of 
equipment required to mitigate or assess the consequence of an 
accident is not reduced. Safety margins are, therefore, not 
decreased.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied.Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219
    NRC Project Director: David B. Matthews

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

    Date of amendment request: July 20, 1995
    Description of amendment request: The proposed amendments would: 1) 
revise three Reactor Protection System/Engineered Safety Features 
Actuation Systems channel trip setpoint limits, 2) 

[[Page 45191]]
add a new setpoint limit for high high steam generator water level, and 
3) incorporate editorial changes to revise the measurement units of one 
setpoint limit and to delete certain references to two-loop operation.
    Basis for proposed no significant hazards consideration 
determination: As requied by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:Specifically, operation of Surry Power Station 
with the proposed change will not:
    (1) Involve a significant increase in either the probability of 
occurrence or consequences of any accident or equipment malfunction 
scenario which is important to safety and which has been previously 
evaluated in the Updated Safety Analysis Report (UFSAR). The effect 
of the proposed change is to ensure that actual plant setpoints 
remain conservative consistent with respect to accident analysis 
assumptions. The proposed change requires safety system actuation 
limits that are more conservative than those currently in Technical 
Specifications. The change does not invalidate currently implemented 
station setpoints or currently applicable accident analysis 
assumptions regarding these setpoints. Consequently, the results and 
conclusions of the current UFSAR accident analyses are not affected 
by these changes. The proposed Technical Specifications change 
revises setpoints used to mitigate accidents and therefore has no 
bearing on the probability of an accident. Further, the change 
ensures that the setpoints used to mitigate an accident bound the 
setpoints used in the accident analyses. Therefore, the probability 
of an accident or consequences of an accident is not adversely 
affected as a result of this change.
    (2) Create the possibility of a new or different type of 
accident than those previously evaluated in the UFSAR. Implementing 
the proposed Technical Specifications setpoint limits cannot create 
the possibility of an accident of a different type than was 
previously evaluated in the UFSAR. Since actual plant setpoints are 
not being affected, new accident precursors will not be introduced. 
Furthermore, spurious challenges to safety systems are also not 
expected to increase in frequency as a result of these changes since 
actual setpoints installed in the plant are not being changed. 
Consequently, no new accident precursors are created as a result of 
the new Technical Specifications setpoint limits.
    (3) Involve a significant reduction in a margin of safety. Since 
the results of the existing UFSAR accident analyses remain bounding, 
safety margins are not impacted. The proposed Technical 
Specifications setpoint limits ensure plant setpoints remain 
conservative and consistent with design base accident analysis 
assumptions including appropriate instrument channel uncertainties 
due to harsh environmental conditions. Therefore, the margin of 
safety as defined in the Technical Specifications bases is 
unaffected.
    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: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219
    NRC Project Director: David B. Matthews

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
Creek Generating Station, Coffey County, Kansas

