[Federal Register Volume 60, Number 158 (Wednesday, August 16, 1995)]
[Notices]
[Pages 42597-42622]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-20816]



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NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating LicensesInvolving 
No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from July 21, 1995, through August 4, 1995. The 
last biweekly notice was published on Wednesday, August 2, 1995 (60 FR 
39430).

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

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

[[Page 42598]]
proven, would entitle the petitioner to relief. A petitioner who fails 
to file such a supplement which satisfies these requirements with 
respect to at least one contention will not be permitted to participate 
as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendments request: March 15, 1995, as supplemented on June 
29, 1995.
    Description of amendments request: The proposed amendments would 
revise the Calvert Cliffs Nuclear Power Plant, Units Nos. 1 and 2, 
Technical Specifications (TSs) Section 6, ``Administrative Controls,'' 
to be consistent with the guidance provided in NUREG-1432, ``Standard 
Technical Specifications, Combustion Engineering Plants.'' The proposed 
changes will relocate several requirements to other documents and 
programs consistent with NUREG-1432 and other NRC guidance addressing 
the administrative section of the TSs such as the ``Final Policy 
Statement on Technical Specification Improvements for Nuclear Power 
Reactors,'' published in the Federal Register on July 22, 1993 (58 FR 
39132).
    The Commission indicated that compliance with the Final Policy 
Statement satisfies Section 182a of the Act. In particular, the 
Commission indicated that certain items could be relocated from the TSs 
to licensee-controlled documents, consistent with the standard 
enunciated in Portland General Electric Co. (Trojan Nuclear Plant), 
ALAB-531, 9 NRC 263, 273 (1979). In that case, the Atomic Safety and 
Licensing Appeal Board indicated that ``technical specifications are to 
be reserved for those matters as to which the imposition of rigid 
conditions or limitations upon reactor operation is deemed necessary to 
obviate the possibility of an abnormal situation or event giving rise 
to an immediate threat to the public health and safety.'' The policy 
statement encouraged licensees to adopt the applicable improved STSs 
and provided some guidance for the conversion from the present plant-
specific TSs to the improved Standard TSs.
    The proposed changes will provide significant human factors 
improvement to the TSs by accomplishing the following: (1) relocating 
existing requirements to licensee controlled documents consistent with 
the policy statement; (2) eliminating requirements which duplicate 
regulations; (3) relocating similar requirements within the same 
section; (4) editorial changes; and (5) adding requirements consistent 
with NUREG-1432.
    In addition, the licensee proposes dual rolls for the Shift 
Technical Advisor (STA) and the establishment of a TS Bases Control 
Program. Allowing the STA to perform dual rolls is not permitted by the 
current TSs, but the current NRC guidance allows the STA to perform a 
dual roll. The proposed new TS Bases Control Program will define the 
appropriate methods and reviews required to implement a TS Bases change 
which is also consistent with the current NRC guidance. Two other 
proposed changes, not specifically covered by the above groupings, 
include a reduction in reporting requirements and utilizing a more 
effective option for estimating doses.
    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. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    Relocating existing requirements to Baltimore Gas and Electric 
Company (BGE)-controlled documents, eliminating requirements which 
duplicate regulations, locating similar requirements within the same 
sections and making necessary editorial corrections to incorporate 
the proposed changes provide Technical Specifications which are 
easier to use. Because existing requirements are relocated to 
established BGE programs where changes to those programs are 
controlled by regulatory requirements, there is no reduction in 
commitment and adequate control is still maintained. Likewise, the 
elimination of requirements which duplicate regulations enhances the 
usability of the Technical Specifications without reducing 
commitments. Locating similar requirements within the same sections 
and making necessary editorial corrections to incorporate the 
proposed changes neither add nor delete requirements, but merely 
clarify and improve the readability and understanding of the 
Technical Specifications. Since the requirements remain the same, 
these changes only affect the method of presentation and would not 
affect possible initiating events for accidents previously evaluated 
or any system functional requirement. Therefore, the proposed 
changes would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.

[[Page 42599]]

    Since the Shift Technical Adviser (STA) is not considered an 
initiator to any previously evaluated accident nor considered in the 
accident's response, the use of a dual role STA would not increase 
the probability or consequences of any previously evaluated 
accident.
    The Technical Specification Bases Control Program provides 
controls which ensure appropriate reviews of changes to the Bases. 
Because NRC approval is still needed for changes to the Bases which 
affect the Technical Specifications, the proposed Program would not 
affect the probability or consequences of a previously evaluated 
accident.
    Eliminating the requirement for submitting two reports which 
place unwarranted administrative burden on both Baltimore Gas and 
Electric Company and the NRC has no affect on the probability or 
consequences of an accident previously evaluated. Therefore, the 
proposed changes would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Replacing the film badge with the electronic personal dosimeter 
provides a more effective, efficient, state-of-the art option for 
estimating dose and would not impact accidents previously evaluated. 
Therefore, the proposed change would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    As discussed previously, relocating existing requirements to 
BGE-controlled documents, eliminating requirements which duplicate 
regulations, locating similar requirements within the same sections 
and making necessary editorial corrections to incorporate the 
proposed changes will not affect any plant system or structure, nor 
will it affect any system functional or operability requirements. 
Consequently, no new failure modes are introduced as a result of the 
proposed changes. Therefore, these types of changes would not create 
the possibility of a new or different type of accident from any 
accident previously evaluated.
    Because the STA does not perform equipment design or equipment 
manipulation, the use of a dual role STA would not create the 
possibility of a new or different type of accident from any accident 
previously evaluated. Since the Technical Specification Bases 
Control Program represents an administrative function performed 
under existing regulatory controls, it too would not create the 
possibility of a new or different type of accident from any 
previously evaluated.
    The addition of new programs which incorporate existing 
Technical Specification requirements and commitments will have no 
effect on the design or operation of the plant and would not create 
the possibility of a new or different type of accident from any 
previously evaluated.
    A reporting function such as report submittals would not change 
the configuration or operation of the plant. Consequently, the 
elimination of the requirement to submit the Startup Report and the 
Special Report dealing with iodine activity levels, would not create 
the possibility of a new or different type of accident from any 
accident previously evaluated.
    Since the operation or configuration of the plant is not changed 
by the type of personal dosimeter, this change would not create the 
possibility of a new or different type of accident from any accident 
previously evaluated.
    Therefore, the proposed changes would not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    Relocating existing requirements to BGE-controlled documents, 
eliminating requirements which duplicate regulations, locating 
similar requirements within the same sections and making necessary 
editorial corrections to incorporate the proposed changes would not 
affect the Updated Final Safety Analysis Report design bases, 
accident analysis assumptions or any margin of safety described in 
the Technical Specification Bases. In addition, these proposed 
changes do not affect effluent release limits, monitoring equipment 
or practices. Therefore, these proposed changes would not involve a 
significant reduction in a margin of safety.
    The use of an STA should provide an additional margin of safety 
in the accident response function of licensed operators beyond that 
considered in the accident analysis. Since the STA is required to 
have the same training and educational qualifications in either the 
individual or dual role, the use of a dual role STA should have 
minimal impact. Consequently, the proposed change would not involve 
a significant reduction in a margin of safety. The Technical 
Specification Bases Control Program is an administrative change 
controlling how Technical Specification basis information is 
reviewed and incorporated. Therefore, this change would not involve 
a significant reduction in a margin of safety.
    The addition of new programs which incorporate existing 
Technical Specification requirements and commitments will have no 
effect on the design or operation of the plant and would not result 
in a significant reduction in the margin of safety.
    Activities described in the Startup Report will continue to be 
performed and corrective action taken when required. Similarly, 
iodine activity levels will continue to be monitored and actions 
taken, including the issuance of a Licensee Event Report when 
conditions warrant. Considering the above, elimination of the two 
reporting requirements would have no impact on the margin of safety.
    Plant operating parameters are not affected by the type of 
personnel monitoring device used and as a consequence, would not 
impact a margin of safety. Since the replacement dosimeter provides 
a more effective mechanism for estimating dose, there is no 
degradation in personal safety levels. Consequently, the proposed 
change would not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Ledyard B. Marsh

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 July 28, 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 adoption 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 

[[Page 42600]]
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 July 28, 1995, applications proposed to 
upgrade only Section 3/4.5 (Emergency Core Cooling Systems) of the 
Dresden and Quad Cities TS.
    Basis for proposed no significant hazards 
considerationdetermination: 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 3/4.5 are based on STS guidelines or later 
operating BWR plants' 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 Emergency Core Cooling 
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 3/4.5 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, the addition of 
requirements which are based on the current safety analysis, and 
some 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 3/4.5 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 Emergency Core Cooling 
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 Emergency Core Cooling 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, the addition of 
requirements which are based on the current safety analysis, and 
some minor curtailments of the current requirements which are based 
on generic guidance or previously approved provisions for other 
stations. Some of the latter 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 3/4.5 
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 Emergency 
Core Cooling Systems when required to mitigate accident conditions; 
therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: for Dresden, Morris Public 
Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities, 
Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: Robert A. Capra

[[Page 42601]]


