[Federal Register Volume 60, Number 138 (Wednesday, July 19, 1995)]
[Notices]
[Pages 37084-37109]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-17565]



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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 June 23, 1995, through July 7, 1995. The 
last biweekly notice was published on July 5, 1995 (60 FR 35058).

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. 

[[Page 37085]]

    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 August 18, 1995, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no 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 

[[Page 37086]]
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.

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 request: September 10, 1993, as 
supplemented June 16, 1995.
    Description of amendment request: 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 operations 
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 September 10, 1993, and June 16, 1995, applications proposed to 
upgrade only Section 3/4.8 (Plant Systems) of the Dresden and Quad 
Cities TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    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 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 3/4.8 are based on STS guidelines or later 
operating BWR plant's NRC accepted changes. Any deviations from STS 
requirements do not significantly increase the probability or 
consequences of any previously evaluated accidents for Dresden or 
Quad Cities Stations. The proposed amendment is consistent with the 
current safety analyses and has 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 Plant Systems are not 
assumed in any safety analysis to initiate any accident sequence for 
Dresden or 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. The additional surveillance requirements 
improve the reliability and availability of all affected systems 
and, therefore, reduce the consequences of any accident previously 
evaluated, as the probability of the systems outlined within Section 
3/4.8 of the proposed Technical Specifications, performing their 
intended function is increased by the additional surveillances.
    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. Some of the changes may involve revision in the 
operation of the station; however, these provide additional 
restrictions which are in accordance with the current safety 
analysis, 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.
    The proposed amendment for Dresden and Quad Cities Station's 
Technical Specification Section 3/4.8 is based on STS guidelines or 
later operating BWR plants' NRC accepted changes. The proposed 
amendment has been reviewed for acceptability at the Dresden or Quad 
Cities Nuclear Power Stations considering similarity of system or 
component design versus the STS or later operating BWRs. Any 
deviations from STS requirements do not create the possibility of a 
new or different kind of accident previously evaluated for Dresden 
or Quad Cities Stations.
    No new modes of operation are introduced by the proposed 
changes. 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. 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 Plant 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 
Plant 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.
    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 

[[Page 37087]]
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 with 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 3/4.8 
implements present requirements, or the intent of present 
requirements in accordance with the guidelines set forth in the STS. 
Any deviations from STS requirements do not significantly reduce the 
margin of safety for Dresden or 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 or 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 or Quad Cities and maintain necessary 
levels of system or component reliability, 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 Plant 
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 request involves no significant hazards consideration.
    Local Public Document Room location: For Dresden, Morris Area 
Public Library District, 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 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: September 17, 1993, as 
supplemented June 30, 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 September 17, 1993, and June 30, 1995, applications proposed to 
upgrade only Section 3/4.6 (Primary System Boundary) of the Dresden and 
Quad Cities TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    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 represent minor curtailments of the 
current requirements which are based on generic guidance or 
previously approved provisions for other stations. The proposed 
amendments for Dresden and Quad Cities Station's Technical 
Specification Section 3/4.6 are based on STS guidelines or later 
operating BWR plant's NRC accepted changes. Any deviations from STS 
requirements do not significantly increase the probability or 
consequences of any previously evaluated accidents for Dresden or 
Quad Cities Stations. The proposed amendment is consistent with the 
current safety analyses and has 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 Primary System Boundary 
are not assumed in any safety analysis to initiate any accident 
sequence for Dresden or 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. The additional surveillance 
requirements improve the reliability and availability of all 
affected systems and therefore, reduce the consequences of any 
accident previously evaluated as the probability of the systems 
outlined within Section 3/4.6 of the proposed Technical 
Specifications, performing its intended function is increased by the 
additional surveillances.
    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 

[[Page 37088]]
for other stations. These changes do not involve revisions to the 
design of the station. Some of the changes may involve revision in 
the operation of the station; however, these provide additional 
restrictions which are in accordance with the current safety 
analysis, 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.
    The proposed amendment for Dresden and Quad Cities Station's 
Technical Specification Section 3/4.6 is based on STS guidelines or 
later operating BWR plants' NRC accepted changes. 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 STS or later operating BWRs. Any 
deviations from STS requirements do not create the possibility of a 
new or different kind of accident previously evaluated for Dresden 
or Quad Cities Stations. No new modes of operation are introduced by 
the proposed changes. 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. 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 Primary System Boundary 
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 Primary System Boundary 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.
    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 with 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 3/4.6 
implements present requirements, or the intent of present 
requirements in accordance with the guidelines set forth in the STS. 
Any deviations from STS requirements do not significantly reduce the 
margin of safety for Dresden or 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 reliability, 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 Primary 
System Boundary 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

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

    Date of amendment request: March 17, 1995
    Description of amendment request: The proposed amendment transfers 
requirements for a cycle specific core operating limit from the 
Technical Specifications to the Core Operating Limits Report. 
Additionally, a reference to a statistical methodology for determining 
uncertainties is being changed to reference a methodology that was 
recently approved by the NRC.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability of an Accident Previously Evaluated.
    The removal of the cycle-dependent value for the departure from 
nucleate boiling ratio (DNBR) reduction from technical 
specifications and placing it into the Core Operating Limits Report 
(COLR) has no impact on plant operation or accident analyses. The 
proposed change is considered to be administrative in nature. 
Technical specifications will continue to require operation within 
the core operational limits for each cycle reload calculated by the 
approved reload design methodologies. The appropriate actions 
required if limits are violated will remain in the technical 
specifications. The reload report presents the results of a cycle-
specific evaluation of accidents and transients addressed in the 
ANO-2 Safety Analysis Report (SAR). The cycle-specific evaluation 
demonstrates that changes in the fuel cycle design and the 
corresponding COLR do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The Modified Statistical Combination of Uncertainties (MSCU) 
methodology statistically combines uncertainties to at least a 95/95 
probability/confidence level. The Proposed change to reference the 
MSCU is administrative in nature. The currently referenced 
methodology is being replaced with a more recently approved 
methodology which has been determined to be applicable to ANO-2. The 
new methodology has been independently reviewed and approved by the 
NRC. This change does not impact either the manner in which the 
operating margin to limits on linear heat rate and DNBR is 
maintained or the manner in which the CPCs respond to transients and 
provide trips. Therefore, this change does not adversely impact 
transient analysis assumptions or results. In addition, the physical 
design or operation of the plant is not impacted by this change. The 
safety analyses will continue to be performed utilizing NRC-approved 
methodologies and specific reload changes will be evaluated per 
10CFR50.59.
    Therefore, these changes do not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The proposed change to relocate the cycle-specific value for the 
DNBR reduction from technical specifications to the COLR is 
administrative in nature. No change to the design, configuration, or 
method of operation of the plant is made by this change. This 
parameter will be determined using NRC-approved methods. Technical 
specifications will continue to require operation within the 
required core operating limits and appropriate actions will be taken 
if the limits are exceeded. The relocation of a cycle-specific 
parameter does not create the possibility of a new or different of 
accident from any accident previously evaluated.
    The proposed change to reference the NRC-approved MSCU 
methodology is administrative in nature. The currently 

[[Page 37089]]
referenced methodology is being replaced with a more recently approved 
methodology which has been determined to be applicable to ANO-2. No 
physical alterations of plant configuration, changes to plant 
operating procedures, or operating parameters are proposed. The 
safety analyses are still performing utilizing NRC-approved 
methodologies and specific reload changes will be evaluated per 
10CFR50.59. No new equipment is being introduced, and no equipment 
is being operated in a manner inconsistent with its design.
    Therefore, these changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    Existing technical specification operability and surveillance 
requirements are not reduced by the proposed change to relocate the 
cycle-specific value for DNBR reduction to the COLR. The development 
of limits for a particular cycle will continue to conform to methods 
described in NRC-approved documentation. Technical specifications 
will still require that the core be operated within these limits and 
specify appropriate actions to be taken if the limits are violated. 
The cycle-specific COLR limits for future reloads will be developed 
based on NRC-approved methodologies. Each reload undergoes a 
10CFR50.59 safety review to assure that operating of the unit within 
the cycle-specific limits will not involve a significant reduction 
in a margin of safety.
    The proposed change to reference the MSCU methodology is 
administrative in nature. The currently referenced methodology is 
being replaced with a more recently approved methodology which has 
been determined to be applicable to ANO-2. The resultant overall 
uncertainty factors using the MSCU methodology are determined and 
applied to at least the same 95/95 probability/confidence level as 
the overall uncertainty factors using the current methodology. NRC 
review and approval of the methodologies used to perform the cycle-
specific reload analyses is not affected by this change. The safety 
analyses are still performed utilizing NRC-approved methodologies 
and specific reload changes will be evaluated per 10CFR50.59.
    Therefore, these changes do not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: March 17, 1995
    Description of amendment request: The proposed amendment deletes 
requirements associated with surveillances to verify position stops for 
High Pressure Safety Injection Emergency Core Cooling System throttle 
valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The HPSI system is not an initiator of a previously evaluated 
accident; therefore, the probability of a previously evaluated 
accident will not be increased by the proposed change. Accidents 
which require the use of HPSI will not have any increased 
consequences since the new injection/isolation valve arrangement is 
at least as reliable as the previous valve arrangement. No part of 
the proposed change has any adverse effect upon the HPSI system 
response or function. The new manual valves will perform the 
throttling function previously performed by the HPSI isolation MOVs 
without reliance upon any electrical equipment (MOV limit switches). 
The proposed change does not affect routing of HPSI piping or affect 
total flow characteristics of the system. The proposed change to 
remove the requirement to verify the correct settings of position 
stops for the HPSI throttle valves is consistent with NUREG-1432, 
restructured ``Standard Technical Specifications - Combustion 
Engineering Plants,'' since the manual throttle valves fixed into 
position serve the function of, and are equivalent to, flow limiting 
orifices.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The proposed change does not change the function or mode of 
operation of the HPSI system. The failure of the new MOVs to 
function will have no different effect than failure of the 
previously installed MOVs and such failure is enveloped by 
assumptions in the existing safety analysis, i.e., redundant trains 
will still be able to function. The new manual valves are less 
likely to fail in operation since they are fixed into position by 
tack-welded locking devices and therefore perform their function 
passively. Inadvertent manipulation of the manual valves will be 
prevented by the locking arrangement. There are no new functions or 
modes being accomplished by the MOVs. The throttling function to be 
performed by the manual valves will be more reliably performed by 
passive components than by active electrical circuits. The change 
eliminates uncertainty in throttle valve position as a result of 
limit switch tolerances and repeatability which form the basis for 
the current surveillance requirement for periodic verification.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.No margin of safety will be reduced or affected by 
the proposed deletion of the surveillance requirement. The new 
manual valves will be throttled to produce a system flow balance 
equivalent to the current one, and the balance will continue to be 
confirmed by surveillance testing in accordance with TS 
requirements.
    Therefore, this change 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: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: March 17, 1995
    Description of amendment request: The proposed amendment revises 
requirements associated with the frequency of containment post-entry 
visual inspections.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The proposed change to the Arkansas Nuclear One-Unit 2 (ANO-2) 
technical specifications (TS) does not involve any system or 
component or condition evaluated as an accident initiator; therefore 
there is no increase in the probability of an accident previously 
evaluated.
    The purpose of this change is to reduce the required number of 
containment inspections following entries at operational modes above 
cold shutdown. This reduction in the number of inspections will 
reduce personnel 