    Date of amendment request: July 25, 1995
    Description of amendment request: This license amendment request 
proposes to revise Technical Specification 4.0.5a and Bases Section 3/
4.4.10 to delete the clause ``(g), except where specific written relief 
has been granted by the Commission pursuant to 10 CFR Part 50, Section 
50.55a(g)(6)(i).'' This proposed change is consistent with NUREG-1482, 
``Guidelines for Inservice Testing and Nuclear Power Plants.''
    Basis for proposed no significant hazards consideration 
determination: As requied by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This proposed change would remove the wording ''...(g), except 
where specific written relief has been granted by the Commission 
pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).'' The Inservice 
Inspection and Testing Programs are described in the technical 
specifications pursuant to 10 CFR 50.55a. In addition, the proposed 
change, in accordance with NUREG-1431 and NUREG-1482, would provide 
relief to the ASME Code requirement in the interim between the time 
of submittal of a relief request until the NRC has issued a safety 
evaluation and granted the relief. The change being proposed is 
administrative in nature and does not affect assumptions contained 
in plant safety analyses, the physical design and/or operation of 
the plant, nor does it affect any technical specification that 
preserves safety analysis assumptions. Any relief from the approved 
ASME Section XI Code requirements will require a 10 CFR 50.59 
evaluation to ensure no technical specification changes or 
unreviewed safety questions exist. Therefore, operation of the 
facility in accordance with the proposed change would not affect the 
probability or consequences of an accident previously analyzed.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This proposed change would remove the wording ''...(g), except 
where specific written relief has been granted by the Commission 
pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).'' The Inservice 
Inspection and Testing Programs are described in the technical 
specifications pursuant to 10 CFR 50.55a. In addition, the proposed 
change, in accordance with NUREG-1431 and NUREG-1482, would provide 
relief to the ASME Code requirement in the interim between the time 
of submittal of a relief request until the NRC had issued a safety 
evaluation and granted the relief. The change being proposed is 
administrative in nature and will not change the physical plant or 
the modes of operation defined in the facility license. The change 
does not involve the addition or modification of equipment nor does 
it alter the design or operation of plant systems. Any relief from 
the approved ASME Section XI Code requirements will require a 10 CFR 
50.59 evaluation to ensure no technical specification changes or 
unreviewed safety questions exist. Therefore, operation of the 
facility in accordance with the proposed change would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    This proposed change would remove the wording ''...(g), except 
where specific written relief has been granted by the Commission 
pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).'' The Inservice 
Inspection and Testing Programs are described in the technical 
specifications pursuant to 10 CFR 50.55a. In addition, the proposed 
change, in accordance with NUREG-1431 and NUREG-1482, would provide 
relief to the ASME Code requirement in the interim between the time 
of submittal of a relief request until the NRC has issued a safety 
evaluation and granted the relief. The change being proposed is 
administrative in nature and will not alter the bases for assurance 
that safety-related activities are performed correctly or the basis 
for any technical specification that is related to the establishment 
or maintenance of a safety margin. Any relief from the approved ASME 
Section XI Code requirements will require a 10 CFR 50.59 evaluation 
to ensure no technical specification changes or unreviewed safety 
questions exist. Therefore, operation of the facility in accordance 
with the proposed change 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 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 locations: Emporia State University, 

[[Page 45192]]
    William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

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.

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

    Date of amendment request: August 11, 1995
    Description of amendment request: The proposed amendment would 
remove Technical Specification Section 3.2, ``Makeup and Purification 
and Chemical Addition Systems,'' and its bases. The pertinent 
requirements and bases applicable to these systems are being 
incorporated in the TMI-1 Updated Final Safety Analysis Report (UFSAR).
    Date of publication of individual notice in Federal Register: 
August 18, 1995 (60 FR 43172)
    Expiration date of individual notice: September 18, 1995
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105

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: April 6, 1995, and superseded 
on August 7, 1995
    Description of amendments request: Amend the Sequoyah Nuclear 
Plant, Units 1 and 2 Technical Specification (TS) to revise the 
numerical values for the overtemperature and overpower delta-
temperature equation constants in TS Table 2.2-1, Reactor Trip System 
Instrumentation Trip Setpoints.
    Date of publication of individual notice in the Federal Register: 
August 15, 1995 (60 FR 42187)
    Expiration date of individual notice: September 14, 1995
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Notice Of Issuance Of Amendments To Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

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

    Date of application for amendments: May 2, 1995
    Brief description of amendments: The amendments remove from the 
technical specifications (TS) plant elevations for the minimum water 
volume required in the spent fuel pool and relocate them to site 
procedures. The TS amendment also includes two changes to correct 
administrative errors in the TS.
    Date of issuance: August 7, 1995
    Effective date: August 7, 1995
    Amendment Nos.: Unit 1 - Amendment No. 97 ; Unit 2 - Amendment No. 
85; Unit 3 - Amendment No. 68
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35060) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 7, 1995.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 Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick 
County, North Carolina

    Date of application for amendments: January 25, 1993, as 
supplemented on December 28, 1993, September 13, 1994, January 13, 
1995, and May 25, 1995. The supplemental submittals did not expand the 
scope of the original Federal Register notice or change the no 
significant hazards determination.
    Brief description of amendments: The amendments allow unit entry 
into Operational Condition 1 (Power Operation) from Operational 
Condition 2 (Startup) with up to eight inoperable control rods, 
provided those control rods are not inoperable due to being immovable 
or untrippable.
    Date of issuance: August 11, 1992
    Effectove date: August 11, 1992 