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

    Date of amendment request: June 17, 1993, as supplemented July 5, 
1995
    Description of amendment request: The initial proposed amendment 
request dated June 17, 1993, was previously noticed in the Federal 
Register on July 21, 1993 (58 FR 39048). The proposed amendment would 
revise Technical Specification 5.3.1, ``Fuel Assemblies'' to provide 
flexibility in the repair of fuel assemblies containing damaged and 
leaking fuel rods by reconstituting the assemblies in accordance with 
the guidance in Generic Letter (GL) 90-02, Supplement 1, ``Alternative 
Requirements For Fuel Assemblies In The Design Features Section Of 
Technical Specifications,'' issued on July 31, 1992. The application is 
also generally consistent with the format and content of the improved 
Standard Technical Specifications for Westinghouse plants provided in 
NUREG-1431.
    Additional information was submitted on July 5, 1995, that added TS 
changes to increase the fuel enrichment limit from 4.0 to 5.0 weight 
percent U-235 that were not previously included the initial June 17, 
1993, amendment application. This additional information is being 
noticed to provide for public comment and opportunity for hearing.
    Basis for proposed no significant hazards 
considerationdetermination: As required by 10 CFR 50.91(a), the 
licensee provided its analysis of the issue of no significant hazards 
consideration (58 FR 39048). The NRC staff's analysis of the July 5, 
1995, supplement against the standards of 10 CFR 50.92(c) is presented 
below.
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    There is no increase in the probability or consequences of an 
accident in the new fuel vault since the only accident that would be 
affected by this change would be a criticality accident and it has been 
shown that the worst-case keff under optimum moderation conditions 
continues to be less than or equal to 0.98.
    There is no increase in the probability of a fuel drop accident in 
the Spent Fuel Storage Pool since the mass of an assembly will not be 
significantly affected by the increase in fuel enrichment. The 
likelihood of other accidents, previously evaluated and described in 
Section 9.1.2 of the Final Safety Analysis Report (FSAR), is also not 
affected by the proposed changes. Since the increase in fuel enrichment 
will allow for extended fuel cycles, it could be postulated that there 
may be a decrease in fuel movement and the probability of an accident 
may likewise be decreased. There is also no increase in the 
consequences of a fuel drop accident in the Spent Fuel Pool since the 
fission product inventory of individual fuel assemblies will not change 
significantly as a result of increased initial enrichment. In addition, 
no change to safety-related systems is being made.
    Therefore, the consequences of a fuel rupture accident remain 
unchanged. In addition, it has been shown that keff is less than 
or equal to 0.95, under all conditions. Therefore, the consequences of 
a criticality accident in the Spent Fuel Pool remain unchanged as well.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not create the possibility of a new or 
different kind of accident since fuel handling accidents (fuel drop and 
misplacement) are not new or different kinds of accidents. Fuel 
handling accidents are already discussed in the FSAR for fuel with 
enrichments up to 4.0 weight % and additional analyses have been 
performed for fuel with enrichment up to 5.00 weight %.
    3.
    The proposed changes do not involve a significant reduction in the 
margin of safety.
    The proposed change does not involve a significant reduction in the 
margin of safety since, in all cases, a spent fuel pool keff less 
than or equal to 0.95 is being maintained. Criticality analyses have 
also been performed that show that the new fuel storage vault will 
remain subcritical under a variety of moderation conditions, from fully 
flooded to optimum moderation. As discussed above, the Spent Fuel Pool 
will remain sufficiently subcritical during any fuel misplacement 
accident.
    Based on this analysis, it appears that the three standards of 10 
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the supplemental amendment submittal involves no 
significant hazards consideration.
    Local Public Document Room location:: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: July 26, 1995
    Description of amendment request: The proposed amendments would 
provide a one-time extension of the allowable outage time from 72 hours 
to 7 days. This extension is necessary to implement a modification to 
the degraded grid protection system and the external grid trouble 
protection system.
    Basis for proposed no significant hazards 
considerationdetermination: 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:

Duke Power Company (Duke) has made the determination that this 
amendment request involves a No Significant Hazards Consideration 
by applying the standards established by NRC regulations in 10 CFR 
50.92. This ensures 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:

    Each accident analysis addressed within the Oconee Final Safety 
Analysis Report (FSAR) has been examined with respect to the change 
proposed within this amendment request. The design basis of the 
auxiliary electrical systems is to supply the required engineered 
safeguards (ES) loads of one unit and the safe shutdown loads of the 
other two units. The systems are arranged so that no single failure 
will jeopardize plant safety.
    The probability of any Design Basis Accident (DBA) is not 
significantly increased by this change. In addition, the 
consequences of the accidents are within the bounds of the FSAR 
analyses. The reliability of the emergency power system is not 
significantly affected by a one time extension of allowable outage 
time for the overhead power path. The underground power path is 
adequate to assure operability of the Oconee ES loads. Finally, the 
enhancement of the Degraded [Grid] Protection System will eliminate 
a concern which was expressed by the EDSFI audit team.
    (2) Create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated:
    Inoperability of the yellow bus is functionally equivalent to 
inoperability of the Keowee Main Step-up Transformer in that it 
renders the overhead emergency power path inoperable. The Keowee 
Main Step-up Transformer is allowed to be inoperable for a period 
not to exceed 28 days. This Technical Specification requirement for 
the 

[[Page 42602]]
Keowee Main Step-up Transformer has been reviewed and approved by the 
NRC. Therefore, operation of ONS [Oconee Nuclear Station] in 
accordance with this Technical Specification amendment will not 
create any failure modes not bounded by previously evaluated 
accidents. Consequently, this change will not create the possibility 
of a new or different kind of accident from any kind of accident 
previously evaluated.
    (3) Involve a significant reduction in a margin of safety:
    The design basis of auxiliary electrical systems is to supply 
the required ES loads of one Unit and safe shutdown loads of the 
other two units. The underground power path is adequate to ensure 
operability of the ES loads during the outage of the yellow bus. The 
reliability of the emergency power system is not significantly 
affected by a one time extension of allowable outage time for the 
overhead power path. Therefore, there will be no significant 
reduction in any margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: July 10, 1995
    Description of amendment request: The proposed amendment would 
modify the technical specifications to minimize the potential for boron 
deletion of the reactor coolant system (RCS) during startup of an 
isolated loop. The changes would permit RCS loop isolation only during 
Modes 5 and 6. RCS loop isolation valves would be required open with 
power removed from each isolation valve operator during Modes 1, 2, 3, 
and 4. Primary grade water would be isolated from the RCS during Modes 
4, 5, and 6, except during planned boron dilution or makeup activities.
    Basis for proposed no significant hazards 
considerationdetermination: 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 proposed amendment would modify the method used to prevent 
an inadvertent boron dilution event during hot shutdown, cold 
shutdown and during refueling. An uncontrolled boron dilution 
transient cannot occur during this mode of operation. Inadvertent 
boron dilution is prevented by administrative controls which isolate 
the primary grade water system isolation valves from the Chemical 
and Volume Control System, except during planned boron dilution or 
makeup activities. Thus unborated water can not be injected into the 
reactor coolant system, making an unplanned boron dilution at these 
conditions highly improbable, since the source of unborated water to 
the charging pumps is isolated. This precludes the primary means for 
an inadvertent boron dilution event in this mode of operation.
    The primary grade water system isolation valves may be opened 
when directed by the control room during this mode of operation only 
for a planned boron dilution or makeup activity. The primary grade 
water system isolation valves will be verified to be locked, sealed 
or otherwise secured in the closed position after the planned boron 
dilution or makeup activity is completed. During planned boron 
dilution events, operator attention will be focused on the boron 
dilution process and any inappropriate blender operation will be 
readily identified.
    The operator has prompt and definite indication of any boron 
dilution from the audible count rate instrumentation supplied by the 
source range nuclear instrumentation. High count rate is alarmed in 
the reactor containment and the control room. In addition a high 
source range flux level is alarmed in the control room. The count 
rate increase is proportional to the subcritical multiplication 
factor.
    The proposed amendment would also modify the method used to 
prevent an adverse reactor transient during startup of an isolated 
reactor coolant loop. Procedures require that the isolated loop 
water boron concentration be verified prior to opening loop 
isolation valves. Procedures also require an isolated loop to be 
drained and refilled from water supplied from the Refueling Water 
Storage Tank (RWST) or Reactor Coolant System (RCS) prior to opening 
either the hot or cold leg isolation valves. Using water from the 
RWST or RCS ensures 1) that the boron concentration of the isolated 
loop is sufficient to prevent a dilution of the active reactor 
coolant loops and reducing the shutdown margin to below those values 
used in safety analyses when the isolated loop is returned to 
service, and 2) that no single failure could cause an isolated loop 
to be filled with unborated water.
    Thus procedures and interlocks prevent inadvertent opening of 
loop isolation valves and require that the startup of an isolated 
loop be performed in a controlled manner that virtually eliminates 
any sudden positive reactivity addition from boron dilution. Thus 
the core cannot be adversely affected by the startup of an isolated 
loop and fuel design limits are not exceeded. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not create the possibility of a new or 
different kind of accident. No new systems, structures or components 
are being proposed. Acceptable alternative administrative controls 
are being proposed to address inadvertent boron dilution and the 
startup of inactive reactor coolant loops.
    The primary source of unborated water will be isolated from 
injecting by the charging pumps into the reactor coolant system 
during hot shutdown, cold shutdown, and refueling, except for 
planned boron dilution events and makeup activities. The proposed 
administrative controls prevent the possible accident previously 
evaluated, i.e., an inadvertent boron dilution event.
    A currently installed interlock to recirculate reactor coolant 
in an isolated loop is proposed to be deleted. In its place, each 
reactor coolant isolated loop will be drained and refilled with 
water supplied from the RWST just before the loop is returned to 
service. This administrative control will prevent any inadvertent 
reactivity transient when returning the loop to service. Thus, the 
proposed administrative controls will prevent the type of accident 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes will continue to ensure that adequate 
protection is provided against an inadvertent boron dilution and the 
adverse effects from the startup of an isolated reactor coolant 
loop. General Design Criteria 10 requirements will not be exceeded 
with respect to demonstrating specified acceptable fuel design 
limits. The required indications and functions are still maintained 
in accordance with current technical specification requirements and 
the shutdown margin is unaffected. Therefore, the proposed change 
will not involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. Library, 663 Franklin 
Avenue, Aliquippa, Pennsylvania 15001.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

[[Page 42603]]


Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley 
Power Station, Unit 1, Shippingport, Pennsylvania