[[Page 37090]]
exposure to radiation and potential heat stress. These inspections are 
to verify that no debris that might be transported to the 
containment sump is left behind at the conclusion of the entry. 
Typically, containment entries above cold shutdown are for specific 
purposes and involve a limited area of containment. The expectation 
for job performance at ANO-2 is that a job site is left cleaner than 
found. The inspection serves as a verification that any materials 
taken into the containment building which might foul the sump 
screens have been removed or have been properly anchored. Performing 
this inspection on a daily frequency will not result in changing the 
work practices at ANO-2, therefore the amount of debris generated or 
left in containment should not increase. The daily inspection will 
be sufficient verification that conditions in containment are not 
degrading; therefore, there will be no significant increase in the 
consequences of an accident previously evaluated.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    Because the proposed amendment will not change the design, 
configuration, or method of operation of the plant, this change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    There will be no adverse effects on margins of safety since 
materials that are considered acceptable to remain in containment 
has not changed. By reducing the number of inspections, no mechanism 
has been created that will generate more debris in containment nor 
have work practices been altered to allow less stringent controls 
over what is taken in or left in containment. Therefore, this change 
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: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: April 4, 1995
    Description of amendment request: The proposed amendment deletes 
requirements associated with part length control element assemblies. 
During the upcoming refueling outage all part length control assemblies 
will be removed from the reactor.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The proposed changes maintain conservative restrictions on the 
operation of those control element assemblies (CEAs) formerly 
specified as part length CEAs (PLCEAs) and are considered to be 
administrative in nature. The Arkansas Nuclear One - Unit 2 (ANO-2) 
Safety Analysis Report (SAR) Chapter 15 accident analyses identify 
four families of analyses associated with the CEAs. Each of these 
analyses is evaluated in the development of the Reload Report for 
each fuel cycle, and the appropriate limitations to insure 
acceptable analysis results are incorporated in the Core Operating 
Limits Report (COLR) for the fuel cycle. The modification replacing 
the PLCEAs with full length CEAs will be evaluated under the 
Arkansas Nuclear One (ANO) 10CFR50.59 process prior to 
implementation. The Reload Report and changes to the COLR are also 
evaluated under the ANO 10CFR50.59 process prior to incorporating 
the identified changes. Movement of the PLCEAs during power 
operation has typically resulted in more dropped CEAs than movement 
of the full length CEAs due to the greater weight of the PLCEAs. 
Replacement of the PLCEAs with full length CEAs should result in a 
reduction in the probability of a dropped CEA.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The proposed changes introduce no new mode of plant operation 
and are considered to be administrative in nature. Operating 
experience has shown that the full length CEAs are capable of 
controlling the axial power distribution function intended for the 
PLCEAs. The PLCEAs will be replaced with the same type of full 
length CEAs used in shutdown and regulating CEA groups.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin Safety.
    The proposed changes may improve overall safety margins. 
Replacement of the PLCEAs with full length CEAs and including these 
Group P CEAs in the CEA drop time testing will allow ANO-2 to credit 
these CEAs in the shutdown margin calculations. This should result 
in an increase in the available shutdown margin during reactor 
operation.
    Therefore, this change 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: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: April 4, 1995
    Description of amendment request: The proposed amendment revises 
the containment cooling response time to reduce the likelihood of a 
water hammer event in service water piping.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or consequences of an Accident Previously Evaluated.
    The containment cooling system and the service water system are 
not considered to be accident initiators for any analyzed accident. 
The containment cooling system functions to mitigate the effects of 
a Main Steam Line Break (MSLB) or Loss of Coolant Accident (LOCA) on 
the containment environment. The proposed change does not affect the 
limiting MSLB analysis as the proposed increase in containment 
cooling response time is only instituted on a loss of off-site 
power. The limiting LOCA analysis has been evaluated with respect to 
the proposed containment cooling response time. Although the 
analysis shows an increase in the containment peak pressure 
(approximately 0.1 psig), this increase in the peak containment 
pressure is not considered significant since the MSLB accident with 
off-site power available is still the overall limiting accident 
condition with respect to containment peak pressure. The containment 
peak conditions for the LOCA and MSLB analyses remain below the 
original Final Safety Analysis Report (FSAR) conditions of 53.4 psig 
and 288 deg.F.
    Therefore, this change does not involve a significant increase 
int he probability or consequences of any accident previously 
evaluated.

[[Page 37091]]

    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The proposed change in containment cooling response time 
introduces no new mode of plant operation. Containment cooling 
response time is an analytical input and is not considered to be the 
initiator of any accident condition.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    The increase in containment cooling response time has been 
evaluated with respect to the accident analyses resulting in peak 
containment pressures. This evaluation has shown no significant 
increase in the resulting peak containment pressure since the 
overall limiting accident with respect to containment pressure is 
still the MSLB with off-site power available. The containment peak 
conditions for the LOCA and MSLB analyses remain below the original 
FSAR conditions of 53.4 psig and 288 deg.F.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Therefore, based upon the reasoning presented above and the 
previous discussion of the amendment request, Entergy Operations has 
determined that the requested change does not involve a significant 
hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: May 25, 1995
    Description of amendment request: The proposed amendment revises 
the Physical Security Plan vital island requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    The accident mitigation features of the plant are not affected 
by the proposed change. This change provides an equivalent level of 
protection to the plant and is adequate for preventing an 
unacceptable risk to public health and safety. This is due to 
continued compliance with existing regulatory requirements, the 
integral defense in depth design of the security program, including 
programs in place to minimize the threat of insiders, and 
historically high system reliability. The SBO (Station Blackout 
diesel) is designed with sufficient capacity to accommodate station 
blackout needs as well as those required for security. Ample 
protection against a design basis security threat continues to be 
provided. Therefore, this change does not increase the probability 
or consequences of an accident previously evaluated.
    The Station Blackout diesel generator has been approved and 
accepted by the Staff pursuant 10CFR50.63. New systems, modes of 
equipment operation, failure modes, or other plant perturbations are 
not introduced by this change. The change provides an equivalent 
level of protection, does not decrease the effectiveness of the 
overall security program and is adequate for preventing an 
unacceptable risk to public health and safety. Ample protection 
against a design basis security threat continues to be provided with 
overall physical protection of the plant maintained. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    This change does not change a safety limit, an LCO (Limiting 
Condition of Operation), or a surveillance requirement on equipment 
required to operate the plant. It is equivalent in level of 
protection, does not decrease the effectiveness of the security 
program and is adequate for preventing an unacceptable risk to 
public health and safety. The SBO diesel generator will provide an 
adequate alternative source of power to security systems. Ample 
protection against a design basis security threat continues to be 
provided. Therefore, this change does not involve a reduction in the 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005
    NRC Project Director: William D. Beckner

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

    Date of amendment request: May 22, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 4.8.1.1.2.e.7 to allow the 
performance of the 24-hour surveillance test of the diesel generators 
during power operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    The proposed change to permit the 24 hour surveillance test of 
the diesels to be performed during power operation does not increase 
the chances for a previously analyzed accident to occur. The 
function of the diesels is to supply emergency power in the event of 
a loss of offsite power. Operation of the diesels is not a precursor 
to any accident. Furthermore, the diesel generator being tested will 
remain operable and will be available to supply emergency loads 
within the required time. In addition, the two remaining diesel 
generators will be operable during the test. Consequently, if an 
offsite disturbance were to occur that affected the operability of 
the diesel being tested, the two remaining diesels would be capable 
of feeding the loads necessary for safe shutdown of the plant. This 
addresses the concerns raised in Information Notice 84-69 regarding 
the operation of emergency diesel generators connected in parallel 
with offsite power. In summary, the proposed changes do not 
adversely affect the performance or the ability of the diesel 
generators to perform their intended function.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment to the 24 hour surveillance test will not 
affect the operation of any safety system or alter its response to 
any previously analyzed accident. The diesel will automatically 
transfer from the test mode if necessary to supply emergency loads 
in the required time. The test mode is used for the monthly 
surveillance of the diesel generators as well, therefore, no new 
plant operating modes are introduced. In the event the diesel fails 
the surveillance test, it will be declared inoperable and the 
actions 

[[Page 37092]]
required for an inoperable diesel will be performed. The remaining two 
diesel generators will be operable and are capable of feeding the 
loads necessary for safe shutdown of the plant.
    Therefore, the proposed change 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 the margin of safety.
    The proposed amendment will not reduce availability of the 
diesel generator being tested to provide emergency power in the 
event of a loss of offsite power. If a loss of offsite power occurs 
during the surveillance test, the diesel generator output breaker 
will be tripped by the directional over-current relay on the ESF 
transformer. The diesel generator will transfer to the emergency 
mode, and the ESF undervoltage logic will initiate Mode II (Loss of 
Offsite Power) operation of the ESF load sequencer to supply 
emergency loads from the diesel generator. If a Loss of Coolant 
Accident occurs during the surveillance test, the diesel generator 
output breaker will be opened by a signal from the Solid State 
Protection System and the preferred offsite source will continue to 
provide power to the ESF bus. The diesel generator will continue to 
run in the emergency mode and would be available to automatically 
supply safety-related loads during any loss of offsite power 
condition. The test mode to emergency mode transfer is tested once 
per cycle in accordance with Surveillance Requirement 
4.8.1.1.2.e.10. In addition, the two remaining generators will be 
operable during the test. Consequently, if an offsite disturbance 
were to occur that affected the operability of the diesel being 
tested, the two remaining diesels would be capable of feeding the 
loads necessary for safe shutdown of the plant. The time required 
for the diesel being tested to pick up emergency loads will not be 
affected by performing the 24 hour surveillance test during power 
operation.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, N.W., Washington, D.C. 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: May 25, 1995
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) on containment leakage, to 
make the action statement consistent with the need to perform Type C 
testing at power, and to replace the surveillance requirements with a 
single requirement to apply the requirements of Appendix J as modified 
by approved exemptions. The proposed amendment would also revise the 
TSs on containment integrity, containment leakage, and containment air 
locks, to eliminate the numerical value of calculated peak containment 
internal pressure related to the design basis accident. In addition, 
there is an associated proposed exemption, from the requirements of 10 
CFR Part 50, Appendix J, to allow the performance of the required 
periodic Type C tests during power operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The change to the action statement of Technical Specification 
3.6.1.2 does not significantly increase the probability of an 
accident because leakage rate testing is not an accident initiator. 
The consequences of an accident previously evaluated are not 
increased by changing the ACTION statement of Technical 
Specification 3.6.1.2 because the requirements for CONTAINMENT 
INTEGRITY are not reduced. The consequences of an accident 
previously evaluated are not increased by the change in the 
surveillance wording because no technical changes are proposed. The 
underlying purpose of the proposed change to the Technical 
Specifications and requested exemption to Appendix J, to allow 
surveillance credit for at-power Type C testing, will not increase 
the consequences of an accident because there are no reductions in 
the requirements to maintain containment integrity.
    The proposed change to delete the numeric value of Pa is purely 
administrative, and has no potential effect on accident initiation 
or consequences.
    2. Does the change create the possibility of a new or different 
kind of accident from any previously evaluated?
    Nothing associated with the requested changes will physically 
change the configuration of the plant or impose new operating 
configurations not previously considered. Leakage rate testing will 
remove components and trains from service; however, this is not 
operationally different from other testing and maintenance 
evolutions that remove components or trains from service, and which 
were previously considered. Consequently, the possibility of a new 
or different kind of accident from any previously evaluated is not 
created.
    3. Does this change involve a significant reduction in the 
margin of safety?
    The margin of safety is not significantly reduced by changing 
the ACTION statement of Technical Specification 3.6.1.2 because the 
requirements for CONTAINMENT INTEGRITY are not reduced. The margin 
of safety is not reduced by the change in the surveillance wording 
because no technical changes are proposed. The underlying purpose of 
the proposed change to the Technical Specifications and requested 
exemption to Appendix J, to allow surveillance credit for at-power 
Type C testing, will not reduce the margin of safety because there 
are no reductions in the requirements to maintain containment 
integrity.
    The proposed change to delete the numeric value of Pa is 
purely administrative, and has no potential effect on the margin of 
safety because the value itself is unchanged.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, N.W., Washington, D.C. 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: May 30, 1995
    Description of amendment request: The proposed amendment would 
increase the spent fuel pool heat load licensing basis to provide 
greater flexibility for normal refueling practices.
    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:

[[Page 37093]]