[[Page 45193]]

    Amendment Nos.: 178 and 209
    Facility Operating License Nos. DPR-71 and DPR-62.
    Date of initial notice in Federal Register: July 7, 1993 (58 FR 
36428) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 11, 1995.Significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297

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: April 5, 1995, as supplemented 
July 31, 1995
    Brief description of amendment: The amendment revises various 
portions of TS 3/4.9, Refueling Operations, to be consistent with 
NUREG-1431, ``Standard Technical Specifications, Westinghouse Plants,'' 
and allows the relocation of applicable sections from the TS that do 
not meet the Commission screening criteria for retention.
    Date of issuance: August 9, 1995
    Effectove date: August 9, 1995
    Amendment No.: 61
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24906) The July 31, 1995 letter provided clarifying 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 August 9, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

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

    Date of application for amendments: January 13, 1995
    Brief description of amendments: The amendments revise the pressure 
alarm setpoint allowable values for the emergency core cooling system 
(ECCS) and reactor core isolation cooling (RCIC) system ``keep filled'' 
pressure instrumentation channels. The purpose of the change is to 
lower the setpoint allowable values for these parameters to more 
realistic values based upon calculations performed by the licensee 
reflecting design changes and system performance. Also, the term 
``setpoint'' is being changed to ``setpoint allowable value'' to 
clarify the use of the values. Additionally, two administrative/
editorial changes are included to delete technical specification 
footnotes which are no longer applicable.
    Date of issuance: August 15, 1995
    Effectove date: Immediately, to be implemented within 90 days.
    Amendment Nos.: 105 and 91
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11128) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 15, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348

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 application of amendments: February 24, 1994, as 
supplemented by letters dated April 19, May 25, August 25, 1994, 
January 4, January 27, February 22, March 15, April 19, and May 31, 
1995
    Brief description of amendments: The amendments provide 
surveillance requirements for a planned modification to the Keowee 
emergency power generators' underground power path breaker closing 
logic.

    Date of issuance:  August 15, 1995
    Effectove date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 210, 210, and 207
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14887) The April 19, May 25, August 25, 1994, January 4, January 27, 
February 22, March 15, April 19, and May 31, 1995, letters provided 
clarifying information that did not change the scope of the February 
24, 1994, application and initial no proposed significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated August 15, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691

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

    Date of application for amendments: March 30, 1995, as supplemented 
May 5, 1995 and June 19, 1995
    Brief description of amendments: These amendments relate to 
separation of the 24-hour emergency diesel generator test and hot 
restart test from the loss of offsite power test.
    Date of issuance: August 8, 1995
    Effectove date: August 8, 1995
    Amendment Nos.: 175 and 169Facility Operating Licenses Nos. DPR-31 
and DPR-41: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 23, 1995 (60 FR 
27339), and July 5, 1995 (60 FR 35072) The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
August 8, 1995.No significant hazards consideration comments received: 
No
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199

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

    Date of application for amendment: April 15, 1995, as supplemented 
by letters on May 20, 1994, and March 8, 1995
    Brief description of amendment: The amendment revises Technical 
Specification Section 6.5.3, ``AUDITS,'' by removing the specified 
frequency for internal audits. These frequency specifications will now 
be located in Appendix E of the GPU Nuclear Operational Quality 
Assurance Plan (1000-PLN-7200.01). A minor editorial change has been 
incorporated into TS 6.5.1.14 correcting a reference in response to a 
finding in the Operational Safety Team Inspection (OSTI) report of 
December 23, 1993.
    Date of issuance: August 7, 1995
    Effectove date: As of the date of issuance to be implemented within 
30 days
    Amendment No.: 181
    Facility Operating License No. DPR-16. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27056) 

[[Page 45194]]
The letters of May 20, 1994, and March 8, 1995, provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of this amendment is contained in a Safety Evaluation dated 
August 7, 1995.No significant hazards consideration comments received: 
No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753