    Date of amendment request: July 11, 1995
    Description of amendment request: The proposed amendment would 
revise the required area of the Reactor Coolant System (RCS) 
overpressure protection system vent from 3.14 square inches to 2.07 
square inches. This vent is provided to relieve a potential RCS 
overpressure condition if the power-operated relief valves (PORVs) are 
not operable. The proposed vent area is equal to the relief area of a 
PORV. A single PORV is capable of providing sufficient relief capacity 
to mitigate potential low temperature overpressurization events.
    Basis for proposed no significant hazards 
considerationdetermination: 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 proposed change is considered to be editorial since it 
replaces the 3.14 square inch vent size stated in overpressure 
protection system (OPPS) Specifications 3.4.9.3, 3.1.2.1.b, and 
3.1.2.3 and Bases 3/4.1.2 and 3/4.4.9 with a 2.07 square inch vent 
size. This ensures the vent size stated in the technical 
specifications is consistent with the actual size of an installed 
PORV. These changes maintain consistency with the analyses 
assumptions and the operation of the OPPS in accordance with 
applicable analyses and the UFSAR [Updated Final Safety Analyses 
Report]. Therefore, we have concluded that these changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated in the UFSAR.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not involve any physical changes to the 
OPPS or their setpoints. These changes do not change any function 
previously provided by the OPPS. These changes do not affect any 
failure modes defined for any plant system or component important to 
safety nor has any new limiting single failure been identified as a 
result of these changes. Therefore, these changes will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated in the UFSAR.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes will not affect the operation of or the 
reliability of the OPPS. These changes do not affect the manner in 
which the plant is operated or involve a change to equipment or 
features that affect the operational characteristics of the plant. 
Therefore, operation of the plant in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: July 20, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3/4.8.1.1 to incorporate guidance 
provided in NRC Generic Letter (GL) 84-15, ``Proposed Staff Actions to 
Improve and Maintain Diesel Generator Reliability,'' and GL 93-05, 
``Line-Item Technical Specification Improvements To Reduce Surveillance 
Requirements For Testing During Power Operation,'' which includes (1) 
revised requirements for testing the operable emergency diesel 
generators (EDGs) for various combinations of inoperable offsite 
circuits and EDGs and (2) revised surveillance requirements for the 
EDGs. The revised surveillance requirements include specifying 
generator voltage, frequency limits, and diesel starting time. In 
addition, several editorial changes would be made to TS 3/4.8.1.1 which 
would be consistent with the guidance provided in the NRC's Improved 
Standard Technical Specifications (NUREG-1431).
    Basis for proposed no significant hazards 
considerationdetermination: As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The probability of occurrence of a previously evaluated accident 
is not increased because the allowable outage times for the offsite 
circuits and diesel generators remain unchanged. The consequences of 
an accident previously evaluated is not increased because reducing 
the diesel generator test frequency and permitting additional test 
evolutions are intended to minimize diesel wear and mechanical 
stress. By eliminating excessive testing, which can lead to 
premature diesel failures and minimizing diesel wear and mechanical 
stress, the diesel generator reliability is increased. The 
consequences of an accident previously evaluated is also not 
increased because the addition of the parameters for generator 
voltage, frequency, and diesel starting time to the surveillance 
requirement will provide additional assurance that the diesel 
generators are performing as assumed in the safety analysis. This 
proposed change does not affect the availability or reliability of 
the offsite circuits.
    Therefore, this change will not increase the probability or 
consequences of an accident previously evaluated due to the 
continued availability and reliability of the A.C. electrical power 
sources.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not alter the method of operating the 
plant. The changes do not introduce any new failure modes and are 
intended to increase the diesel generator reliability and provide 
additional assurance that the diesels are performing as assumed in 
the safety analysis. The revision to the various action statements 
and surveillance requirements provide assurance that the diesel 
generators will be able to power their respective safety systems if 
required. The proposed changes do not impact the performance of any 
safety system.
    Therefore, this proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The margin of safety is not reduced because the A.C. electrical 
power sources will continue to provide sufficient capacity, 
capability, redundancy, and reliability to ensure availability of 
necessary power to engineered safety feature (ESF) systems. The ESF 
systems will continue to function, as assumed in the safety 
analyses, to ensure that the fuel, reactor coolant system and 
containment design limits are not exceeded. The elimination of 
excessive testing on the diesel generators are permitting additional 
test evolutions, which result in less diesel wear and mechanical 
stress, are intended to increase diesel reliability. The increased 
reliability of the diesels adds to the ability of the A.C. 
electrical power source to provide power to ESF systems. The 
proposed additions to the surveillance requirements will provide 
additional assurance of the ability of the A.C. electrical power 
sources to provide power to ESF systems.
    Therefore, this proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 

[[Page 42604]]
amendment request involves no significant hazards consideration.
    Local Public Document Room Location:  B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley 
PowerStation, Unit 2, Shippingport, Pennsylvania

    Date of amendment request: July 24, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3/4.4.11, ``Relief Valves,'' and 
associated Bases to make Unit 2 TS 3/4.4.11 consistent with Unit 1 TS 
3/4.4.11, which was revised by Unit 1 License Amendment No. 187 issued 
on May 15, 1995. The proposed amendment would also generally reflect 
the guidance provided in NRC Generic Letter 90-06 and in the NRC's 
Improved Standard Technical Specifications (NUREG-1431).
    Basis for proposed no significant hazards 
considerationdetermination: 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?
    Implementation of these changes will increase the availability 
of the power-operated relief valves (PORVs) and their associated 
block valves. The increased availability is obtained through 
maintaining power to the block valves which are closed to control 
PORV seat leakage. Maintaining power to the block valve provides the 
flexibility of reopening the valves to control reactor coolant 
system pressure. The proposed change modifies Specification 3.4.11 
actions, a surveillance requirement, and Bases to generally reflect 
the requirements of Generic Letter (GL) 90-06, and the guidance 
provided in NUREG-1431, ``Improved Standard Technical 
Specifications'' (ISTS) and is consistent with the changes the NRC 
approved for Unit No. 1. A revised stress analysis has been 
completed that takes credit for the speed at which the block valve 
opens when manually reducing reactor coolant system pressure. The 
block valve relatively slow opening speed reduces the peak pressure 
surge and results in acceptable downstream piping stress values. The 
PORV downstream piping has been evaluated assuming manual vent path 
operation with cold loop seal slug flow and it has been determined 
that the piping supports can accept these design transient loads. 
The proposed change to the action statement to close the block valve 
to isolate a PORV and maintain power to the block valve does not 
significantly increase the probability of a small break loss of 
coolant accident. No PORV function has been deleted and the PORV and 
block valve continue to be capable of being manually closed at any 
time. As a result of the change to action ``a,'' an exception to the 
stroking requirements is no longer required, therefore, reference to 
action ``a'' in Surveillance Requirement 4.4.11.2 has been deleted. 
Closing the block valve for a PORV that is not capable of being 
manually cycled and removing power to the block valve assures that 
the valve will not be inadvertently opened when the condition of the 
PORV is uncertain.
    The changes remain consistent with the analysis assumptions 
regarding the operation of the PORVs and block valves and provides 
increased assurance of their availability in mitigating the 
consequences of a steam generator tube rupture (SGTR) accident. The 
requirements of GL 90-06 are substantially addressed in the ISTS 
which have been incorporated here except for specific design 
differences. Minor editorial changes involving capitalization have 
been incorporated to maintain the format and content and do not 
affect any of the requirements, the accident analyses, or the 
operation of the plant. Therefore, we have concluded that these 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated in the UFSAR 
[Updated Final Safety Analysis Report].
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes to the action statements for the PORVs and 
the associated block valves will improve the availability of these 
valves for normal operation and for mitigation of a SGTR accident. 
The proposed changes do not involve any physical changes to the 
PORVs or their setpoints. These changes do not delete any design 
basis accident function previously provided by the PORV vent path 
nor has the probability of inadvertent opening been increased. 
Accordingly, no new limiting single failure has been identified as a 
result of these changes. Therefore, these changes will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated in the UFSAR.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes have been incorporated to provide the 
capability to manually stroke the vent path using the block valve to 
control the pressure surge as a PORV opens. The resultant downstream 
piping forces were found acceptable, therefore, power can be 
maintained to the block valve when the block valve has been closed 
to isolate a PORV because of excessive seat leakage. This will allow 
operation of the PORVs in a manner similar to the guidance provided 
in GL 90-06 to improve PORV availability. These changes will improve 
the operator use of an isolated PORV since it is now analyzed to be 
manually cycled with the block valve closed and power maintained so 
the operator can use the PORV if required to mitigate the effects of 
a SGTR accident. This is consistent with the intent of the ISTS and 
does not affect the UFSAR, therefore, operation of the plant in 
accordance with the proposed amendment would not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location:  B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 1500l.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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 minimum water level that is required to be maintained over 
irradiated fuel assemblies during latching and unlatching of control 
element assemblies.
    Basis for proposed no significant hazards 
considerationdetermination: 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 fuel handling accident analysis assumes that a fuel assembly 
is dropped during fuel handling. During the latching and unlatching 
of the CEAs, the upper guide structure is in place and the CEDM 
extension shaft assemblies are disconnected from their CEA for 
subsequent removal with the vessel upper guide structure. The 
dropping of a CEA from the maximum height of six inches will not 
damage that particular fuel assembly or any surrounding fuel 
assemblies since this movement is confined to within the upper guide 
structure and the guide tubes of the associated fuel assembly during 
this activity. This less than six inches of movement does not have 
the potential to result in a fuel handling accident; therefore, an 
increase in the probability of this accident does not occur. The 
requirement to have at least 23 feet of water over the top of the 
irradiated fuel assemblies during fuel and CEA movement ensures 
that, should a fuel handling accident occur, the resulting offsite 
dose consequences are mitigated. The six inch movement of the CEA 
during CEA decoupling does not constitute fuel or CEA 

[[Page 42605]]
movement which would result in a fuel handling accident. As such, 
Technical Specifications are unchanged with respect to the 
mitigating requirements for a fuel handling accident.
    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 design, configuration, 
or method of operation of the plant; therefore, it does not create 
the possibility of a new or different kind of accident. Because no 
new equipment is being introduced, and no equipment is being 
operated in a manner inconsistent with its design, the possibility 
of equipment malfunction is not increased. The proposed change adds 
an exception to the applicability section and is bounded by the 
existing fuel handling accident analysis.
    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.
    There is no reduction in margin of safety in that 23 feet of 
water is still maintained over the irradiated fuel assemblies 
anytime there is a potential for a fuel handling accident. Adding 
the exception of the latching and unlatching of the CEAs to the 
applicability section does not involve a change in the accident 
analysis for fuel handling which remains bounding.
    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., et al., Docket No. 50-416, Grand Gulf 
Nuclear Station, Unit 1, Claiborne County, Mississippi