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because:
    (a) The Spent Fuel Pool conditions are not indicative of 
accident initiators.
    (b) Design and operability requirements of equipment important 
to safety are not affected.
    (c) If only one Spent Fuel Pool cooling train is available, 
boiling would not occur and the Spent Fuel Pool components would 
remain within their design basis.
    (d) The complete loss of Spent Fuel Pool cooling event has 
previously been analyzed and described in Supplement 6 to the Safety 
Evaluation Report, Appendix BB. The dose consequences for this event 
have been evaluated and the safety evaluation is described in 
Updated Safety Analysis Report Section 9.1.3.3.4. The results of the 
evaluation show that the Spent Fuel Pool components would remain 
within their design bases. Also, the dose consequences of iodine 
release as a result of Spent Fuel Pool boiling are significantly 
below the allowable dose limits of 10 CFR 100.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously because:
    (a) The operability of safety-related equipment is not impacted.
    (b) The probability of safety-related equipment malfunctioning 
is not increased.
    (c) The scope of the change does not establish a potential new 
accident precursor.
    (d) The Spent Fuel Pool design considers design basis heat loads 
for the modified refueling procedure which includes a full-core 
offload.
    (e) For the design basis case, the integrity of the Spent Fuel 
Pool Boraflex is not adversely impacted.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety because:
    (a) No fuel damage would occur as a result of the proposed 
change.
    (b) Technical Specification operability and surveillance 
requirements are not reduced.
    (c) The Spent Fuel Pool boiling doses would be significantly 
below the allowable dose limits of 10 CFR 100.
    (d) The modified refueling procedure (full-core offload) 
continues to have acceptable margins of safety.
    (e) For the design basis case, the integrity of the Spent Fuel 
Pool Boraflex is not adversely impacted.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, N.W., Washington, D.C. 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: June 9, 1995
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 4.1, ``Site Location,'' to 
incorporate a description of the exclusion area boundary. The proposed 
change is necessary to ensure the content of the TS conforms to Section 
182 of the Atomic Energy Act of 1954.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration which 
is presented below:
    (1) The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any of the parameters or conditions that could contribute to 
the initiation of any accidents previously evaluated. In addition, 
the physical location of the [exclusion area boundary] EAB has not 
been changed; a description of its location has merely been added to 
the TS. Thus, the proposed change cannot increase the probability or 
the consequences of any accident previously evaluated.
    (2) The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any parameter or condition that could contribute to the 
initiation of any accidents. Thus, the proposed change cannot create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    (3) The proposed change only affects regulatory controls on the 
accepted configuration of the EAB. The proposed change does not 
involve an actual change to the location of the EAB. The proposed 
change will restore compliance with the Atomic Energy Act of 1954 
and require prior NRC approval of any changes to the physical 
location of the EAB. As a result, IP has concluded that the proposed 
change will not result in a reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727
    Attorney for licensee: Leah Manning Stetzner, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th St., Decatur, 
Illinois 62525
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: February 3, 1995, as supplemented April 
25, 1995 (AEP:NRC:1166Q and 1166R)
    Description of amendment request: The proposed amendment would 
allow continued use of a steam generator (SG) tube support plate 
interim plugging criteria for fuel cycle 15. The change would allow SG 
tubes with bobbin coil eddy current indications less than or equal to 
2.0 volts at tube support plate intersections to remain in service if 
the projected end-of-cycle distribution of crack indications is shown 
to result in primary-to-secondary leakage less than 12.6 gpm during a 
postulated steam line break (SLB). The change would also allow 
indications greater than 2.0 volts but less than or equal to 5.6 volts 
to remain in service if a motorized rotating pancake coil probe 
inspection does not detect degradation.
    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:
    In accordance with the three factor test of 10 CFR 50.92(c), 
implementation of the proposed license amendment is analyzed using 
the following standards and found not to 1) involve a significant 
increase in the probability or consequences of an accident 
previously evaluated; 2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; 
or 3) involve a significant reduction in margin of safety. 
Conformance of the proposed amendment to the standards for a 
determination of no significant hazards as defined in 10 CFR 50.92 
(three factor test) is shown in the following paragraphs.
    1) Operation of Cook Nuclear Plant Unit 1 in accordance with the 
proposed license amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. Testing of model boiler specimens for free span tubing 
(no tube support plate restraint) at room temperature conditions 
show burst pressures in excess of 5000 psi for indications of outer 
diameter stress corrosion cracking with voltage measurements as high 
as 19 volts. Burst testing performed on pulled tubes from Cook 
Nuclear Plant Unit 1 with up to a 2.02 volt indication shows 
measured burst pressure in excess of 10,000 psi at room temperature. 
Burst testing performed on pulled tubes from other plants with up to 
7.5 volt indications show burst pressures in excess of 6,300 psi at 
room temperatures. Correcting for the effects of temperature on 

[[Page 37094]]
material properties and minimum strength levels (as the burst testing 
was done at room temperature), tube burst capability significantly 
exceeds the safety factor requirements of RG [Regulatory Guide] 
1.121 [``Bases for Plugging Degraded PWR Steam Generator Tubes'']. 
As stated earlier, tube burst criteria are inherently satisfied 
during normal operating conditions due to the proximity of the tube 
support plate [TSP]. Test data indicates that tube burst cannot 
occur within the tube support plate, even for tubes which have 100% 
throughwall electric-discharge machined notches 0.75 inch long, 
provided the tube support plate is adjacent to the notched area. 
Since tube-to-tube support plate proximity precludes tube burst 
during normal operating conditions, use of the criteria must, 
therefore, retain tube integrity characteristics which maintain the 
RG 1.121 margin of safety of 1.43 times the bounding faulted 
condition (steam line break) pressure differential.
    During a postulated main steam line break, the TSP has the 
potential to deflect during blowdown, thereby uncovering the 
intersection. Based on the existing data base, the RG 1.121 
criterion requiring maintenance of a safety factor of 1.43 times the 
steam line break pressure differential on tube burst is satisfied by 
7/8 inch diameter tubing with bobbin coil indications with signal 
amplitudes less than 9.6 volts, regardless of the indicated depth 
measurement. A 2.0 volt plugging criteria compares favorably with 
the 9.6 volt structural limit considering the previously calculated 
growth rates for ODSCC [outer diameter stress corrosion cracking] 
within Cook Nuclear Plant Unit 1 SGs. Considering a voltage growth 
component of 0.8 volts (40% voltage growth based on 2.0 volts BOC 
[beginning of cycle]), and a nondestructive examination uncertainty 
of 0.40 volts (20% voltage uncertainty based on 2.0 volts BOC), when 
added to the BOC IPC [interim plugging criteria] of 2.0 volts, 
results in a bounding EOC [end of cycle] voltage of approximately 
3.2 volts for cycle 15 operation. A 6.4 volt safety margin exists 
(9.6 structural limit - 3.2 volt EOC - 6.4 volt margin).
    For the voltage/burst correlation, the EOC structural limit is 
supported by a voltage of 9.6 volts. Using this structural limit of 
9.6 volts, a BOC maximum allowable repair limit can be established 
using the guidance of RG 1.121. The BOC maximum allowable repair 
limit should not permit a significant number of EOC indications to 
exceed the 9.6 volt structural limit and should assure that 
acceptable tube burst probabilities are attained. By adding NDE 
[nondestructive examination] uncertainty allowances and an allowance 
for crack growth to the repair limit, the structural limit can be 
validated. The previous IPC submittal established the conservative 
NDE uncertainty limit of 20% of the BOC repair limit. For 
consistency, a 40% voltage growth is extremely conservative for Cook 
Nuclear Plant Unit 1. Therefore, the maximum allowable BOC repair 
limit (RL) based on the structural limit of 9.6 volts can be 
represented by the expression:
    RL + (0.2 x RL) + (0.4 x RL) = 9.6 volts, or,the maximum 
allowable BOC repair limit can be expressed as,
    RL = 9.6 volt structural limit/1.6 = 6.0 volts
    This structural repair limit supports this application for cycle 
15 IPC implementation to repair bobbin indications greater than 2.0 
volts based on RPC [rotating pancake coil] confirmation of the 
indication. Conservatively, an upper limit of 5.6 volts will be used 
to repair bobbin indications which are above 2.0 volts but do not 
have confirming RPC calls.
    The conservatism of this repair limit is shown by the EOC 13 
(Spring 1994) eddy current data. The overall average voltage growth 
was determined to be on 1.4% (of the BOC voltage). In addition, the 
EOC 13 maximum observed voltage increase was 0.40 volts, and 
occurred in a tube with a BOC indication of 0.96 volts. The 
applicability of cycle 14 growth rates for cycle 15 operation will 
be confirmed prior to return to service of Cook Nuclear Plant Unit 
1. Similar large structural margins are anticipated.
    Relative to the expected leakage during accident condition 
loadings, it has been previously established that a postulated main 
steam line break outside of containment but upstream of the main 
steam isolation valve represent the most limiting radiological 
condition relative to the IPC. In support of implementation of the 
IPC, it will be determined whether the distribution of crack 
indications at the tube support plate intersections at the end of 
cycle 15 are projected to be such that primary to secondary leakage 
would result in site boundary doses within a small fraction of the 
10 CFR 100 guidelines. A separate calculation has determined this 
allowable steam line break leakage limit to be 12.6 gpm. Although 
not required by the Cook Nuclear Plant design basis, this 
calculation uses the recommended Iodine-131 transient spiking values 
consistent with NUREG-0800 [Standard Review Plan], and the T/S 
[technical specification] reactor coolant system activity limit of 
1.0 micro curie per gram dose equivalent Iodine-131. The projected 
steam line break leakage rate calculation methodology prescribed in 
[Draft] GL 94-XX [``Voltage-Based Repair Criteria for the Repair of 
Westinghouse Steam Generator Tubes Affected by Outside Diameter 
Stress Corrosion Cracking,'' August 12, 1994] and WCAP 14277 [``SLB 
Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at 
TSP Intersections''] will be used to calculate EOC 15 leakage, based 
on actual EOC 14 distributions and EOC 15 projected distributions. 
Due to the relatively low voltage growth rates at Cook Nuclear Plant 
Unit 1 and the relatively small number of indications affected by 
the IPC, steam line break leakage prediction per GL 94-XX is 
expected to be significantly less than the acceptance limit of 12.6 
gpm in the faulted loop.
    Prior to issue of GL 94-XX, projected EOC 14 leak rates were 
calculated, based on draft NUREG-1477 [``Voltage-Based Interim 
Plugging Criteria for DG Tubes, Draft for Comments''], for a total 
of twelve cases, the combination of six probability-of-leak 
correlations and two leak rate calculation methodologies. Results of 
the calculations show that the projected EOC 14 leak rates ranged 
from 0.001 gpm to 1.360 gpm. These results are well below the 12.6 
gpm allowable; therefore, implementation of the 2 volt IPC during 
cycle 15 would not adversely affect SG tube integrity and results in 
acceptable dose consequences.
    Current GL 94-XX methodology requires only the log-logistic 
probability of leakage correlation be used. Projected EOC 14 SLB 
leakage using this function was calculated to be only 0.001 gpm. 
Based on the relatively few numbers of intersections at Cook Nuclear 
Plant Unit 1 to which the IPC are applied and extremely small Cook 
Nuclear Plant Unit 1 plant-specific growth rate, a similar value 
would be expected based on the EOC 14 eddy current data. The 
inclusion of all IPC intersections in the leakage model, along with 
application of a probability of detection of 0.6, will result in 
extremely conservative leakage estimations, especially so since 
close examination of the available data shows that indications of 
less than 2.8 volts will not be expected to leak during SLB 
conditions. All Unit 1 IPC indications are expected to be below 2.8 
volts at the EOC 15 conditions.
    The proposed amendment does not result in any increase in the 
probability or consequences of an accident previously evaluated 
within the Cook Nuclear Plant Unit 1 FSAR.
    2) The proposed license amendment does not create the 
possibility of a new or different kind of accident previously 
evaluated.
    Implementation of the proposed SG tube IPC does not introduce 
any significant changes to the plant design basis. Use of the 
criteria does not provide a mechanism which could result in an 
accident outside of the region of the tube support plate elevations. 
Neither a single or multiple tube rupture event would be expected in 
a SG in which the plugging criteria has been applied (during all 
plant conditions).
    Specifically, we will continue to implement a maximum leakage 
rate limit of 150 gpd (0.1 gpm) per SG to help preclude the 
potential for excessive leakage during all plant conditions. The 
cycle 15 T/S limits imposed on primary to secondary leakage at 
operating conditions are: a maximum of 0.4 gpm (600 gpd) for all SGs 
with a maximum of 150 gpd allowed for any one SG.
    The RG 1.121 criteria for establishing operational leakage rate 
limits that require plant shutdown are based upon leak-before-break 
considerations to detect a free span crack before potential tube 
rupture during faulted plant conditions. The 150 gpd limit should 
provide for leakage detection and plant shutdown in the event of the 
occurrence of an unexpected single crack resulting in leakage that 
is associated with the longest permissible crack length. Regulatory 
Guide 1.121 acceptance criteria for establishing operating leakage 
limits are based on leak-before-break considerations such that plant 
shutdown is initiated if the leakage associated with the longest 
permissible crack is exceeded. The longest permissible crack is the 
length that provides a factor of safety of 1.43 against bursting at 
faulted conditions maximum pressure differential. A voltage 
amplitude of 9.6 volts for typical ODSCC corresponds to meeting this 
tube burst requirement at a lower 95% prediction limit on the burst 
correlation 