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

    Date of application for amendment: April 19, 1994, supplemented 
March 8, 1995
    Brief description of amendment: The amendment revises the TMI-1 
Technical Specification (TS) Section 6.5.3 to remove the specified 
frequency of various licensee-conducted audits, including those related 
to quality assurance, fire protection, security, emergency 
preparedness, and offsite dose calculations. The frequencies for 
conduct of these audits will now be specified in the licensee's 
Operational Quality Assurance Plan, which requires NRC approval for 
significant changes. The Commission has determined that these audit 
frequencies need not be in the TS to assure public health and safety.
    Date of issuance: August 14, 1995
    Effectove date: August 14, 1995
    Amendment No.: 195
    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 March 8, 1995, submittal provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of 
this amendment is contained in a Safety Evaluation dated August 14, 
1995.No significant hazards consideration comments received: No.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: December 13, 1994, as 
supplemented April 3, 1995
    Brief description of amendment: The amendment revises Table 
3.6.1.2-1 to allow a maximum leakage of 24.0 scfh for each of the 8 
main steam isolation valves instead of the current 6.0 scfh.
    Date of issuance: August 10, 1995
    Effectove date: As of the date of issuance to be implemented within 
60 days
    Amendment No.: 67
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: January 18, 1995 (60 FR 
3675) The April 3, 1995, letter provided clarifying information that 
did not change the initial no proposed significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated August 10, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126

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

    Date of application for amendments: March 15, 1995 (published in 
Federal Register as March 15, 1994) as supplemented by letter dated 
August 5, 1995
    Brief description of amendments: These amendments modify the 
Susquehanna Steam Electric Station Technical Specification Table 3.6.3-
1, Primary Containment Isolation Valves, concerning the scope of Type C 
testing on specified emergency core cooling system and reactor core 
isolation cooling containment isolation valves. Specifically, the 
subject valves on systems which terminate below the minimum water level 
of the suppression pool will no longer require Type C testing but will 
instead be tested using requirements of the American Society of 
Mechanical Engineers' Section XI Code.
    Date of issuance: August 15, 1995
    Effectove date: August 15, 1995
    Amendment Nos.: 149 and 119
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.The supplemental letter did not 
change the proposed no significant hazards consideration determination 
nor the Federal Register notice.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20521) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 15, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701

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

    Date of application for amendments: March 31, 1995, as supplemented 
by letter dated June 22, 1995
    Brief description of amendments: These amendments delete from the 
Technical Specifications of each unit, the operational condition 
restriction in Surveillance Requirement 4.8.1.1.2.d.7, which requires 
that 24-hour emergency diesel generator testing be performed with at 
least one unit in operational condition 4 or 5 (cold shutdown or 
refueling).
    Date of issuance:  August 15, 1995
    Effectove date: Units 1 and 2, effective as of the date of issuance 
and shall be implemented within 60 days
    Amendment Nos.:  150 and 120
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20523) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 15, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701

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

    Date of application for amendments: November 21, 1994, as 
supplemented by letters dated February 21, 1995, March 28, 1995, April 
10, 1995, May 24, 1995, and June 23, 1995
    Brief description of amendments: These amendments change the 
Technical Specifications for the two units by deleting reference to the 
main steamline isolation valve (MSIV) leakage control system and its 
associated primary containment isolation valves, and increase the 
allowable leakage rate for any MSIV and the total maximum 

[[Page 45195]]
pathway leakage for all four main steam lines.
    Date of issuance: August 15, 1995
    Effectove date: Units 1 and 2 as of date of issuance and shall be 
implemented within 30 days
    Amendment Nos.: 151 and 121
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 4, 1995 (60 FR 
503) The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 15, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701

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

    Date of application for amendment: March 2, 1995
    Brief description of amendment: The amendment extends the 
surveillance test intervals for the snubber systems to support 24-month 
operating cycles. Surveillance test interval extensions are denoted as 
being performed ``every 24 months'' or ``at least once per 24 months'' 
consistent with the guidance provided in Generic Letter (GL) 91-04, 
``Changes in Technical Specification Surveillance Intervals to 
Accommodate 24-Month Fuel Cycle,'' dated April 2, 1991. The NRC staff 
has determined that the proposed Technical Specification changes are in 
accordance with GL 91-04, and are, therefore, acceptable.
    Date of issuance: August 8, 1995
    Effectove date: As of the date of issuance to be implemented within 
30 days
    Amendment No.: 226
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24916) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 8, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126