    Date of amendment request: July 21, 1995
    Description of amendment request: The proposed change requests that 
the current expiration date for license NPF-29 be changed to reflect 
the issuance date of the new license granted Grand Gulf on November 1, 
1984. The change consists of extending the expiration date to 40 years 
from the date of issuance of license NPF-29 (November 1, 1984 to 
November 1, 2024).
    Basis for proposed no significant hazards 
considerationdetermination: As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:
    a. No significant increase in the probability or consequences of 
an accident previously evaluated results from this change.
    The proposed change does not affect the design or operation of 
any plant system. The effect of 40 years of full power operations 
has previously been evaluated and documented in the Updated Final 
Safety Analysis Report (UFSAR). The design life of structures, 
systems and components is controlled by existing plant problems 
[sic., programs] and processes that are not affected by this change. 
The proposed change will simply allow Grand Gulf to achieve its 
original planned 40 years of service. Equipment associated with 
initiating event frequencies or accident mitigation must continue to 
meet all applicable maintenance and operability requirements 
regardless of license duration (It is also interesting to note that 
the license duration limitation of 40 years, as contained in 10 CFR 
50.51 is not a limitation resulting from concerns over plant aging 
effects. ``In fact, the limit was a compromise between the efforts 
of the Justice Department and electric cooperatives, who championed 
a 20-year limit on the basis of antitrust concerns, and the view of 
the utility industries that a longer period was necessary to ensure 
full amortization of a nuclear power plant.'' (56 FR 64961, December 
13, 1991)). Therefore, the probability or consequences of previously 
analyzed accidents are not significantly increased.
    b. The change would not create the possibility of a new or 
different kind of accident from any previously analyzed.
    The proposed change will not add any plant equipment or 
introduce any new modes of plant operation. The change will only 
amend the operating license to allow 40 years of full power 
operations. The proposed change does not affect the current 
maintenance or surveillance practices, which are designed to 
maintain and monitor the current service life of plant structures, 
systems and components in accordance with regulatory requirements. 
Therefore, the proposed change does not create the possibility of 
new equipment failure modes or a new or different kind of accident 
from any accident previously evaluated.
    c. The change would not involve a significant reduction in a 
margin of safety.
    The proposed change does not involve a significant reduction in 
a margin of safety since it only provides for 40 years of full power 
operations for which the plant is designed. Current Technical 
Specification surveillance requirements (e.g. associated with 10 CFR 
50 Appendix H) and other regulatory requirements remain in place and 
will ensure continued compliance with applicable safety margins.
    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: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, Mississippi 39120
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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 20, 1995
    Description of amendment request: The proposed Technical 
Specifications (TS) changes would remove the surveillance interval text 
for the 10 CFR Part 50, Appendix J, Type A test (Integrated Leak Rate 
Test or ILRT), and Drywell-to-Suppression Chamber (bypass) leakage test 
specified in TS Surveillance Requirements (SR) 4.6.1.2.a, 4.6.1.2.b, 
and 4.6.2.1.e.
    Basis for proposed no significant hazards 
considerationdetermination: 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 TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The primary containment and the suppression chamber are not 
considered to be accident initiators, they are accident mitigators. 
There are no physical or operational changes to the containment or 
suppression structure, system or components being made as a result 
of the proposed changes. These changes will not impose different 
requirements and adequate control of information will be maintained. 
These TS changes will not alter assumptions made in the safety 
analysis and licensing basis. Therefore, the proposed TS changes to 
eliminate the details of the test intervals will not increase the 
probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes remove the specific surveillance test 
interval text from TS and address the interval by direct reference 
to the applicable regulation. The proposed TS changes do not make 
any physical or operational changes to existing plant systems or 
components. Furthermore, the primary containment and suppression 
chamber act as 

[[Page 42606]]
accident mitigators not initiators. Therefore, the possibility of a new 
or different kind of accident than from any accident previously 
evaluated is not introduced.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    LGS [Limerick Generating Station] TS Bases 3/4 6.1.2 state that 
surveillance testing is consistent with 10 CFR 50, Appendix J and 
does not specify a SR test interval. TS Bases 3/4 6.2, describing 
the bypass test does not specify a SR test interval. However, the 
NRC Safety Evaluation related to amendment Nos. 68 (Unit 1) and 31 
(Unit 2) concluded that it is acceptable for the drywell-to-
suppression chamber test frequency to coincide with the 10 CFR 50, 
Appendix J, Type A test, since individual vacuum breaker leakage 
tests are an acceptable alternative to an integrated suppression 
pool bypass test during outages for which a Type A containment 
integrated leak rate test is not conducted. The alternative bypass 
test requirement, TS SR 4.6.2.1.f, is not affected by these changes.
    The Type A test, and bypass SR test intervals are adequately 
presented in the test implementing procedures, and TS will directly 
reference 10 CFR 50, Appendix J, for the appropriate test interval.
    Therefore, the proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: 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

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

    Date of amendment request: July 21, 1995
    Description of amendment request: The proposed amendment would 
change Technical Specifications Section 6.0 (Administrative Controls) 
to replace the title-specific list of members on the Plant Operating 
Review Committee (PORC) with a more general statement of membership 
requirements. The scope of disciplines represented on the PORC would 
also be expanded to include nuclear licensing and quality assurance. 
The proposed amendment would also change the title ``Resident Manager'' 
to ``Site Executive Officer.'' This title change would not affect the 
reporting relationship, authority, or responsibility of the position.
    Basis for proposed no significant hazards 
considerationdetermination: 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:
    Operation of the Indian Point 3 Nuclear Power Plant in 
accordance with the proposed amendment would not involve a 
significant hazards consideration as defined in 10 CFR 50.92, since 
it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes are administrative in nature and do not 
involve plant equipment or operating parameters. There is no change 
to any accident analysis assumptions or other conditions which could 
affect previously evaluated accidents. The proposed changes will not 
decrease the organization's ability to respond to a design basis 
accident.
    2. Create the possibility of a new or different kind of accident 
from those previously evaluated.
    Since the proposed changes are administrative in nature and do 
not involve hardware design, modifications or operation, the 
possibility of new or different accidents is not created.
    3. Involve a significant reduction in the margin of safety.
    The proposed title change for the Resident Manager is an 
administrative change and does not affect the responsibilities, 
authority, or reporting relationships for this management position. 
Replacing the title specific list of PORC members with a statement 
of membership requirements for the committee does not reduce the 
effectiveness of the committee to advise the Resident Manager (Site 
Executive Officer) on matters regarding nuclear safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle, 
New York, New York 10019.
    NRC Project Director: Ledyard B. Marsh

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

    Date of amendment request: March 30, 1995
    Description of amendment request: The proposed change to the 
Technical Specifications (TS) would change TS Table 3.3.1-2, ``Reactor 
Protection System Response Times'', TS Table 3.3.2-3, ``Isolation 
System Instrumentation Response Time'', TS Table 3.3.3-3, ``Emergency 
Core Cooling System Response Times'', and associated Bases. The 
proposed changes to the above-referenced TS Tables would eliminate the 
requirement to perform response time testing for certain classes of 
equipment.
    Basis for proposed no significant hazards 
considerationdetermination: As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The purpose of the proposed Technical Specification change is to 
eliminate response time testing requirements for selected 
instrumentation in the Reactor Protection System, Isolation System, 
and Emergency Core Cooling System. However, because of the continued 
application of other existing Technical Specification requirements 
such as channel calibrations, channel checks, channel functional 
tests, and logic system functional tests, the response time of these 
systems will be maintained within the acceptance limits assumed in 
plant safety analyses and required for successful mitigation of an 
initiating event. The proposed Technical Specification changes do 
not affect the capability of the associated systems to perform their 
intended function within their required response time.
    The BWR Owners' Group has completed an evaluation (NEDO-32291, 
``System Analyses for the Elimination of Selected Response Time 
Testing Requirements'') which demonstrates that response time 
testing is redundant to the other Technical Specification 
requirements listed in the preceding paragraph. These other tests 
are sufficient to identify failure modes or degradation in 
instruments response time and ensure operation of the associated 
systems within acceptance limits. There are no known failure modes 
that can be detected by response time testing that cannot be 
detected by the other Technical Specification tests. Hope Creek 
Generating Station is specifically bounded by the assumptions and 
justifications in General Electric Company Licensing Topical Report, 
NEDO-32291, ``System Analyses for Elimination of Selected Response 
Time Testing Requirements.''
    2. Will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    As discussed above, the proposed Technical Specification changes 
do not affect the capability of the associated systems to perform 
their intended function within the acceptance limits assumed in 
plant safety analyses and required for successful mitigation of an 
initiating event. The proposed elimination of response time testing 
would not result in any new 

[[Page 42607]]
equipment, operating modes, or plant configurations.
    3. Will not involve a significant reduction in a margin of 
safety.
    The current Technical Specification response times are based on 
the maximum allowable values assumed in the plant safety analyses. 
These analyses conservatively establish the margin of safety. As 
described above, the proposed Technical Specification changes do not 
affect the capability of the associated systems to perform their 
intended functions within the allowed response time used as the 
basis for the plant safety analyses. Plant and system response to an 
initiating event will remain in compliance within the assumptions of 
the safety analyses, and therefore the margin of safety is not 
affected.
    Although not explicitly evaluated, the proposed Technical 
Specification changes will provide an improvement to plant safety 
and operation by:
    a) Reducing the time safety systems are unavailable
    b) Reducing safety system actuations
    c) Reducing shutdown risk
    d) Limiting radiation exposure to plant personnel
    e) Eliminating the diversion of key personnel to conduct 
unnecessary testing.
    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: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

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

    Date of amendment request: April 18, 1995
    Description of amendment request: The proposed changes to the 
Technical Specifications (TS) would change TS Table 4.3.7.1-1 
``Radiation Monitoring Instrumentation Surveillance Requirements.'' 
This change would increase the channel functional test interval from 
monthly to quarterly for each instrument.
    Basis for proposed no significant hazards 
considerationdetermination: As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change involves no hardware changes, no changes to 
the operation of any systems or components, and no changes to 
existing structures. Increasing the interval between channel 
functional tests for the radiation monitoring instrumentation 
represent changes that do not affect plant safety and do not alter 
existing accident analyses.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed change is procedural in nature concerning the 
channel functional test frequency for the radiation monitoring 
instrumentation not already on a quarterly surveillance. The channel 
functional test methodology for these instruments remains unchanged. 
The proposed changes, while slightly increasing the possibility of 
an undetected instrument error, will not create a new or unevaluated 
accident or operating condition.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed change is in accordance with recommendations 
provided by the NRC regarding the improvement of Technical 
Specifications. These changes will result in perpetuation of current 
safety margins while reducing regulatory burden and decreasing 
equipment degradation.
    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: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