[[Page 37095]]
coupled with 95/95 lower tolerance limit material properties. Alternate 
crack morphologies can correspond to 9.6 volts so that a unique 
crack length is not defined by the burst pressure versus voltage 
correlation. Consequently, typical burst pressure versus through-
wall crack length correlations are used below to define the 
``longest permissible crack'' for evaluating operating leakage 
limits.
    Consistent with the Cycle 13 and Cycle 14 license amendment 
requests for IPC and Section 5 of Enclosure 1 of the GL, operational 
leakage limits will remain at 150 gpd per SG. Axial cracks leaking 
at this level are expected to provide leak before break (LBB) 
protection at both the SLB pressure differential of 2560 psi and, 
while not part of any established LBB methodology, LBB protection 
will also be provided at a value of 1.43 times the SLB pressure 
differential. Thus, the 150 gpd limit provides for plant shutdown 
prior to reaching critical crack lengths for steam line break 
conditions. Additionally, this leak-before-break evaluation assumes 
that the entire crevice area is uncovered during blowdown. Partial 
uncovery will provide benefit to the burst capacity of the 
intersection.
    3) The proposed license amendment does not involve a significant 
reduction in margin of safety.
    The use of the voltage based bobbin probe interim tube support 
plate elevation plugging criteria at Cook Nuclear Plant Unit 1 is 
demonstrated to maintain SG tube integrity commensurate with the 
criteria of RG 1.121. Regulatory Guide 1.121 describes a method 
acceptable to the NRC staff for meeting GDC [General Design 
Criteria] 14, 15, 31, and 32 by reducing the probability or the 
consequences of SG tube rupture. This is accomplished by determining 
the limiting conditions of degradation of SG tubing, as established 
by inservice inspection, for which tubes with unacceptable cracking 
should be removed from service. Upon implementation of the criteria, 
even under the worst case conditions, the occurrence of ODSCC at the 
tube support plate elevations is not expected to lead to a SG tube 
rupture event during normal or faulted plant conditions. The EOC 15 
distribution of crack indications at the tube support plate 
elevations will be confirmed by analysis and calculation to result 
in acceptable primary to secondary leakage during all plant 
conditions and that radiological consequences are not adversely 
impacted.
    In addressing the combined effects of a LOCA [loss-of-coolant 
accident] and SSE [safe-shutdown earthquake] on the SG component (as 
required by GDC 2), it has been determined that tube collapse may 
occur in the SGs at some plants. This is the case as the tube 
support plates may become deformed as a result of lateral loads at 
the wedge supports at the periphery of the plate due to the combined 
effects of the LOCA rarefaction wave and SSE loadings. Then, the 
resulting pressure differential on the deformed tubes may cause some 
of the tubes to collapse.
    There are two issues associated with SG tube collapse. First, 
the collapse of SG tubing reduces the RCS [reactor coolant system] 
flow area through the tubes. The reduction in flow area increases 
the resistance to flow of steam from the core during a LOCA which, 
in turn, may potentially increase peak clad temperature. Second, 
there is a potential that partial through-wall cracks in tubes could 
progress to through-wall cracks during tube deformation or collapse.
    Consequently, since the leak-before-break methodology is 
applicable to the Cook Nuclear Plant Unit 1 reactor coolant loop 
piping, the probability of breaks in the primary loop piping is 
sufficiently low that they need not be considered in the structural 
design of the plant. The limiting LOCA event becomes either the 
accumulator line break or the pressurizer surge line break. Loss of 
coolant accident loads for the primary pipe breaks were used to 
bound the Cook Nuclear Plant Unit 1 smaller breaks. The results of 
the analysis using the larger break inputs show that the LOCA loads 
were found to be of insufficient magnitude to result in SG tube 
collapse or significant deformation.
    Addressing RG 1.83 [``Inservice Inspection of PWR Steam 
Generator Tubes''] considerations, implementation of the bobbin 
probe voltage based interim tube plugging criteria of 2.0 volts is 
supplemented by enhanced eddy current inspection guidelines to 
provide consistency in voltage normalization, a 100% eddy current 
inspection sample size at the tube support plant elevations per T/S, 
and MRPC [Motorized Rotating Pancake Coil] inspection requirements 
for the larger indications left in service to characterize the 
principal degradation as ODSCC.
    As noted previously, implementation of the tube support plate 
elevation plugging criteria will decrease the number of tubes which 
must be repaired. The installation of SG tube plugs reduces the RCS 
flow margin. Thus, implementation of the IPC will maintain the 
margin of flow that would otherwise be reduced in the event of 
increased tube plugging.
    Based on the above, it is concluded that the proposed license 
amendment request does not result in a significant reduction in 
margin with respect to plant safety as defined in the Final Safety 
Analysis Report or any Bases of the plant T/Ss.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment requests: May 26, 1995 (AEP:NRC:1207)
    Description of amendment requests: The proposed amendments would 
change multiple operating limits on both units. The primary change 
would allow operation of Cook Unit 1 with steam generator plugging 
levels as high as 30% in each steam generator. The second group of 
changes would modify the overtemperature delta T and overpower delta T 
reactor trip setpoints for both units and increase the allowed 
degradation of the Unit 1 auxiliary feedwater pumps consistent with 
Unit 2. The third group of changes would reduce the required shutdown 
margin in modes 1, 2, 3, and 4, increase the allowable centrifugal 
charging pump head degradation, reduce the minimum refueling water 
storage tank temperature, and revise the peak pressure of the long-term 
containment integrity analysis in the bases. Finally, certain 
administrative changes are also proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    10 CFR 50.92 specifies that the holder of an operating license 
or construction permit of a nuclear power facility participate in 
determining whether a change to the T/S's current licensing basis 
(CLB) involves a significant hazards consideration. Prior to 
implementation of a change to the CLB, the Nuclear Regulatory 
Commission must review and make a final determination, pursuant to 
the procedures in 10 CFR 50.91, that a proposed amendment to the 
operating license involves no significant hazards considerations. In 
order to satisfactorily complete the review, the proposed amendment 
to the CLB must not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated,
    2. create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. involve a significant reduction in a margin of safety.
    For the purpose of performing a significant hazards 
consideration analysis, the four groups of technical specification 
changes discussed under Description of Changes can be reduced to 
three groups. In evaluating significant hazards, the first three 
groups of proposed technical specifications will be considered 
together. The miscellaneous change and the administrative change 
will each be considered separately.

[[Page 37096]]

    Determination Of No Significant Hazards For Changes Based On 
Analyses And Evaluations (Groups 1, 2, and 3)
    Criterion 1
    Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The analyses which were performed to support the first three 
groups of proposed changes were performed in accordance with 
approved methodologies and acceptance criteria applicable to Cook 
Nuclear Plant. The proposed technical specification changes do not 
involve postulated initiators for analyzed events. Therefore, the 
probability of accidents can not be affected. The analyses and 
evaluations performed all met applicable acceptance criteria. 
Therefore, the consequences of accidents previously evaluated are 
unaffected.
    Criterion 2
    Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The analyses which were performed to support the second and 
third groups of proposed changes address increases in operating 
margin for accident mitigators. They do not create the possibility 
of new accidents. The first group of proposed changes to reduce 
minimum measured primary flow, increase the DNB [departure from 
nucleate boiling] temperature limit, and reduce the reactor coolant 
system volume have been analyzed or evaluated. The proposed DNB 
limit is consistent with the DNB design and does not constitute an 
accident initiator. The new volume results from the new value of 
allowed tube plugging and is consistent with the analysis. It is not 
an accident initiator.
    The impact of the reduced primary flow in the primary system was 
analyzed or evaluated, as appropriate. All applicable criteria were 
satisfied. No new or different kind of accident resulted.
    Criterion 3
    Do the proposed changes involve a significant reduction in a 
margin of safety?
    No. The margin of safety is provided for the primary pressure 
boundary and other components in part by applicable design codes. 
The margin of safety for the various accidents and transients is 
maintained by the analysis acceptance criteria. Since the components 
remain in compliance with the codes and standards in effect when 
Cook Nuclear Plant was licensed and applicable acceptance criteria 
are met, the margin of safety is not reduced by the 30% SGTP [steam 
generator tube plugging] program.
    Determination Of No Significant Hazards For Administrative 
Changes (Group 4)
    Criterion 1
    Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed change involves the surveillances for 
mitigating equipment. Therefore, it has no impact on probability. 
The proposed change also has no impact on the consequences of an 
accident because the criteria for operable RHR [residual heat 
removal] and SI [safety injection] pumps does not change. The change 
is only in the parameter that will be compared with the required 
criteria, the differential pressure instead of the discharge 
pressure.
    Criterion 2
    Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. Nothing is changed with regard to accident initiators. The 
surveillance criteria for the RHR and SI pumps, which are mitigating 
equipment, is unchanged. The proposed change can have no impact on 
accident initiators.
    Criterion 3
    Does the proposal involve a significant reduction in a margin of 
safety?
    No. The proposal does not change the requirements for a pump to 
be operable. Only the parameter compared to acceptance criteria 
changes. The underlying criteria is unchanged. Therefore, there is 
no change in the margin of safety.
    Conclusion
    It is concluded that operation of Cook Nuclear Plant units 1 and 
2 with the changes proposed above does not involve any significant 
hazards as defined in 10 CFR 50.92
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment requests: June 15, 1995 (AEP:NRC:0896V)
    Description of amendment requests: The proposed amendments would 
change the 18 month emergency diesel generator (EDG) surveillance test 
from a 24-hour run to an 8-hour run.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Per 10 CFR 50.92, a proposed change does not involve a 
significant hazards consideration if the change does not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated,
    2. create the possibility of a new of different kind of accident 
from any accident previously evaluated, or
    3. involve a significant reduction in a margin of safety.
    Criterion 1
    The safety function of the EDGs is to supply ac electrical power 
to plant safety systems whenever the preferred ac power supply is 
unavailable. Through surveillance requirements, the ability of the 
EDGs to meet their load and timing requirements is tested and the 
quality of the fuel and the availability of the fuel supply are 
monitored. Reduction of the 24 hour run to 8 hours will not reduce 
the surveillance factors under consideration and will sufficiently 
exercise the EDG and its support systems to identify potential 
conditions that could lead to performance degradation. Based on 
these considerations, it is concluded that the proposed changes do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Criterion 2
    The proposed changes do not involve physical changes to the 
plant or changes in plant operating configuration. The changes only 
involve the reduction of 18 month 24 hour EDG surveillance test 
duration. Thus, it is concluded that the proposed changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    Criterion 3
    Although the duration of the 18 month 24 hour EDG surveillance 
test would be reduced, the EDG components will continue to be 
sufficiently exercised such that the ability to detect incipient and 
degraded conditions will be maintained. The proposed changes have 
been determined to be compatible with our plant operating experience 
and commensurate with past surveillance test results. Based on these 
considerations, it is concluded that the proposed changes do not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: John N. Hannon

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: June 15, 1995
    Description of amendment request: The proposed amendment would 
revise 