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

    Date of application for amendments: February 5, 1993, supplemented 
April 13, June 11 and November 17, 1993
    Brief description of amendments: The amendment eliminates the 
Steam/Feedwater Flow Mismatch and Low Steam Generator Water Level 
Reactor Trip due to the installation of the digital feedwater control 
system incorporating a median signal selector.
    Date of issuance: August 7, 1995
    Effectove date: Unit 1, as of the date of issuance, to be 
implemented by the startup following the twelfth refueling outage, Unit 
2, as of the date of issuance, to be implemented by the startup 
following the current outage
    Amendment Nos.: 173 and 154
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1993 (58 FR 
25864) The April 13, June 11, and November 17, 1993 submittals provided 
clarifying 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 August 7, 1995.No significant hazards consideration comments 
received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

South Carolina Electric & Gas Company, South Carolina Public 
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear 
Station, Unit No. 1, Fairfield County, South Carolina

    Date of amendment request: March 11, 1994
    Description of amendment request: The amendment decreases the 
allowable time for operation with one inoperable residual heat removal 
(RHR) relief valve from 7 days to 72 hours. This amendment request has 
been submitted in response to Generic Issue 94 as discussed in Generic 
Letter 90-06.
    Date of issuance: August 11, 1995
    Effectove date: August 11, 1995
    Amendment No.: 125
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 22, 1994 (59 FR 
32236) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 11, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Fairfield County Library, 300 
Washington Street, Winnsboro, South Carolina 29180

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

    Date of amendment request: February 14, 1994 (TXX-94045), as 
supplemented by letter dated May 23, 1995 (TXX-95147)
    Brief description of amendments: The amendments incorporated 
appropriate references to and provisions of the new 10 CFR Part 20 
regulations. These changes revised a definition and aspects of 
radiological effluent technical specifications, clarified the 
administrative specification for reporting individual annual exposures 
greater than 100 mrem by work/job function, and revised the 
administrative specifications for providing alternative measures for 
control of access to high radiation areas and designating record 
retention for radioactive shipments.
    Date of issuance: August 11, 1995
    Effectove date: August 11, 1995
    Amendment Nos.: Unit 1 - Amendment No. 42; Unit 2 - Amendment No. 
28
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1994 (59 FR 
22016) The additional information contained in the supplemental letter 
dated May 23, 1995, was clarifying in nature and thus, within the scope 
of the initial notice and did not affect the staff's proposed no 
significant hazards consideration determinations. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated August 11, 1995.No significant hazards consideration 
comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, 
Vermont Yankee Nuclear Power Station, Vernon, Vermont

    Date of application for amendment: March 31, 1994, as supplemented 
by letters dated September 9, 1994, and June 22, 1995
    Brief description of amendment: The amendment modifies the 
requirements for avoidance and protection from thermal hydraulic 
instabilities to be 

[[Page 45196]]
consistent with the Boiling Water Reactor (BWR) Owners Group long-term 
solution Option I-D described in the Licensing Topical Report, ``BWR 
Owners Group Long-Term Stability Solutions Licensing Methodology, NEDO-
31960 June 1991'' and NEDO-31960, Supplement 1, Dated March 1992. NEDO-
31960 and NEDO-31960, Supplement 1, were accepted by the NRC staff in a 
letter to L.A. England (BWR Owners Group) dated July 12, 1993.

    Date of issuance: August 9, 1995

    Effectove date: As of the date of issuance to be implemented within 
30 days

    Amendment No.: 146

    Facility Operating License No. DPR-28. Amendment revised the 
Technical Specifications.

    Date of initial notice in Federal Register: January 4, 1995 (60 FR 
507) The September 9, 1994, and June 22, 1995, submittals provided 
clarifying 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 August 9, 1995. No significant hazards consideration comments 
received: No

    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301

    Dated at Rockville, Maryland, this 23rd day of August.

    For The Nuclear Regulatory Commission

Elinor G. Adensam,
Acting Director, Division of Reactor Projects III/IV, Office of Nuclear 
Reactor Regulation

[Doc. 95-21389 Filed 8-29-95; 8:45 am]

BILLING CODE 7590-01-F