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

    Date of amendment request: May 4, 1995
    Description of amendment request: The proposed change to the 
Technical Specifications (TS) would change TS 3/4.6.1.8, ``Drywell and 
Suppression Chamber Purge System'', to increase the annual operational 
limit for the drywell and suppression chamber purge system from 120 to 
500 hours.
    Basis for proposed no significant hazards 
considerationdetermination: As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change involves no hardware changes and no changes 
to existing structures. Increasing the annual operational limit of 
the drywell and suppression chamber purge system will not increase 
the probability of a loss-of-coolant accident. While increased usage 
of the purge system will result in a slight increase in the 
possibility that these valves will be open during a LOCA, it will 
not alter or impact previous LOCA analyses.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed change will not result in an unanalyzed condition. 
While the increase in purge system operation will slightly increase 
the possibility of the containment vent and purge valves being open 
at the onset of a LOCA event, the valves have been established as 
capable of isolating the containment within five seconds. This is 
well within the bounds of existing LOCA analyses which assume an 
open duration of 175 seconds. Therefore, this change will not 
require a new or different accident analysis.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed change will not alter existing systems, equipment, 
components, or structures. The method of operating the drywell and 
suppression chamber purge system will not be altered by the 
increased annual usage. While there is a slight increase in the 
possibility of purge operations at the onset of a LOCA, any 
resulting release would be insignificant and bounded by existing 
LOCA analyses. Operation of the drywell and suppression chamber 
purge system based on these proposed changes will remain within the 
guidance provided in the NRC's Branch Technical Position CSB 6-4.
    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:  Pennsville Public Library, 
190 S. Broadway, Pennsville, New Jersey 08070
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

Saxton Nuclear Experimental Corporation (SNEC), Docket No. 50-146, 
Saxton Nuclear Experimental Facility (SNEF), Bedford County, 
Pennsylvania

    Date of amendment request: June 2, 1995, as supplemented on June 
23, 1995.
    Description of amendment request: The proposed changes to the 
technical specifications are administrative in 

[[Page 42608]]
nature. The proposed amendment would revise the organization structure 
associated with the SNEF to allow General Public Utilities Nuclear 
Corporation resources to be applied to SNEC activities within their 
normal organizational structure; eliminating the need to identify and 
compartmentalize a portion of the organization as specific to SNEC. The 
proposed amendment would also revise the description and drawing of the 
SNEF site to reflect multiple gates in the SNEF fence.
    Basis for proposed no significant hazards 
considerationDetermination: As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below: The proposed changes 
do not involve a significant hazards considerations because the changes 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The administrative changes will not impact the physical 
condition of the containment vessel as it relates to the risk of 
fire, flood or radiological hazard.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    In its present condition, the only accidents applicable to the 
site are those addressed above.
    3. Involve a significant reduction in a margin of safety.
    The proposed administrative changes would have no effect on any 
margins of safety for any evaluated accidents.
    The NRC staff has reviewed the analysis of the licensee 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:  Saxton Community Library, 911 
Church Street, Saxton, Pennsylvania 16678Attorney for the Licensee: 
Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts, and Trowbridge, 
2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: Seymour H. Weiss

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 30, 1995
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) for the pressurizer power 
operated relief valves (PORVs) to follow the guidance of Generic Letter 
(GL) 90-06, Generic Issue 70, and the improved Westinghouse 
Standardized Technical Specifications (NUREG-1431, Rev. 1).
    Basis for proposed no significant hazards 
considerationdetermination: 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 because 
the physical characteristics of the PORVs and their block valves 
remain unchanged. No changes to any hardware or software that 
affects these components is planned.
    The PORVs are pressure relieving devices and only two failure 
modes need to be considered. The first is that one or more PORVs or 
block valves fail to open when required. This is not
    a significant concern and is not a credible cause of any 
accident. The second mode is failing to close which includes 
depressurization of the RCS [reactor coolant system] and a reactor 
trip on low pressurizer pressure or overtemperature [delta]T. The 
consequences for the more limiting Pressurizer Safety Valve 
Accidental Depressurization event has been analyzed with acceptable 
results.
    There is no increase in the consequences of an accident as a 
result of this change, because only one PORV is required to mitigate 
the consequences of a design basis Steam Generator Tube Rupture. 
There is sufficient redundancy to ensure one PORV is available to 
perform this function even if one PORV is inoperable or incapable of 
being manually cycled. The validation of the Emergency Operating 
Procedures on the VCSNS [Virgil C. Summer Nuclear Station] simulator 
demonstrated that one pressurizer PORV has sufficient capacity to 
depressurize the RCS in a time frame which will not cause the 
offsite doses presented in the FSAR [Final Safety Analysis Report] 
to be exceeded.
    The PORVs are utilized to depressurize the RCS and equalize the 
pressure between the primary and secondary systems. This stops the 
intrusion of RCS water into the secondary which can be released into 
the atmosphere. By the time the PORVs are called upon, the affected 
steam generator (SG) has been identified and steps have been taken 
to isolate the faulted SG. This acts to minimize the radiological 
impact on the health and safety of the public. In all cases, the 
dose results are within 10 CFR 100 limits.
    2. The possibility of an accident or a malfunction of a 
different type than any previously evaluated is not created.
    The proposed TSCR [TS Change Request] does not involve any 
physical changes to the plant or decrease the number of PORVs and 
block valves that must be capable of performing their intended 
function. These components are used to mitigate the effects of 
postulated events and their failure has already been considered. The 
worst case failure, either not opening or not closing, has been 
evaluated and is bounded by other more limiting accidents.
    3. The margin of safety has not been significantly reduced.
    The currently approved TS permits all three PORVs and/or their 
block valves to be inoperable as long as precautions are taken to 
assure that RCS would not leak-by, assuming single failures and 
spurious operation. The proposed TSCR would require a minimum of two 
PORVs and block valves to be operable, or at least capable of being 
manually cycled, in Modes 1, 2, and 3. This is in fact an increase 
in margin and provides for greater reliability with the added 
benefit that the probability of challenges to the pressurizer code 
safety valves will be lessened.
    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 (SCE&G), 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: July 28, 1995
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to exclude the requirement to 
perform the slave relay test of the 36-inch containment purge supply 
and exhaust valves on a quarterly basis while the plant is in Modes 1, 
2, 3, or 4.
    Basis for proposed no significant hazards 
considerationdetermination: 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?
    No, the probability or consequences of an accident previously 
evaluated would not be increased since no credit is taken for the 
valves in FSAR [Final Safety Analysis Report] Chapter 15.
    The only credible accident discussed in FSAR Chapter 15 that 
applies to these valves is a fuel handling accident inside 

[[Page 42609]]
containment (15.4.5.1). The analysis assumes the escaped gases are 
released instantaneously to the environment via the Reactor
    Building purge system. The analysis does not take credit for 
these valves nor for filtration or holdup time during release. The 
result of the analysis is acceptable and offsite doses are within 
the limits of 10 CFR 100.
    TS 3.6.1.7 requires that these valves be sealed shut during 
Modes 1, 2, 3, and 4. When sealed shut, these valves will not open 
via any signal.
    With these valves already in a shut position, neither the 
probability nor the consequences of an accident are increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any previously evaluated?
    No, the 36'' [inch] containment purge exhaust and supply valves 
will not be placed in a condition different from that evaluated 
previously.
    The only credible accident discussed in FSAR Chapter 15 that 
applies to these valves is a fuel handling accident inside 
containment (15.4.5.1). The analysis assumes the escaped gases are 
released instantaneously to the environment via the Reactor Building 
purge system. The analysis does not take credit for these valves nor 
for filtration or holdup time during release. The result of the 
analysis is acceptable and offsite doses are within the limits of 10 
CFR 100.
    Additionally, TS 3.6.1.7. requires that these valves be sealed 
shut during Modes 1, 2, 3, and 4. When sealed shut, these valves 
will not open via any signal.
    3. Does the change involve a significant reduction in the margin 
of safety?
    TS 4.3.2.1. requires that this slave relay test be performed 
quarterly. This surveillance is accomplished for the 36'' [inch] 
containment purge exhaust and supply valves by cycling the 
respective K615 relay. This will not provide assurance that the 
valve will perform its safety function since the valve is sealed 
closed. The proposed change will exclude the requirement to perform 
the K615 relay test (auto actuation logic and actuation relays - 
slave relay test) on a quarterly basis while the plant is in Modes 
1, 2, 3,or 4.
    TS 3.6.1.7. requires that these valves be sealed shut during 
Modes 1, 2, 3, and 4. When sealed shut, these valves will not open 
via any signal. Since this relay would not be needed to supply a 
signal to place these valves in the closed position, the margin of 
safety is not affected.
    Based on the preceding analysis, SCE&G has determined that this 
change does no 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: 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. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, 
Alabama

    Date of amendment request: June 2, 1995 (TS 353)
    Description of amendment request: The proposed amendment supports 
replacement of the existing power range neutron monitoring equipment 
and implements ARTS/MELLL [average power range monitor and rod block 
monitor technical specifications/maximum extended load line limit] 
analysis improvements.
    Basis for proposed no significant hazards 
considerationdetermination: As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Group A Changes: This proposed TS change is associated with the 
NUMAC PRNM [nuclear measurement analysis and control power range 
neutron monitor] retrofit design. The proposed TS change involves 
modification of the LCOs [limiting condition for operations] and SRs 
[surveillance requirements] for equipment designed to mitigate 
events which result in power increase transients. For the APRM 
[average power range monitor] system mitigative action is to block 
control rod withdrawal or initiate a reactor scram which terminates 
the power increase when setpoints are exceeded. For the RBM [rod-
block monitor] system mitigative action is to block continuous 
control rod withdrawal prior to exceeding the MCPR [minimum critical 
power ratio] safety limit during a postulated Rod Withdrawal Error 
[RWE]. The worst case failure of either the APRM or the RBM systems 
is failure to initiate mitigative action (failure to scram or block 
rod withdrawal). Failure to initiate mitigative action will not 
increase the probability of an accident. Thus, the proposed change 
does not increase the probability of an accident previously 
evaluated.
    For the APRM and the RBM systems, the NUMAC PRNM design, 
together with revised operability requirements (LCOs) and revised 
testing requirements (SRs), results in equipment which continues to 
perform the same mitigation functions under identical conditions 
with reliability equal to or greater than the equipment which it 
replaces. Because there is no change in mitigation functions and 
because reliability of the functions is maintained, the proposed 
change does not involve an increase in the consequences of an 
accident previously evaluated.
    Group B Changes: This proposed change is associated with 
implementation of the ARTS/MELLL analysis. The proposed change will 
permit expansion of the current allowable power/flow operating 
region and will apply a new methodology for assuring that fuel 
thermal and mechanical design limits are satisfied. Reference 3 
evaluates operation in the MELLL region with assumed implementation 
of the ARTS changes. The conclusion of reference 3 is that for all 
events and parameters considered there is adequate design margin for 
operation in the MELLL region. Because operation in the MELLL region 
maintains adequate design margin, the proposed change does not 
significantly increase the probability of an accident previously 
evaluated.
    In support of operation in the MELLL region, the proposed change 
modifies flow-biased APRM scram and rod block setpoints and 
implements new RBM power-biased setpoints. This potentially changes 
the way in which the APRM and RBM systems perform their mitigation 
functions. However, no credit for the flow-biased APRM scram or rod 
block is taken in mitigation of any design basis event; thus, 
changing the APRM setpoints does not impact the consequences of any 
accident previously evaluated. The proposed changes to the RBM 
system potentially impact mitigation of the RWE. However, per 
discussion in reference 3, the proposed RBM changes will assure that 
the RWE is not a limiting event; thus, the consequences of the RWE 
are not increased. The proposed change does not increase the 
consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes (Group A and Group B) involve modification 
and replacement of the existing power range neutron monitoring 
equipment, modification of the setpoints and operational 
requirements for the APRM and RBM systems, implementation of a new 
methodology for administering compliance with fuel thermal limits, 
and operation in an extended power/flow domain. These proposed 
changes do not modify the basic functional requirements of the 
affected equipment, create any new system interfaces or 
interactions, nor create any new system failure modes or sequence of 
events that could lead to an accident. The worst case failure of the 
affected equipment is failure to perform a mitigation action, and 
failure of this mitigative equipment does not create the possibility 
of a new or different kind of accident. The proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    Group A Changes: This proposed TS change is associated with the 
NUMAC PRNM retrofit design. The NUMAC PRNM change does not impact 
reactor operating parameters nor the functional requirements of the 
power 