[[Page 37097]]
the definition for logic system functional test and revise the 
surveillance interval for emergency core cooling system logic system 
functional testing from 6 months to 18 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed revisions to change the Cooper Nuclear Station 
(CNS) Emergency Core Cooling System (ECCS) logic system functional 
testing surveillance intervals from once/6 months to once/18 months 
do not involve a significant increase in the probability or 
consequences of an accident previously evaluated. The change in 
surveillance interval to once/18 months is necessary to coincide 
with scheduled refueling outages. The expansion of the scope of the 
logic system functional tests will ensure that once/18 months all 
contacts providing an automatic safety function in the ECCS logic 
systems will be tested. Revising the test frequency to once/18 
months will prevent CNS from being required to install jumpers and/
or test blocks during power operation, temporarily rendering various 
safety functions inoperable, and potentially challenging safety 
systems.
    This proposed change will not result in any hardware changes to 
the facility, nor will it introduce any new mode of operation. 
Conversely, not changing the surveillance frequency would contribute 
to a slight, but measurable increase in the probability of an 
accident. Therefore, this change will not result in a significant 
increase in the probability of any accident previously evaluated.
    This change will not result in a significant increase in the 
consequences of any accident previously evaluated. The District has 
evaluated the change in logic system reliability due to the 
increased proposed surveillance interval and determined it to be 
negligible. This conclusion is supported by a review of the 
surveillance history associated with the ECCS logic system 
functional tests which demonstrates that the logic systems perform 
reliably. Therefore, this change will not result in a significant 
reduction in the reliability or performance of the ECCS, and 
therefore, will not result in a significant increase in the 
consequences of any accident previously evaluated.
    The change to the definition for ``Logic System Functional 
Test'' will not result in an increase in the probability or 
consequences of any accident previously evaluated. This change will 
only provide clarification of the definition for performing these 
tests.
    These changes are also consistent with the NUREG-1433, 
``Standard Technical Specifications, General Electric Plants, BWR/
4,'' dated September, 1992. Therefore, these changes have been 
previously reviewed and accepted by the NRC, and have been 
implemented at other plants.
    2. Does the proposed change create the possibility for a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes revise the ECCS logic system functional 
testing surveillance intervals and the definition of that testing to 
be consistent with the Standard Technical Specifications, and 
therefore reflect current NRC guidance. The proposed changes do not 
involve any plant design changes nor any new mode of operation. 
Therefore, these proposed changes do not create the possibility for 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change create a significant reduction in 
the margin of safety?
    The proposed changes to the CNS ECCS logic system functional 
testing surveillance intervals do not create a significant reduction 
in the margin of safety. As discussed above, the District has 
revised its logic system functional testing to ensure that all 
contacts providing an automatic safety function in the ECCS logic 
systems are tested during this surveillance; thus, this change in 
testing scope will ensure that all essential functions in these 
logic systems are periodically tested.
    The proposed changes will extend the ECCS logic system 
functional testing intervals to coincide with refueling outages. 
This will prevent CNS from being required to install jumpers and/or 
test blocks during power operation which would temporarily defeat 
safety system capability, and have the potential of challenging 
plant safety systems and/or degrading logic system reliability. The 
District has also determined that the change in test frequency will 
have a negligible impact on logic system reliability. Therefore, 
since these changes will continue to ensure the reliability of the 
ECCS logic systems, and thereby the capability of those systems to 
respond to accidents, these proposed changes do not create 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: Auburn Public Library, 118 
15th Street, Auburn, NE 68305
    Attorney for licensee: Mr. G.D. Watson, Nebraska Public Power 
District, Post Office Box 499, Columbus, Nebraska 68602-0499
    NRC Project Director: William D. Beckner

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

    Date of amendment request: June 15, 1995
    Description of amendment request: The proposed amendment would 
change the definition for an alteration of the reactor core to one that 
is consistent with the intent of the improved standard technical 
specifications. The proposed amendment also makes administrative 
changes to several technical specification pages.
    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:
    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.
    Revising the definition of core alteration would not affect the 
probability or consequences of a fuel handling accident, since the 
movement of fuel within the reactor vessel would still be considered 
a CORE ALTERATION. Additionally, movement of a fuel assembly 
continues to be performed under the supervision of a senior licensed 
operator. Therefore, the potential for inadvertent positioning of a 
fuel assembly would not be affected by the change to the definition 
of a core alteration.
    Other activities which were not specifically excluded as core 
alterations in the existing technical specifications are now 
excluded. These activities do not affect the reactivity of the core.
    Based upon the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    All required systems will continue to operate as before. 
Therefore, there is no possibility of a new or different kind of 
accident. The change in definition of a core alteration cannot 
create the possibility of a new type of accident since those 
activities which affect reactivity and could affect the initiating 
events for accidents will remain classified as core alterations.
    3. Involve a significant reduction in the margin of safety.
    Refueling operations which have the potential to alter the 
reactivity potential of the core will continue to be defined as core 
alterations. The margin of safety associated with those evolutions 
will not be altered as a result of the revised definition. As a 
result of the revised definition, evolutions which take place within 
the reactor vessel core region with the vessel head installed, or 
with the reactor vessel completely defueled, will not be considered 
core alterations. This does not constitute a reduction in the margin 
of safety since there is no impact on core reactivity potential 
during these conditions.

[[Page 37098]]

    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: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270
    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: June 7, 1995
    Description of amendment request: The proposed amendment would 
increase the temperature limit below which reactor coolant sampling and 
analysis for dissolved oxygen is not required. Specifically, the 
temperature limit stated in the footnotes to Technical Specification 
Surveillance Requirement 4.4.7 and to Table 3.4-2 would be increased to 
250 deg.F from 180 deg.F.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration. The 
NRC staff has reviewed the licensee's analysis against the standards of 
10 CFR 50.92(c). The NRC staff's review is presented below.
    A. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated (10 
CFR 50.92(c)(1)) because the proposed changes merely increase the 
temperature limit below which sampling of reactor coolant for 
dissolved oxygen and maintaining the dissolved oxygen below the 
specified limit would not be required. The proposed limit is 
consistent with data which shows that there is no significant 
oxygen-induced corrosion to reactor coolant system (RCS) components 
at or below the limit. The changes do not affect the manner by which 
the facility is operated and do not change any structures, systems, 
or components. Since there is no change to the facility or to the 
way it is operated, there is no effect upon the probability or 
consequences of any accident previously analyzed.
    B. The 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 the manner by which 
the facility is operated or change any structure, system, or 
component. The proposed changes merely raise the temperature limit 
above which dissolved oxygen must be maintained within the specified 
limit. The changes are consistent with data for oxygen-induced 
corrosion of RCS components.
    C. The changes do not involve a significant reduction in a 
margin of safety (10 CFR 50.92(c)(3)) because the proposed changes 
are consistent with data for oxygen-induced corrosion of RCS 
components.
    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: Thomas Dignan, Esquire, Ropes & Gray, One 
International Place, Boston MA 02110-2624
    NRC Project Director: Phillip F. McKee

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

    Date of amendment requests: June 29, 1995 (Reference LAR 95-04)
    Description of amendment requests: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Nuclear Power Plant, Unit Nos. 1 and 2, to add Mode 1 applicability to 
TS 3/4.4.2.2, ``Safety Valves - Operating,'' and to change the low- 
temperature overpressure protection (LTOP) system enable temperature 
for Mode 4 applicability from 323 degrees F to 270 degrees F in TS 3/
4.4.2.1, ``Safety Valves - Shutdown.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes have no effect on plant operation. The 
proposed changes correct the applicability of TS 3/4.4.2.2, 
consistent with the NRC safety evaluation for License Amendments 
(LAs) 98 for Unit 1 and 97 for Unit 2, and LAs 100 for Unit 1 and 99 
for Unit 2 dated March 9, 1995, and April 13, 1995, respectively.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes are administrative in nature. Further, the 
proposed changes would not result in any physical alteration to any 
plant system, and would not be a change in the method by which any 
safety-related system performs its function.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed administrative changes correct TS 3/4.4.2.2 
applicability, consistent with previous NRC review and approval of 
LAs 98 and 97 and LAs 100 and 99, as described in the associated 
safety evaluations. Further, these proposed changes have no effect 
on current operating methodologies or actions that govern plant 
performance.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Project Director: William H. Bateman

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

    Date of amendment request: June 5, 1995
    Description of amendment request: The proposed changes will revise 
Technical Specification (TS) Section 3/4.1.5, ``Standby Liquid Control 
System,'' (SLCS), to remove the minimum flow rate requirement for the 
SLCS pumps from TS Section 3/4.1.5.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The proposed Technical Specifications (TS) change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.

[[Page 37099]]

    The proposed TS change will remove the minimum flow rate 
requirement for the Standby Liquid Control System (SLCS) pumps from 
Technical Specifications Section 3/4.1.5. The proposed TS change 
does not involve any physical change in the plant configuration or 
the SLCS pumps operation. The SLCS is not used during normal plant 
operation; therefore, there is no impact on any accident initiators. 
The proposed TS change does not change the plant response to 
transients in any way that could increase the likelihood of an 
accident. The consequences of previously evaluated accidents are not 
affected since the SLCS pumps and the balance of the SLCS will 
continue to perform as designed, in accordance with the Anticipated 
Transient Without Scram (ATWS) Rule specified in 10CFR50.62. The 
SLCS pumps will continue to be tested periodically for operability 
in accordance with TS 4.0.5 Surveillance Requirements for American 
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B 
& PV) Code Class 2 pumps, and the testing frequency remains 
unchanged.
    Therefore, the proposed TS change does not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS change will remove the minimum flow rate 
requirement for the Standby Liquid Control System (SLCS) pumps from 
Technical Specifications Section 3/4.1.5. The SLCS and the SLCS 
pumps will continue to function as currently designed. There are no 
physical changes being performed to the SLCS or plant configuration. 
The proposed TS change does not introduce a new failure mode for the 
SLCS pumps. Physical and electrical redundancy and separation 
criteria are not impacted by this proposed TS change. There is no 
change to the Redundant Reactivity Control System (RRCS) logic which 
could create an accident or transient of a different type.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident, from any 
accident previously evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The following TS Bases were reviewed for potential reduction in 
the margin of safety:
    3/4.1.5 Standby Liquid Control System
    4.0.5 Surveillance Requirements
    The margin of safety as defined in the TS Bases will remain the 
same. The specific flow rate requirement for the Standby Liquid 
Control System (SLCS) pumps is being removed from the TS since the 
Anticipated Transient Without Scram (ATWS) equation ensures 
acceptable flow rates. The SLCS pumps, which are safety-related, are 
not physically modified or impacted by the proposed TS change. The 
pumps will continue to be tested for operability, in accordance with 
TS 4.0.5 Surveillance Requirements for ASME B & PV Code Class 2 
pumps, and the testing frequency remains unchanged. This testing 
will ensure that the SLCS pumps operate in accordance with the 
existing design basis for the SLCS.
    Therefore, the proposed TS change does not involve a reduction 
in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: John F. Stolz

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

    Date of amendment request: June 20, 1995
    Description of amendment request: The proposed amendment (PA-190) 
would permit SMUD to change the reviewer qualifications of the 
Permanently Defueled Technical Specification (PDTS) D6.5.3 from those 
required by ANSI N18.1-1971, Section 4.4 to those of Section 4. In 
addition, PDTS D6.9.6b, Environmental Reports, would be changed to 
permit annual reporting instead of the current semi-annual schedule.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has reviewed 
the proposed changes against each of the no significant hazards 
consideration criteria in 10 CFR 50.92, and, based on their safety 
analysis, concludes:
    A significant increase in the probability or consequences of an 
accident previously evaluated will not be created, because the 
proposed PDTS changes (1) are administrative in nature, (2) have no 
effect on any credible accidents previously evaluated in the Rancho 
Seco Defueled Safety Analysis Report (DSAR) (i.e., the dropped fuel 
assembly accident, the loss of off-site power condition, or a 
radwaste tank rupture), (3) will not reduce the effectiveness of the 
reviews conducted because the Rancho Seco Qualified Reviewer 
training program ensures Qualified Reviewers have adequate skills to 
competently perform the required reviews and the Plant Review 
Committee will continue to conduct their second level review 
function, and (4) will only affect the timing and management of the 
required Environmental Reports submittals to the NRC.
    PA-190 will not create the possibility of a new or different 
type of accident than previously evaluated, because the proposed 
PDTS changes (1) do not modify the configuration of the facility or 
affect facility operation during the PDM [permanently defueled 
mode], (2) are administrative in nature, and (3) do not provide any 
new mechanisms by which an accident can occur.
    The proposed PDTS amendment will not involve a significant 
reduction in the margin of safety, because the proposed changes do 
not affect the operation of Rancho Seco or any plant systems. Also, 
The PDTS bases do not rely on (1) Qualified Reviewer qualification 
requirements or (2) submittal of PDTS D6.9.6b Environmental Reports 
to the NRC to provide a margin of safety for plant operation during 
the PDM. The Rancho Seco Qualified Reviewer program relies on 
training and not the ANSI N18.1 qualification requirements to ensure 
the PDTS D6.5.3 required reviews are competently performed. 
Therefore, the proposed changes will not involve a reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis, and based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration. 
Local Public Document Room location: 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