[[Page 42610]]
range neutron monitoring system. The replacement equipment continues to 
provide information, enforce control rod blocks and initiate reactor 
scrams under appropriate specified conditions. The proposed change 
does not revise any safety margin requirements. The replacement 
APRM/RBM equipment has improved channel trip accuracy compared to 
the current system and meets or exceeds system requirements 
previously assumed in setpoint analysis. Thus, the ability of the 
new equipment to enforce compliance with margins of safety equals or 
exceeds the ability of the equipment which it replaces. The proposed 
change does not involve a reduction in a margin of safety.
    Group B Changes: This proposed change is associated with 
implementation of recommendations presented in the ARTS/MELLL 
analysis. Operation in the MELLL region does not affect the ability 
of the plant safety-related trips or equipment to perform their 
functions, nor does it cause any significant increase in offsite 
radiation doses resulting from any analyzed event. Analyses 
documented in reference 3 demonstrate that for operation in the 
MELLL region adequate margin to design limits is maintained. 
Implementation of the ARTS improvements provides flow- and power-
dependent thermal limits which maintain existing margins of safety 
in normal operation, anticipated operational occurrences and 
accident events. Implementation of power-biased RBM setpoints 
improves the margin of safety in a postulated RWE by assuring that 
the RWE is not a limiting event. The proposed change does not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Athens Public Library, South 
Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: June 8, 1995 (TS 361)
    Description of amendment request: The proposed amendment clarifies 
the definition of operability for the RHRSW system standby coolant 
supply capability and revises the instrument numbers for several 
instruments that have been upgraded.
    Basis for proposed no significant hazards 
considerationdetermination: As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change to TS 3.5.C.3 clarifies the operability 
requirements of the standby coolant supply capability. It does not 
change or degrade the nuclear safety characteristics of the RHRSW 
and RHR systems and will not affect the intent of the TS. The 
operation of the standby coolant supply capability is not a 
precursor to any design basis accident or transient analyzed in the 
BFN FSAR. The proposed changes to instrument numbers are 
administrative changes for the upgraded drywell temperature and 
pressure instrumentation. The proposed changes do not affect the 
design basis or the safety functions of the Primary Containment 
system, since the function and instrumentation range is not changed. 
Therefore, the probability of occurrence or the consequences of an 
accident or malfunction of equipment important to safety previously 
evaluated in the safety analysis report has not been increased.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The possibility for an accident or malfunction of a different 
type than any evaluated previously in the safety analysis report is 
not created by this change. The change to TS 3.5.C.3 adds the 
indication of associated valves of the function involved and a 
clarification of operability for the standby coolant supply 
connection to be commensurate with the RHR cross-connect capability. 
The proposed changes to instrument numbers are administrative 
changes effected by the upgrade of instrumentation. There are no 
automatic actions affected or compromised by these changes.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change to TS 3.5.C.3 does not affect any acceptable 
limit of operation or analysis assumption in the TS or Bases. The 
changes affect neither setpoints, calibration intervals, nor 
functional test intervals. The change does not affect any acceptable 
limit of operation or analysis assumption found in the TS or their 
bases. The proposed administrative changes to the instrument numbers 
do not affect the setpoint, calibration interval or function of the 
instrumentation. These changes do not affect any limiting conditions 
of operation or analysis assumption in the TSs or their bases. 
Therefore, the change does not reduce the margin of safety as 
defined in the basis for any TS.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Athens Public Library, South 
Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: June 16, 1995 (TS 360)
    Description of amendment request: The proposed change will revise 
the BFN Units 1, 2, and 3 Technical Specifications (TS) to permit the 
Traversing In-Core Probe (TIP) system to be considered operable with 
less than five TIP machines operable. The proposed amendment will allow 
the utilization of substitute data in lieu of data from inaccessible 
TIP measurement locations. The substitute data will be derived from 
either symmetric TIP measurement locations (under certain core 
conditions) or from normalized TIP data as calculated by the on-line 
core monitoring system.
    Basis for proposed no significant hazards 
considerationdetermination: As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The TIP system is not used to prevent, or mitigate the 
consequences of any previously analyzed accident or transient; nor 
are any assumptions made in any accident analysis relative to the 
operation of the TIP system. The primary containment isolation 
function (TIP withdrawal) is not affected. The
    proposed TS change does not alter the fundamental process 
involved in calibrating neutron instrumentation (LPRMs) [local power 
range monitors], but requires that only the equipment associated 
with the TIP channels necessary for recalibrating LPRMs and for core 
monitoring functions be operable. Collection and storage of TIP data 
without using all TIP channels is acceptable because TIP machine 
normalization factors are ultimately derived from the most recent 
full core TIP set, which intercalibrates the TIP machines in a 
common core location.
    Additionally, the use of symmetric detectors and analytical 
values as substitute data for inaccessible TIP channels does not 
compromise the ability of the process computer to accurately 
represent the spatial neutron flux distribution of the reactor core. 


[[Page 42611]]
The core monitoring methodology is presently based on symmetry of rod 
patterns and fuel loading. This is not changed but extended to use a 
higher order of symmetry (octant symmetry) which exists with ``type 
A'' sequence rod patterns. Therefore, this change does not increase 
the probability or consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve the installation of any new 
equipment, or the modification of any equipment designed to prevent 
or mitigate the consequences of accidents or transients. Therefore, 
the proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The total core TIP reading uncertainties will remain within the 
assumptions of the licensing basis. Therefore, the margin of safety 
to the MCPR [minimum critical power ratio] safety limits is not 
reduced. The ability of the process computer to accurately represent 
the spatial neutron flux distribution for the reactor core is not 
compromised. Additionally, the computer's ability to accurately 
predict the LHGR [linear heat generation rate], APLHGR [average 
planar linear heat generation rate], MCPR and its ability to provide 
for LPRM calibration is not compromised. Therefore, the proposed 
changes do not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Athens Public Library, South 
Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: October 21, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3/4.6.1.2, ``Primary Containment 
Leakage.'' The changes would clarify that the main steam line isolation 
valves leakage is accounted for separately from the integrated primary 
containment leak rate or combined local leak rate results. Also, two 
references would be deleted, the test duration for use of Bechtel 
Corporation Topical Report BN-TOP-1 would be clarified, and the 
requirement to perform the third integrated leak rate in each 10-year 
service period in conjunction with the 10-year plant inservice 
inspection would be deleted. Exemptions to 10 CFR Part 50 Appendix J, 
``Primary Reactor Containment Leakage Testing for Water-Cooled Power 
Reactors,'' are also being requested in conjunction with the proposed 
TS changes.
    Basis for proposed no significant hazards 
considerationdetermination: 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:
    Part A - Formalize the Approval for Excluding the Main Steam 
Line Isolation Valve Leakages from Inclusion in i) the Overall 
Integrated Primary Containment Leak Rate and ii) the Combined Local 
Leak Rate, and Clarify that the Main Steam Lines are Not Required to 
be Vented and Drained for Type A Testing
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Since Appendix J was originally envisioned, alternative means of 
meeting the intent of these requirements have been developed which 
provide an equivalent level of protection of the public health and 
safety. However, since some of these alternatives deviate from the 
specific wording of Appendix J, exemptions are appropriate for these 
alternatives. Implicit in the FSAR treatment of the main steam line 
leakage, as well as the TS requirements for main steam line leakage, 
are several deviations from the specific requirements of Appendix J. 
Although PNPP's methods and practices for Appendix J testing have 
been previously described in correspondence to the NRC, a formal 
exemption was not recognized to be needed at that time in that the 
NRC's approval was perceived to be received by the issuance of the 
PNPP TS. Exemption to four separate paragraphs of 10 CFR 50 Appendix 
J will document the approvals previously received and incorporated 
into the TS for main steam line isolation valve testing during the 
initial licensing of the PNPP. This TS change adds references to 
footnotes within the TS LCO 3.6.3.1 to clarify which conditions 
represent exemptions to Appendix J. These exemptions are described 
in the Bases.
    PNPP utilized the criteria described in the Standard Review Plan 
(SRP), Section 15.6.5, Appendix D, ``Radiological Consequences of a 
Design Basis Loss-of-Coolant Accident: Leakage from Main Steam 
Isolation Valve Leakage Control System (Rev. 1 - July 1981).'' This 
is an alternative, NRC approved method for assessing the MSIV 
leakage contribution and determining the radiological consequences.
    In accordance with the SRP, the safety analysis for a design 
basis LOCA includes the maximum main steam line leak rate separately 
from the maximum containment leak rate. Within Appendix J it is 
implied that Type A tests are intended to measure the primary 
containment overall integrated leak rate, but this vas before the 
SRP Section was developed which allows the MSIV contribution to be 
accounted for separately in the safety analysis. Therefore, the MSIV 
leak rate should not be included in the measurement of the ILRT. 
Including the MSIV leakage in the combined local leak rate limit is 
also not necessary since a specific Type C MSIV leak rate has been 
specified in TS 3.6.1.2.
    In summary, there is no change in the probability or 
consequences of any accident since the addition of the references 
and footnotes to clarify the TS LCO and Actions do not change the 
design of the plant, nor the operational characteristics of any 
plant system, nor the procedures by which the Operators run the 
plant. These changes only cite formal Appendix J exemptions which 
are requested to document the approval previously received. A formal 
request for exemption to the applicable paragraphs of 10 CFR 50 
Appendix J is also being submitted in a separate letter in 
conjunction with this proposed TS change.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
There are no design changes being made that would create a new type 
of accident or malfunction, and the method and manner of plant 
operation remains unchanged. The only change being made is an 
exemption to 10 CFR 50 Appendix J which will be cited in the TS to 
document the implicit and explicit approvals of the PNPP design and 
testing methods for main steam line isolation valves. The 
requirements and bases for which the formal exemption is sought are 
currently presented and implemented in the licensing basis and the 
TS for PNPP. The objective of the regulation is being met and will 
continue to be met. The exemption to 10 CFR 50 Appendix J is being 
submitted in a separate letter in conjunction with this proposed TS 
change.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    These changes do not involve a significant reduction in the 
margin of safety because they are administrative in nature. The 
proposed change will only cite the NRC exemption that grants the 
deviation from Appendix J. The proposed changes do not affect any 
USAR design bases or accident assumptions. Therefore, the proposed 
changes do not reduce the margin of safety as defined in the bases 
for any Technical Specification.
    Part B - Revise Surveillance Requirement 4.6.1.2 to Eliminate 
Unnecessary References and ClarifY the Use of BN-TOP-1 

[[Page 42612]]

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Surveillance Requirement 4.6.1.2 is proposed to be revised to 
eliminate the direct reference to the ANSI Standards N45.4 and N56.8 
within the text, because these same Standards are listed within 
Appendix J. It is unnecessary to repeat the references to the 
Standards within the Technical Specifications because the PNPP is 
still required to be in compliance with the regulations. No 
additional benefits are gained and licensee flexibility to upgrade 
to later versions of the Standards is reduced since a Technical 
Specification change is necessary to change the version of the 
Standard to which PNPP is committed. This change removes a redundant 
requirement to list these Standards in the Technical Specifications. 
Therefore, this change cannot involve a significant increase in the 
probability or consequences of an accident because the regulation is 
still required to be met.
    A reference to Topical Report BN-TOP-1 continues to be retained 
within Surveillance Requirement 4.6.1.2, and the use of the report 
is clarified to be for test durations less than 24 hours. This 
reference is retained within the TS since a reference to BN-TOP-1, 
though not specifically included within Appendix J, is allowed by 
Section 7.6 of ANSI N45.4-1972 and has been approved for PNPP use by 
the NRC. The TS Bases are also proposed to be revised to include a 
statement that the use of BN-TOP-1 is in accordance with Appendix J.
    These changes result in no changes to plant systems and have no 
effect on accident conditions or assumptions. These proposed changes 
do not affect possible initiating events for accidents previously 
evaluated, or any system functional requirements. Hence, these 
changes are purely administrative in that they are designed to 
eliminate a redundant requirement and clarify the applicability and 
acceptability of an alternative leak rate testing provision within 
the TS. These changes do not affect plant operation in any way. 
Therefore, the proposed changes do not affect the probability or 
consequences of any accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no design changes being made that would create a new 
type of accident or malfunction, and the method and manner of plant 
operation remains unchanged. These changes eliminate a redundant 
requirement and clarify the applicability and acceptability of 
alternative leak rate testing provisions within the TS. Since the 
alternative leak rate testing provisions have been approved by the 
NRC, the objective of the regulation continues to be met. Therefore, 
the proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    These changes do not involve a significant reduction in the 
margin of safety because they are administrative in nature and 
either eliminate a redundant requirement or clarify the 
applicability and acceptability of an alternative, NRC approved, 
leak rate testing provision within the TS. The proposed changes do 
not affect any USAR design bases or accident assumptions. Therefore, 
the proposed changes do not reduce the margin of safety as defined 
in the Bases for any Technical Specification.
    Part C - Decouple Performance of the Third Type A Test from the 
Shutdown for the 10-Year Plant Inservice Inspection
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change revises Surveillance Requirement 4.6.1.2.a 
by removing the second sentence requiring that the third test of 
each containment Integrated Leak Rate Test (ILRT) set be conducted 
during the shutdown for the 10-year plant inservice inspection. A 
request for an exemption to 10 CFR 50 Appendix J, Paragraph 
III.D.l(a) is also being submitted in conjunction with this proposed 
change. Note that this change is also included in the proposed 
Appendix J rule changes currently under consideration and has been 
approved for several other plants. The deletion of this requirement 
from the Technical Specifications does not impact plant safety 
because the 10 CFR 50 Appendix J requirement that three Type A 
containment ILRT tests to be performed over a 10 year period is not 
affected. This change only removes an unnecessary connection between 
the two regulations.
    The proposed change results in no changes to plant systems. The 
proposed change has no effect on accident conditions or assumptions. 
The proposed change does not affect possible initiating events for 
accidents previously evaluated, or any system functional 
requirements. Hence, the proposed change removes an unnecessary tie 
between regulations and does not affect plant operation in any way.
    In summary, there is no change in the probability or 
consequences of any accident since the revision of the existing 
Surveillance Requirement to reflect the removal of an unnecessary 
tie between regulations does not change the design of the plant, nor 
the operational characteristics of any plant system, nor the 
procedures by which the Operators run the plant.
    2. The propose change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change removes an unnecessary tie between 
regulations. The objective of the regulation continues to be met. 
There are no design changes being made that would create a new type 
of accident or malfunction, and the method and manner of plant 
operation remains unchanged. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not involve a significant reduction in 
the margin of safety because they are administrative in nature and 
remove an unnecessary tie between requirements. The proposed change 
does not affect any USAR design bases, accident assumptions. or 
Technical Specification Bases. Therefore, the proposed change does 
not reduce the margin of safety as defined in the bases for any TS.
    Based upon the above considerations, it has been concluded that 
the proposed changes do not involve significant hazards 
considerations.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: June 9 and 30, 1995
    Description of amendment request: The licensee has requested a one-
time extension of the performance intervals for certain Technical 
Specification Surveillance Requirements (SRs). Affected SRs include 
valve testing, and undervoltage instrumentation testing.
    Basis for proposed no significant hazards 
considerationdetermination: 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 TS change requests one-time only extensions of the 
surveillance intervals related to: a) ASME Section XI valve leak 
rate, stroke and timing, and position indication testing; b) 
Accident Monitoring Instrumentation related to valve position 
indication testing; c) Division 1, 2, and 3 Degraded Voltage and 
Undervoltage instrumentation LSFT; and, d) leak rate testing for 
hydrostatically tested containment isolation valves.
    Based on the discussion in the License Amendment Request which 
shows:
    i) The extension of the interval for ASME Section XI stroke and 
timing, leak rate measurement and position indication testing 

[[Page 42613]]
requirements are acceptable based on results of past testing which 
indicates a margin to TS limits will be maintained;
    ii) The extension of the interval for Position Indication 
Calibration as specified in Table 4.3.7.5-1, Item 17 is acceptable 
based on the testing results from the past two refueling outages 
that indicate no failures have occurred:
    iii) LSFT interval extension for the Division 1, 2, and 3 
Degraded Voltage and Undervoltage instrumentation is acceptable 
based on the NRC Safety Evaluation Report (Peach Bottom Atomic Power 
Plant, Units 2 and 3, dated August 2, 1993) which supported 
extension of the interval for LSFT from 18 to 24 months. This was 
based on the small probability of relay or contact failure relative 
to mechanical component failure probability and, therefore, the 
increase in LSFT interval represented no significant change in the 
overall safety system unavailability; and,
    iv) The extension of the interval for hydrostatic leak testing 
of containment isolation valves is acceptable based on the 
consistently low past leak rate data which is a small percentage of 
the TS limits.
    Therefore, from the above it is shown that the proposed changes 
will not significantly increase the probability of an accident 
previously evaluated.
    The proposed TS change requests one-time only extensions of the 
surveillance intervals related to TS SR 4.3.3.1, Table 4.3.3.1-1, 
Items D.1 and D.2, Division 1, 2, and 3 Degraded Voltage and 
Undervoltage instrumentation calibration. [...] extension of the 
interval for this instrumentation is acceptable based on the testing 
results from the past two refueling outages. No failures have 
occurred which would negate the assurance that the instrumentation 
would function as required for the requested extended period. 
Accordingly, the proposed change will not significantly increase the 
probability or consequences of an accident previously evaluated.
    2. The proposed change would not create the possibility of a new 
or
    different kind of accident from any accident previously 
evaluated.
    The proposed TS change requests one-time extensions of the 
surveillance intervals for ASME Section XI valve testing, 
instrumentation calibration, instrument channel LSFT, containment 
isolation valve hydrostatic leak rate testing. The proposed changes 
do not necessitate a physical alteration to the plant (no new or 
different type of equipment will be installed). In that the 
requested extension durations are small as compared to the overall 
interval allowed by TS, NRC and industry evaluations support 
extension of LSFT, and past testing results provide confidence of no 
effect on equipment availability by extending the surveillance 
interval, the change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed TS change requests one-time extensions of the 
surveillance intervals for the Division 1, 2, and 3 Undervoltage and 
Degraded Voltage instrumentation calibration. The proposed changes 
do not necessitate a physical alteration to the plant (no new or 
different type of equipment will be installed). In that the 
requested extension durations are small as compared to the overall 
interval allowed by TS and past testing results provide confidence 
of no effect on equipment availability by extending the surveillance 
interval, the change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change will not involve a significant reduction 
in the margin of safety.
    The proposed TS change requests a one-time extension of the 
surveillance intervals for ASME Section XI valve testing, 
instrumentation calibration, instrument channel LSFT, and 
containment isolation valve hydrostatic leak rate testing. The 
proposed changes do not necessitate a physical alteration to the 
plant (no new or different type of equipment will be installed). In 
that the requested extension durations are small as compared to the 
overall interval allowed by TS, NRC and industry evaluations support 
extension of LSFT, and past testing results provide confidence of no 
effect on equipment availability by extending the surveillance 
interval, the change does not involve a significant reduction in the 
margin of safety.
    The proposed TS change requests a one-time extension of the 
surveillance intervals for the division 1, 2, and 3 Undervoltage and 
Degraded Voltage instrumentation calibration. The proposed changes 
do not necessitate a physical alteration to the plant (no new or 
different type of equipment will be installed). In that the 
requested extension durations are small as compared to the overall 
interval allowed by TS and past testing results provide confidence 
of no effect on equipment availability by extending the surveillance 
interval, the 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: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