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

    Date of amendment request: June 19, 1995
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to delete the scheduler 
requirements for Type A testing (Overall Integrated Containment Leakage 
Rate) to be performed at 40 plus or minus 10 month intervals and to 
delete the schedular requirements for Type B and C tests to be 
performed at 24 month intervals.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The probability or consequences of an accident previously 
evaluated is not significantly increased.
    There is no increase in the probability of an accident since 
there is no work planned that would affect containment integrity. 
The testing of containment isolation valves and 

[[Page 37100]]
other containment penetration sealing devices is not postulated as an 
accident precursor or initiating event.
    Type A testing is capable of determining the total leakage from 
both local leak paths as well as gross containment leakage paths. 
Our Type B and C testing has consistently provided accurate leakage 
rates for valves and penetrations.
    Administrative controls govern maintenance and testing such that 
there is very low probability that unacceptable maintenance or 
alignments can occur. After maintenance on containment isolation 
valves (CIVs) and penetrations, a local leak rate test (LLRT) is 
required to be performed. All work on valves also requires that an 
independent valve lineup be performed. As a result, Type A testing 
is not required to accurately quantify the leakage through 
containment penetrations.
    Any specific exemptions to the requirements of Appendix J will 
require approval by the NRC before implementation. The proposed 
change in itself does not affect reactor operations and does not 
change radiological consequences.
    Therefore, this proposed change does not involve a significant 
increase in the possibility or consequences of an accident 
previously evaluated.
    2. The possibility of an accident or a malfunction of a 
different type than any previously evaluated is not created.
    The proposed TS change request (TSCR) does not involve any 
physical changes to the plant, affect the operation of the plant, or 
change testing methods or acceptance criteria. The history of 
containment testing verifies that containment integrity has been 
maintained.
    The scheduler change that is proposed should not significantly 
decrease the level of confidence in the ability of the reactor 
building to limit offsite doses to allowable values. No accident or 
malfunction can be the result of the change in test schedule or 
frequency.
    Since the proposed TSCR will not directly impact equipment, 
procedures or operations, the changes will not create the 
possibility of any new or different kind of accident from any 
previously evaluated.
    3. The margin of safety has not been significantly reduced.
    The reason for performing ILRTs [integrated leakage rate tests] 
is to assure that the leakage paths are identified, and any accident 
release will be restricted to those paths assumed in
    the safety analysis. The purpose for the schedule is to assure 
that containment integrity is verified on a periodic basis.
    Revising the schedule does not mean that containment integrity 
will be compromised. Type B and C testing will still be performed. 
The requirements in 10 CFR 50 Appendix J still require the testing 
to be performed periodically.
    The testing previously performed has shown that acceptable 
results were obtained. The ILRT results minus the LLRT results 
demonstrate that most of the increases in leakage are the result of 
LLRT increases. These changes in Type B and C leakage are tracked 
and corrective action is initiated at a specific action level.
    Therefore, the margin of safety has not been significantly 
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: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
    NRC Project Director: Frederick J. Hebdon

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: June 19, 1995
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to change the required test 
frequency for the reactor building spray nozzle flow test from once per 
five years to once per ten years.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The probability or consequences of an accident previously 
evaluated is not significantly increased.
    This change does not effect the probability or consequences of 
an accident. The Reactor Building Spray System is normally idle, 
with the exception of testing. The possibility of the introduction 
of foreign material or corrosion products to restrict flow is 
minimized because of the use of 304 stainless steel as construction 
material. This change results in an extension of the testing 
periodicity only.
    2. The possibility of an accident or a malfunction of a 
different type than any previously evaluated is not created.
    This change results in an extension of the testing periodicity 
only and does not result in an accident not previously evaluated.
    3. The margin of safety has not been significantly reduced.
    The Reactor Building Spray System is normally idle, with the 
exception of testing. The possibility of the introduction of foreign 
material or corrosion products to restrict flow is minimized because 
of the use of 304 stainless steel as construction material. Industry 
wide spray system reliability, as demonstrated by the performance of 
these tests, justifies this change in the frequency of the nozzle 
flow test. This results in an extension of the testing periodicity 
only and will not significantly reduce 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: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
    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: June 29, 1995 (TS 95-14)
    Description of amendment request: The proposed change would, under 
certain stated administrative controls, allow both sets of containment 
personnel airlock doors to be open during core alterations and fuel 
movements. The administrative controls that would be added to Limiting 
Condition for Operation 3.9.4.b would allow both airlock doors to be 
open if one personnel airlock door in each airlock is capable of 
closure, and one train of the Auxiliary Building Gas Treatment System 
is operable in accordance with Specification 3.9.12. In addition, 
proposed changes to Surveillance Requirement 4.9.4 and 4.9.4.a would 
replace the requirement to determine that the containment building 
penetrations are in the ``closed/isolated'' condition with the need to 
determine that they are in the ``required'' condition, and delete the 
requirement to verify that the penetrations are in their required 
condition and the requirement to test the Containment Ventilation 
isolation valves ``within 100 hours prior to the start of'' core 
alterations or movement of irradiated fuel in the containment building. 
Related changes to the Bases would supply amplifying information.
    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 

[[Page 37101]]
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 to TS 3.9.4, Containment Building 
Penetrations, would allow the containment personnel airlocks (PALs) 
to be open during fuel movement and core alteration. The PALs are 
not an initiator to any accident. The position of the PAL doors 
(open or closed) during fuel movement and core alterations has no 
affect on the probability of any accident previously evaluated.
    All doses from a fuel handling accident (FHA) for the proposed 
change remain well below the 10 CFR 100 limits. The proposed change 
will reduce the dose to workers inside containment in the event of a 
FHA by allowing more rapid egress from containment. The wear on the 
PAL doors will significantly be decreased; therefore, increasing the 
reliability of the PAL doors in the event of an accident.
    Since the probability of a FHA is not affected by the airlock 
door positions, and the doses remain within acceptable limits, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    As previously stated, the PAL doors are not accident initiators. 
The open PAL doors do not represent a significant change in the 
configuration of the plant; therefore, does not create a new or 
different type of accident from any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The margin of safety provided for an FHA inside containment 
remains well below the 10 CFR 100 limits. Therefore, this proposed 
change to allow the PAL doors to remain open during fuel movement or 
core alterations does not involve a significant reduction in the 
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

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

    Date of amendment request: September 2, 1992
    Description of amendment request: The proposed amendment revises 
the surveillance criteria for the source range monitors (SRMs) to 
incorporate a more conservative signal-to-noise (S/N) ratio, as 
recommended by General Electric for this system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration. The 
NRC staff has reviewed the licensee's analysis against the standards of 
10 CFR 50.92(c). The NRC staff's review is presented below:
    1.
    The proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The SRM instrumentation is not assumed to be an initiator of any 
analyzed event. The SRM instrumentation provides monitoring of neutron 
flux levels to give the control room operator early indication of 
unexpected subcritical multiplication that could be indicative of an 
approach to criticality. As such, action could be taken on the 
indication to avert or minimize the consequences of the event. However, 
the SRM function is not relied upon in any design bases or transient 
analysis. Rod motion interlocks and other instrumentation are relied on 
in the accident analysis to avert an accident. The change in acceptable 
count rate and signal-to-noise ratio preserves the confidence level of 
the General Electric design. As a result, the consequences of any 
analyzed events are unaffected because the change does not alter any 
system or component design assumptions or operation. Therefore, no 
significant increase in the probability or consequences of an accident 
previously evaluated will be involved.
    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 in SRM count rate and S/N ratio values does not 
change modes of plant operation or require physical modifications. The 
WNP-2 design basis accident and transient analyses do not rely on the 
SRMs to assume plant safety. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident.
    3.
    The proposed change does not involve a significant reduction in a 
margin of safety.
    The proposed change does not involve a significant reduction in a 
margin of safety. The design basis to assure SRM operability is based 
on an instrument count rate that will assure the SRMs will provide 
early indication of subcritical multiplication with a 95-percent 
confidence level. Requiring the count rate to be greater than or equal 
to 0.7 counts per second (cps) with a S/N ratio greater than or equal 
to 20, or greater than or equal to 3 cps with a S/N ratio greater than 
or equal to 2 (vs. a count rate of greater than or equal to 0.5 cps 
with a S/N ratio greater than or equal to 2 in current TS) ensures the 
design 95-percent confidence level is maintained when verifying SRM 
operability. Therefore, the margin of safety is not affected by this 
change.
    Based on this review, it appears that the three standards of 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: William H. Bateman

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

    Date of amendment request: June 6, 1995
    Description of amendment request: The proposed amendment would 
change Technical Specification 6.9.3.2. The change would add references 
to three topical reports describing analytical methods that may be used 
in determining reactor core operating limits for reload licensing 
applications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment does not remove or modify existing 
Technical Specification 

[[Page 37102]]
requirements or safety limits. The Technical Specifications will 
continue to require operations within analyzed core operating limits 
and appropriate actions be taken when, or if, limits are exceeded. 
There will be no changes to the physical design of the plant as a 
result of adding the proposed references to Section 6.9.3.2. The 
results of analytical determination of core operating limitations is 
not assumed as the initiator of any analyzed event, and the approved 
safety analysis is still applicable. Therefore, the proposed 
amendment to Technical Specification 6.9.3.2 does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment does not remove or modify existing 
Technical Specification requirements or safety limits. The Technical 
Specifications will continue to require operation within analyzed 
core operating limits and appropriate actions be taken when, or if, 
limits are exceeded. The technical methodology outlined in the three 
new reports is in accordance with the accepted principals, and the 
specific reports proposed for inclusion in the Technical 
Specifications by this request have been previously approved by NRC 
for use at WNP-2 as a basis for core reload analyses. Therefore, the 
proposed amendment to Technical Specification 6.9.3.2 does not 
create the possibility of a new or different type of accident from 
any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Plant safety limits are established through LCOs, limiting 
safety systems settings, and safety limits specified in the 
Technical Specifications. There will be no changes to either the 
physical design of the plant or to any of these settings and limits 
as a result of adding the proposed references to Section 6.9.3.2. 
The ability to mitigate the consequences of all accidents previously 
evaluated will be maintained and nuclear safety is not adversely 
affected because the technical methodology outlined in the three new 
reports is in accordance with accepted principals, and the specific 
reports proposed for inclusion in the Technical Specifications by 
this request have been previously approved by NRC for use at WNP-2 
as a basis for core reload analyses. Therefore, the proposed 
amendment to Technical Specification 6.9.3.2 does not significantly 
reduce any margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: William H. Bateman

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

    Date of amendment request: June 6, 1995
    Description of amendment request: The proposed amendment would 
change the Index of the WNP-2 technical specifications by deleting 
reference to the Bases pages. Consistent with the requirements of 10 
CFR 50.36(a), which states that the Bases shall not become part of the 
technical specifications, the Bases information will be consolidated 
into a controlled plant document.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes are administrative and do not remove or 
modify existing Technical Specification requirements or safety 
limits. There will be no changes to the physical design of the plant 
as a result of the proposed change. The Bases information, per 10 
CFR 50.36(a), is not part of the Technical Specifications and will 
be consolidated into a controlled plant document. Future changes to 
the Bases will be evaluated per 10 CFR 50.59. Therefore, the 
proposed changes to the Technical Specification Index do not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes are administrative and do not remove or 
modify existing Technical Specification requirements or safety 
limits. There will be no changes to the physical design of the plant 
or alteration of any operational practice as a result of the 
proposed change. The Bases information, per 10 CFR 50.36(a), is not 
part of the Technical Specifications and will be consolidated into a 
controlled document. Future changes to the Bases will be evaluated 
under 10 CFR 50.59. Therefore, the proposed changes to the Technical 
Specifications Index do not create the possibility of a new or 
different type of accident.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Plant safety limits are established through LCOs, limiting 
safety system settings, and safety limits specified in the Technical 
Specifications. There will be no changes to either the physical 
design of the plant or to any of these settings and limits as a 
result of modifying the Technical Specification Index. The ability 
to mitigate the consequences of all accidents previously evaluated 
will be maintained and nuclear safety is not impacted. The Bases 
information, per 10 CFR 50.36(a), is not part of the Technical 
Specifications and will be consolidated into a controlled document. 
Future changes to the Bases will be evaluated under 10 CFR 50.59. 
Therefore, the proposed amendment does not significantly reduce any 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: William H. Bateman

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

    Date of amendment request: June 6, 1995
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 6.0, ``Administrative Controls'' 
for WNP-2. Specifically, the changes would (a) reflect Supply System 
titles for senior management throughout TS 6.0, (b) modify the Plant 
Operations Committee (POC) composition to specify members according to 
functional areas rather than by organizational titles (c) replace the 
Plant Manager as the POC Chairman with an individual appointed by the 
Plant General Manager, and (d) make an editorial correction.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a) the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The senior management title changes are title changes only and 
will not impact the plant safety responsibilities associated with 
these positions. The removal of the Plant Operations Committee (POC) 
organizational 

[[Page 37103]]
titles and replacement with functional areas, and the elimination of 
the Plant Manager as the POC Chairman, will not impact the POC 
function because membership qualifications will continue to be 
consistent with the unit staff qualifications in TS 6.3.1 for those 
POC members and alternates considered part of the unit staff. Those 
designated POC members and alternates not considered part of the 
unit staff will possess skills and knowledge commensurate with their 
organizational positions. The proposed change ensures that POC will 
continue to be comprised of personnel who are experienced, have 
varied expertise, and are involved in daily plant activities. In 
maintaining the qualification requirements for members of POC, the 
POC will continue to fulfill its review and advisory 
responsibilities specified in TS 6.5.1.6 and TS 6.5.1.7. The 
proposed changes do not involve any physical changes to plant 
systems, structures, or components (SSC) or the manner in which the 
SSC are operated, maintained, modified, tested, or inspected. The 
changes therefore do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Because the proposed changes are of an organizational nature and 
their implementation does not involve physical changes to the plant 
SSC or the manner in which the SSC are operated and maintained, the 
proposed changes do not create the possibility of a new or different 
kind of accident. The proposed changes do not introduce any new 
modes of operation or alter system setpoints which could create a 
new or different kind of accident. Therefore, the possibility of a 
new or different kind of accident from any accident previously 
evaluated is not created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The senior management title changes do not impact the management 
responsibilities or functions associated with ensuring plant safety. 
Changes proposed in the POC composition will allow the scope of 
available expertise to be expanded without changing the POC function 
or responsibilities. Maintaining the current level of personnel 
qualifications and experience ensures the POC will continue to meet 
its TS review and advisory requirements. The proposed changes will 
not impact the basis for any Technical Specification related to the 
establishment of, or maintenance of, nuclear safety margins. 
Therefore, operation of the facility in accordance with the proposed 
amendment does not involve a significant reduction in a margin of 
safety.
    The NRC staff has revieywed 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: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: William H. Bateman

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: March 28, 1994
    Brief description of amendments: The amendments change the minimum 
condensate storage tank indicated level from 25 feet to 29.5 feet to 
ensure that the condensate storage tank contains a sufficient volume of 
water. In addition, an editorial change was made to Technical 
Specification 3.7.1.3 for Unit 3 to be consistent with Units' 1 and 2 
technical specifications.
    Date of issuance: July 6, 1995
    Effective date: July 6, 1995
    Amendment Nos.:  Unit 1 - Amendment No. 94; Unit 2 - Amendment No. 
82; Unit 3 - Amendment No. 65
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29625) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 6, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.

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: April 6, 1995, as supplemented 
by letter dated June 7, 1995.
    Brief description of amendments: These amendments involve 
improvements delineated in Generic Letter 93-07, ``Modification of the 
Technical Specification Administrative Control Requirements for 
Emergency and Security Plans,'' changes in plant review board, and 
miscellaneous minor changes.
    Date of issuance: July 7, 1995
    Effective date: July 7, 1995
    Amendment Nos.:  Unit 1 - Amendment No. 95; Unit 2 - Amendment No. 
83; Unit 3 - Amendment No. 66
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 23, 1995 (60 FR 
27335) The June 7, 1995, letter provided clarifying information and did 
not change the initial no sigificant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 7, 1995.No 

[[Page 37104]]
significant hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.

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

    Date of application for amendments: February 16, 1993, as 
supplemented by letter dated May 2, 1995.
    Brief description of amendments: This application upgrades the 
current custom Technical Specifications (TS) for Dresden and Quad 
Cities to the Standard Technical Specification contained in NUREG-0123, 
``Standard Technical Specification General Electric Plants BWR/4.'' 
This application upgrades only Section 3/4.10 (Refueling Operations).
    Date of issuance: June 23, 1995
    Effective date: Immediately, to be implemented no later than 
December 31, 1995, for Dresden Station and June 30, 1996, for Quad 
Cities Station.
    Amendment Nos.: 136, 130, 157, and 153
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 23, 1995 (60 FR 
27337) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 23, 1995. No significant 
hazards consideration comments received: No Local Public
    Document Room location: for Dresden, Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities, 
Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021.

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

    Date of application for amendments: December 23, 1994
    Brief description of amendments: The amendments revise the 
Technical Specifications by increasing the allowable U-235 enrichment 
of fuel to be stored in the new fuel storage vault.
    Date of issuance: June 22, 1995
    Effective date: June 22, 1995
    Amendment Nos.: 164 and 152
    Facility Operating License Nos. DPR-39 and DPR-48: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8742) The Commission's related evaluation of the amendments is 
contained in an Environmental Assessment dated June 8, 1995, and a 
Safety Evaluation dated June 22, 1995. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station Units 1 and 2, Lake County, Illinois
    Date of application for amendments: March 24, 1995
    Brief description of amendments: The amendments recognize 
performing containment leakage rate tests in accordance with 10 CFR 
Part 50, Appendix J, and approved exemptions.
    Date of issuance: June 30, 1995
    Effective date: June 30, 1995
    Amendment Nos.: 165 and 153
    Facility Operating License Nos. DPR-39 and DPR-48: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20516) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 30, 1995. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, 
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport, 
Pennsylvania
    Date of application for amendments: June 8, 1993, as supplemented 
June 15, 1995
    Brief description of amendments: These amendments revise item 2 of 
Technical Specification 6.9.1.14, ``Core Operating Limits Report,'' for 
Unit 1 and Unit 2, to specify the use of the BASH methodology instead 
of an earlier Westinghouse methodology. The BASH methodology is a 
Westinghouse improved and updated methodology which can be used to 
evaluate a large break loss-of-coolant accident. The BASH methodology 
was approved by the NRC staff on November 13, 1986.
    Date of issuance: June 27, 1995
    Effective date: As of the date of issuance, to be implemented 
within 60 days
    Amendment Nos.: 189 and 71
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 7, 1993 (58 FR 
36433) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 27, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam 
Electric Station, Unit 3, St. Charles Parish, Louisiana
    Date of amendment request: August 9, 1994
    Brief description of amendment: The amendment changed the Appendix 
A Technical Specifications (TSs) by revising the Administrative 
Controls Section of the TSs for Waterford 3 by removing the functions 
under review and audit from the TSs and by relocating those items in 
the quality assurance program manual. In addition the amendment removed 
the review and audit functions for the emergency plan and implementing 
procedures, and security plan from the list of responsibilities of the 
plant operation review committee in the TSs. These requirements will be 
retained in emergency plan or security plan as appropriate.
    Date of issuance: July 6, 1995

Effective date: July 6, 1995, to be implemented within 60 days.

    Amendment No.: 109
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 14, 1994 (59 
FR 47167) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 6, 1995.No significant 
hazards consideration comments received: No. Local Public Document Room 
location: University of New Orleans Library, Louisiana Collection, 
Lakefront, New Orleans, LA 70122.

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

    Date of application for amendments: April 3, 1995
    Brief description of amendments: These amendments will incorporate 
line-item TS improvements to Specifications 3/4.8.1 ``Electrical Power 
Systems-A.C. Sources,'' and 4.8.1.2.2 ``Electrical Power Systems-
Shutdown.'' The changes are consistent with 

[[Page 37105]]
recommendations for Emergency Diesel Generator (EDG) Surveillance 
Requirements in NUREG-1366, and regulatory guidance provided in Generic 
Letter (GL) 93-05 and GL 94-01. This issuance also contains FPL's 
commitment to implement a maintenance program for monitoring and 
maintaining EDG performance for both St. Lucie Units consistent with 10 
CFR 50.65 and the guidance of Regulatory Guide 1.160.
    Date of Issuance: June 29, 1995
    Effective Date: June 29, 1995
    Amendment Nos.: 138 and 78
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24910) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 29, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

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

    Date of application for amendments: November 19, 1995
    Brief description of amendments: The amendments relocate the 
requirements of Technical Specification 3/4.3.4, Turbine Overspeed 
Protection, to Section 16.3 of the Vogtle Final Safety Analysis Report 
(FSAR). In addition, the surveillance intervals for exercising the high 
pressure turbine stop valves, the low pressure turbine intermediate 
stop valves and intercept valves, and the high pressure turbine control 
valves are extended after relocation to the FSAR.
    Date of issuance: July 3, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 88 and 66
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the technical Specifications.
    Date of initial notice in Federal Register: February 16, 1994 (59 
FR 7689) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 3, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830.

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

    Date of application for amendments: December 27, 1994
    Brief description of amendments: The amendments revise the 
frequency of conducting leak testing of containment purge valves with 
seals made of resilient material from every 3 months to each refueling 
outage.
    Date of issuance: July 7, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 89 and 67
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 1995 (60 FR 
6301) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 7, 1995. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830.

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

    Date of amendment request: February 15, 1995.
    Brief description of amendments: The amendments modified (by 
relocation to the Technical Requirements Manual) Technical 
Specification (TS) 3/4.3.3.7, Chemical Detection Systems, and TS 3/
4.8.4.1, Electrical Equipment Protective Devices - Containment 
Penetration Conductor Overcurrent Protective Devices, and the 
associated Bases.
    Date of issuance: July 6, 1995
    Effective date: July 6, 1995, to be implemented within 30 days.
    Amendment Nos.:  Unit 1 - Amendment No. 76; Unit 2 - Amendment No. 
65
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16189) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 6, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

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

    Date of application for amendment: March 28, 1995
    Brief description of amendment: The amendment revises Technical 
Specifications (TS) Table 3.2-A by clarifying or correcting entries to 
the table. The amendment also revises the TS Bases to describe more 
clearly the logic arrangements in Table 3.2-A.
    Date of issuance: June 14, 1995
    Effective date: June 14, 1995
    Amendment No.: 212
    Facility Operating License No. DPR-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20519) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 14, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S. E., Cedar Rapids, Iowa 52401.

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

    Date of application for amendments: March 31, 1995
    Brief description of amendments: The amendments modify the 
Containment Ventilation System Technical Specifications (and associated 
Bases) to allow limited containment purge operation in Modes 1, 2, 3, 
and 4 for pressure control, ALARA [as low as is reasonably achievable], 
and respirable air quality considerations.
    Date of issuance: June 23, 1995
    Effective date: June 23, 1995
    Amendment Nos.: 195 and 181
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20520) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 23, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

[[Page 37106]]


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

    Date of application for amendment: March 17, 1995
    Brief description of amendment: The amendment allows a one-time 
extension of the required test interval for the overall integrated 
containment leak rate test (Type A test). This extension allows the 
third Type A test of the second 10-year service period to be performed 
during the refueling outage that will follow the end of Cycle 15. 
Concurrently, the Commission has also granted a one-time schedular 
exemption to allow an extension of one cycle for the performance of the 
10 CFR Part 50, Appendix J, Type A test.
    Date of issuance: July 6, 1995
    Effective date: July 6, 1995
    Amendment No.: 196
    Facility Operating License No. DPR-58. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20519) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 6, 1995.No significant 
hazards consideration comments received: No. Local Public
    Document Room location: Maud Preston Palenske Memorial Library, 500 
Market Street, St. Joseph, Michigan 49085.

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

    Date of application for amendments: December 5, 1994, as 
supplemented January 9, 1995 and May 15, 1995.
    Date of application for amendments: The amendments revise the 
Prairie Island Technical Specifications to allow containment airlock 
doors to remain open during core alterations provided certain 
conditions are met. In its May 15, 1995, letter, the licensee withdrew 
the portion of its original application which dealt with containment 
penetrations during core alterations. The staff granted the licensee's 
request to withdraw all aspects of its application concerning the 
opening of containment penetrations during core alterations.
    Date of issuance: July 3, 1995
    Effective date: July 3, 1995
    Amendment Nos.: 119/112
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 1995 (60 FR 
6306). The January 9 and May 15, 1995, letters provided updated 
Technical Specification pages and clarifying information in response to 
discussions with the staff during various teleconferences conducted 
during the review process. This information was within the scope of the 
original application and did not change the staff's initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated July 3, 1995. No Significant hazards consideration comments 
received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

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

    Date of amendment request: February 10, 1995
    Brief description of amendment: The amendment relocates the 
requirements for the incore instrumentation (ICI) system from the 
technical specifications to the Updated Safety Analysis Report (USAR).
    Date of issuance: June 26, 1995
    Effective date: June 26, 1995
    Amendment No.: 167
    Facility Operating License No. DPR-40. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 15, 1995 (60 FR 
14025) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 26, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.

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

    Date of application for amendments: April 19, 1995 (LAR 95-03)
    Brief description of amendments: The amendments would allow an 
emergency diesel generator (EDG) hot restart test within 5 minutes of a 
2-hour run at the continuous rating instead of an EDG loss of offsite 
power load sequencing test within 5 minutes of the 24-hour endurance 
run.
    Date of issuance: June 26, 1995
    Effective date: June 26, 1995
    Amendment Nos.: Unit 1 - Amendment No. 105; Unit 2 - Amendment No. 
104
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 23, 1995 (60 FR 
27340) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 26, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

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

    Date of application for amendments: February 16, 1994, as 
supplemented by letter dated April 25, 1995 (Reference LAR 94-05)
    Brief description of amendments: The amendment revises Technical 
Specifications 3/4.7.2, ``Steam Generator Pressure/Temperature 
Limitation,'' 3/4.7.7, ``Snubbers,'' 3/4.7.8, ``Sealed Source 
Contamination,'' 3/4.7.11, ``Area Temperature Monitoring,'' and 3/
4.7.13, ``Flood Protection,'' in accordance with the Commission's final 
policy statement for relocation of current technical specifications to 
licensee controlled documents that do not satisfy any of the policy 
statement criteria.
    Date of issuance: July 6, 1995
    Effective date: July 6, 1995
    Amendment Nos.: Unit 1 - Amendment No. 106; Unit 2 - Amendment No. 
105
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17603) The April 25, 1995, supplemental letter provided additional 
clarifying information and did not change the initial no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
July 6, 1995.No significant hazards consideration comments received: 
No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

[[Page 37107]]


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: February 22, 1995
    Brief description of amendment: The amendment modifies operability 
and surveillance requirements for the reactor vessel overfill 
protection instrumentation that initiates feedwater pump turbine and 
main turbine trips on high reactor vessel water level. The NRC staff 
has determined that the proposed Technical Specification (TS) changes 
will have no adverse impact on plant safety and will enhance the 
current TSs by adding operability requirements for the reactor vessel 
overfill protection system. Therefore, the proposed TS changes are 
acceptable.
    Date of issuance: June 19, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 225
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24915) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 19, 1995.No significant 
hazards consideration comments received: No Local Public Document Room 
location: Reference and Documents Department, Penfield Library, State 
University of New York, Oswego, New York 13126.

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

    Date of application for amendment: July 28, 1994 and December 15, 
1994
    Brief description of amendment: This amendment makes changes to TS 
Section 3/4.8.1 ``AC SOURCES.'' The staff found it appropriate to 
combine these two applications into one amendment. The amendment 
removes the surveillance requirements, methodology and frequency for 
Emergency Diesel Generator (EDG) fuel oil from the TS and relocates 
them in a controlled plant procedure, VSH.SS-CA.ZZ-0013(Q) ``Procedures 
for Testing Diesel Fuel and 2 Fuel Oil at Artificial Island 
for PSE&G Nuclear Operations.'' The changes also delete an unnecessary 
lab test for the fuel oil and extend the surveillance frequency from 
once per 92 days to once per 184 days. In addition and in accordance 
with 10 CFR 50.90, this amendment removes TS Surveillance Requirement 
4.8.1.1.2.h.1 in order that PSE&G can utilize plant-controlled programs 
to govern diesel generator maintenance. To ensure procedural 
consistency and reduce the impact of this change on Hope Creek 
procedures, the remaining Surveillance Requirements of TS 4.8.1.1.2.h 
are not renumbered.
    Date of issuance: June 29, 1995
    Effective date: June 29, 1995
    Amendment No.: 74
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 
45034) and April 26, 1995 (60 FR 20526) The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
June 29, 1995.No significant hazards consideration comments received: 
No
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070.

Public Service Electric & Gas Company, Docket No. 50-272, Salem 
Nuclear Generating Station, Unit No. 1, Salem County, New Jersey 
Date of application for amendment: April 4, 1995

    Brief description of amendment: The amendment allows a one-time 
interval extension for the Type A test required by 10 CFR Part 50, 
Appendix J. Instead of conducting the test during the twelfth refueling 
outage, it can now be conducted during the thirteenth refueling outage, 
but no later than June 1997.
    Date of issuance: July 5, 1995
    Effective date: July 5, 1995
    Amendment No.: 171
    Facility Operating License No. DPR-70: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 23, 1995 (60 FR 
27341) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 5, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.

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

    Date of application for amendments: April 6, 1995; supplemented May 
26, 1995 (TS 94-19)
    Brief description of amendments: The amendments revise action 
statements to eliminate starting of emergency diesel generators in 
order to verify their operability whenever one of the required 
electrical power sources is inoperable or a diesel is inoperable unless 
the diesel inoperability is due to a common cause failure.
    Date of issuance: June 29, 1995
    Effective date: June 29, 1995
    Amendment Nos.: 205 and 195
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20529) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 29, 1995.No significant 
hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library,1101 Broad Street, Chattanooga, Tennessee 37402.

The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania 
Power Company, Toledo Edison Company, Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio Date of 
application for amendment: September 27, 1993 and December 16, 1994

    Brief description of amendment: The amendment revised the Technical 
Specification Section 6.8.1, ``Unit Staff Qualifications,'' to make it 
consistent with the current requirements of Part 55 of Title 10 of the 
Code of Federal Regulations.
    Date of issuance: June 27, 1995
    Effective date: June 27, 1995
    Amendment No.: 70
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64604) and February 1, 1995 (60 FR 6310). The Commission's related 
evaluation of the amendment is contained in an Environmental Assessment 
dated February 28, 1995, and a Safety Evaluation, dated June 27, 
1995.No significant hazards consideration comments received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.

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

    Date of application for amendment: July 16, 1993
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) 3/4.8.1.1 and 3/

[[Page 37108]]
4.8.1.2. The changes address the minimum required storage volumes of 
the Emergency Fuel Oil storage and day tanks.
    Date of issuance: July 6, 1995Effective date: July 6, 1995
    Amendment No.: 100
    Facility Operating License No. NPF-30: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17607) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 6, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

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

    Date of application for amendments: January 24, 1995
    Brief description of amendments: The amendments change the ``as-
found'' test criterion for the pressurizer safety valves from plus or 
minus 1% to plus or minus 3%
    Date of issuance: June 29, 1995
    Effective date: June 29, 1995
    Amendment Nos.: 200 and 200
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 12, 1995 (60 FR 
18631) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 29, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.

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

    Date of application for amendment: July 12, 1994
    Brief description of amendment: The amendment modifies the 
technical specifications (TS) to remove instrument response time limit 
tables for the reactor protection system, isolation actuation system, 
and emergency core cooling system from the TS. The affected instrument 
response time limit tables will be located in the Final Safety Analysis 
Report (FSAR).
    Date of issuance: June 26, 1995
    Effective date: June 26, 1995, to be implemented within 30 days of 
issuance.
    Amendment No.: 139
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 
45036). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 26, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352.

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

    Date of application for amendments: January 24, 1995, as 
supplemented by letters dated February 24, April 25, May 24, and June 
1, 1995
    Brief description of amendments: These amendments revise Point 
Beach Nuclear Plant Technical Specification (TS) Section 15.6.5, 
``Review and Audit,'' and TS Section 15.7.8, ``Administrative 
Controls.'' The quality assurance audit frequencies and the section on 
emergency plan reviews are relocated to other documents, and the period 
for radioactive effluent reporting is increased to annual. In addition, 
the references to ``Semiannual Monitoring Report'' are changed to 
``Annual Monitoring Report'' throughout TS Sections 15.7 and 16.5. 
Administrative changes are also included.
    Date of issuance: July 5, 1995
    Effective date: July 5, 1995
    Amendment Nos.: Unit 1 - 162: Unit 2 - 166
    Facility Operating License Nos. DPR-24 and DPR-27. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11142). The February 24, April 25, May 24, and June 1, 1995, submittals 
provided supplemental information that did not change the initial 
proposed no significant hazards consideration determination.The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated July 5, 1995.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.

Notice Of Issuance Of Amendment To Facility Operating License And 
Final No Significant Hazards Consideration Determination

    During the period since publication of the last biweekly notice, 
individual notices of issuance of amendments have been issued for the 
facilities as listed below. These notices were previously published as 
separate individual notices. They are repeated here because this 
biweekly notice lists all amendments that have been issued for which 
the Commission has made a final determination that an amendment 
involves no significant hazards consideration.
    In this case, a prior Notice of Consideration of Issuance of 
Amendment, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing was issued, a hearing was requested, and 
the amendment was issued before any hearing because the Commission made 
a final determination that the amendment involves no significant 
hazards consideration.
    Details are contained in the individual notice as cited.

Commonwealth Edison Company, Docket No. 50-295, Zion Nuclear Power 
Station Unit 1, Lake County, Illinois

    Date of amendment request: May 17, 1995
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to allow 154 steam generator 
tubes that potentially exceed the repair or plugging criteria to remain 
in service for the remainder of the current Unit 1 operating cycle.
    Date of publication of individual notice in Federal Register: May 
25, 1995 (60 FR 27798)
    Expiration date of individual notice: June 26, 1995
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.

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

    Date of amendment request: June 14, 1995
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) to allow the hot restart 
sequence loading test of the emergency diesel generators to be 
performed independent of the 24 hour endurance test.

[[Page 37109]]

    Date of publication of individual notice in Federal Register: June 
30, 1995 (60 FR 34308)
    Expiration date of individual notice: July 31, 1995

    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.

    Dated at Rockville, Maryland, this 12th day of July 1995.

    For the Nuclear Regulatory Commission

Jack W. Roe,
Director, Division of Reactor Projects - III/IV,Office of Nuclear 
Reactor Regulation

[Doc. 95-17565 Filed 7-18-95; 8:45 am]

BILLING CODE 7590-01-F