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

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: July 19, 1995
    Description of amendments request: Amend the Sequoyah Nuclear 
Plant, Units 1 and 2 Technical Specification to incorporate new 
requirements associated with steam generator tube inspections and 
repair.
    Date of publication of individual notice in the Federal Register: 
August 1, 1995 (60 FR 39198)
    Expiration date of individual notice: August 31, 1995
    Local Public Document Room Location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Notice Of Issuance Of Amendments To Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these 

[[Page 42614]]
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 applications for amendments: December 30, 1993 and July 12, 
1994. The December 30, 1993, application was supplemented by letters 
dated November 30, 1994, May 24, 1995, and June 21, 1995, and the July 
12, 1994, application was supplemented by letter dated June 21, 1995.
    Brief description of amendments: The amendments (1) revise the 
degraded voltage relay trip setpoint and (2) enhance the current 
presentation of the information regarding the loss-of-voltage relay 
setpoint. A time-voltage curve has been added to the technical 
specifications as a more accurate characterization of the inverse-time 
relay response.
    Date of issuance: July 21, 1995
    Effective date: July 21, 1995, to be implemented within 45 days of 
issuance.
    Amendment Nos.: Unit 1 - Amendment No. 96; Unit 2 - Amendment No. 
84; Unit 3 - Amendment No. 67
    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 and August 
17, 1994 (59 FR 29625 and 59 FR 42334) The November 30, 1994, May 24, 
1995, and June 21, 1995, letters 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 21, 1995.No 
significant hazards consideration comments received: No.
    Local Public Document Room Location:  Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.

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

    Date of application for amendment: February 6, 1995
    Brief description of amendment: The amendment allows the relocation 
of cycle-specific core operating limits of Figure 3.1-1, Shutdown 
Margin versus Boron Concentration in Technical Specification (TS) 
3.1.1.2, Shutdown Margin- Modes 3, 4, and 5, to the plant Core 
Operating Limits Report.
    Date of issuance: August 1, 1995
    Effective date: August 1, 1995
    Amendment No. 59
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications
    Date of initial notice in Federal Register: March 15, 1995 (60 FR 
14017) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 1, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room Location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

    Date of application for amendment: March 30, 1995, as supplemented 
July 6, 1995. The July 6, 1995, submittal did not change the initial no 
significant hazards consideration determination; it contained 
clarifying information only.
    Brief description of amendment: The amendment revises the Emergency 
Diesel Generator (EDG) surveillance requirements contained in TS 3/
48.1.1.2 to be consistent with NUREG-1431, ``Standard Technical 
Specifications, Westinghouse Plants,'' and to eliminate the need for 
duplicate EDG testing being performed to satisfy the requirements of 
the Station Blackout Rule and the Maintenance Rule.
    Date of issuance: August 1, 1995
    Effective date: August 1, 1995
    Amendment No.: 60
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20515) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 1, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room Location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

    Date of application for amendments: June 8, 1995, which superseded 
the December 16, 1994, request in its entirety, and additional 
correspondence dated November 30, 1994, April 27, May 5, May 11 and 
June 23, 1995.
    Brief description of amendments: The amendments revised Figure 3.4-
4a ``Nominal PORV Pressure Relief Setpoint Versus RCS Temperature for 
the Cold Overpressure Protection (LTOP) System'' in the Braidwood Unit 
1's Technical Specifications. The revision extends the applicability of 
Figure 3.4-4a from 5.37 effective full power years (EFPY) to 16 EFPY. 
In addition, the amendments remove the 638 psig administrative limit 
line from the LTOPS curve, because the appropriate instrument 
uncertainties and discharge piping pressure limits have been 
incorporated in the new curve. Finally, the amendments contains 
administrative changes to Figure 3.4-4a and its associated index page.
    Date of issuance: July 24, 1995
    Effective date: July 24, 1995
    Amendment Nos.: 64 and 64
    Facility Operating License Nos. NPF-72 and NPF-77: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 21, 1995 (60 FR 
32360). The June 23, 1995, letter, corrected a collating error in the 
June 8, 1995, submittal and did not change the initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated July 24, 1995.No significant hazards consideration 
comments received: No
    Local Public Document Room Location: Wilmington Public Library, 201 
S. Kankakee Street, Wilmington, Illinois 60481.

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

    Date of application for amendments: March 23, 1994, as supplemented 
on July 26, 1994, and subsequently superseded by a submittal dated 

[[Page 42615]]
February 15, 1995. The February 15, 1995, request was supplemented on 
February 28, 1995.
    Brief description of amendments: The amendments approve a maximum 
moderator temperature coefficient (MTC) of +7 pcm/ deg.F and relocate 
specification of the cycle specific MTC from the Technical 
Specifications to the operating limits report. The staff also approved 
the methodology proposed by the licensee for ensuring that the plants 
continue to meet the anticipated transient without scram (ATWS) rule 
(10 CFR 50.62) during operation with cycle specific MTCs.
    Date of issuance: July 27, 1995Effective date: Immediately, to be 
implemented within 30 days.
    Amendment Nos.: Byron Units 1 and 2 - 73, 73 and Braidwood Units 1 
and 2 - 65, 65
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 12, 1995 (60 FR 
18623) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 27, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room Location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

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: July 29, 1992, as supplemented 
January 14, 1993, February 16, 1993, and May 9, 1995.
    Brief description of amendments: The amendments upgrade the current 
custom Technical Specifications for Dresden and Quad Cities to the 
Standard Technical Specifications contained in NUREG-0123, ``Standard 
Technical Specification General Electric Plants BWR/4.'' These 
amendments upgrade only Section 3/4.3, ``Reactivity Control.''
    Date of issuance: July 27, 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.:  137, 131, 158, and 154
    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: June 23, 1993 (58 FR 
34071) 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: 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. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: December 14, 1994
    Brief description of amendments: The amendments revise the 
surveillance test intervals and allowed outage times for certain 
actuation instrumentation in the reactor protection, isolation, 
emergency core cooling, control rod withdrawal block, monitoring and 
feedwater/main turbine trip systems. The amendments also include 
changes to the feedwater/main turbine trip limiting condition for 
operation required actions, several mode related changes to the nuclear 
instrumentation and rod block specifications, shiftly channel check 
requirements for several systems, and several editorial changes to 
correct errors and remove outdated footnotes.
    Date of issuance: August 2, 1995
    Effective date: Immediately, to be implemented within 90 days.
    Amendment Nos.: 104 and 90
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11128) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 2, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room Location:  Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.

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

    Date of application for amendment: May 17, 1995, as supplemented on 
June 2, June 16, and July 12, 1995.
    Brief description of amendment: The amendment allows a limited 
number of steam generator tubes with roll transition indications to 
remain in service until the September 1995 refueling outage.
    Date of issuance: July 26, 1995
    Effective date:  July 26, 1995
    Amendment No.: 167
    Facility Operating License No. DPR-39: The amendment revises the 
Technical Specifications. The June 2, June 16, and July 12, 1995, 
submittals provided additional clarifying information that did not 
change the initial proposed no significant hazards consideration 
determination. The information, however, included changes to details of 
the administrative limits mentioned in the initial proposed no 
significant hazards consideration determination.Public comments 
requested as to proposed no significant hazards consideration 
determination: Yes (60 FR 27798). This notice provided an opportunity 
to submit comments on the Commission's proposed no significant hazards 
consideration determination. No comments have been received. The notice 
also provided for an opportunity to request a hearing by June 26, 1995, 
but indicated that if the Commission makes a final no significant 
hazards consideration determination any such hearing would take place 
after issuance of the amendment. The Commission's related evaluation of 
the amendments, finding of exigent circumstances and final no 
significant hazards consideration determination is contained in a 
Safety Evaluation dated July 26, 1995.
    Local Public Document Room Location:  Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.

Consumers Power Company, Docket No. 50-155, Big Rock Point Plant, 
Charlevoix County, Michigan

    Date of application for amendment:  December 15, 1994
    Brief description of amendment: The amendment revises Technical 
Specification 11.3.1.5 ACTION a. to eliminate the need to demonstrate 
that the actuation circuitry of the unaffected reactor depressurization 
system channels is operable. In addition, the amendment makes an 
editorial change to correct a typographical error.
    Date of issuance:  July 28, 1995
    Effective date: July 28, 1995
    Amendment No.: 115
    Facility Operating License No. DPR-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20516) The Commission's related evaluation of the amendment is 
contained in a Safety 

[[Page 42616]]
Evaluation dated July 28, 1995. No significant hazards consideration 
comments received: No.
    Local Public Document Room Location: North Central Michigan 
College, 1515 Howard Street, Petoskey, Michigan 49770.

Consumers Power Company, Docket No. 50-155, Big Rock Point Plant, 
Charlevoix County, Michigan

    Date of application for amendment: March 4, 1993, as revised April 
14, 1993, as supplemented April 19 and May 31, 1995
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) to conform to the wording of the revised 10 CFR 
Part 20, ``Standards for Protection Against Radiation,'' and to reflect 
a separation of chemistry and radiation protection responsibilities.
    Date of issuance: August 2, 1995
    Effective date: August 2, 1995
    Amendment No.:  16
    Facility Operating License No. DPR-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1993 (58 FR 
28053), as corrected June 1, 1993 (58 FR 31222). The supplemental 
submittals were noticed on June 21, 1995 (60 FR 32361). The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation datedNo significant hazards consideration comments 
received: No.
    Local Public Document Room Location: North Central Michigan 
College, 1515 Howard Street, Petoskey, Michigan 49770.

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van 
Buren County, Michigan

    Date of application for amendment: April 7, 1994, as 
supplementedApril 27, 1995.
    Brief description of amendment: This amendment relocates certain 
Technical Specifications (TS) that contain fuel cycle-specific 
parameter limits that change with core reloads to a Core Operating 
Limits Report. TS bases have also been revised to refer to limits 
relocated to the COLR. A portion of the amendment request was denied. A 
separate Notice of Denial of Amendment has been sent to the Federal 
Register for publication.
    Date of issuance: July 26, 1995
    Effective date: July 26, 1995
    Amendment No.: 169
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27053) The April 27, 1995, submittal provided clarifying information 
which was within the scope of the initial application and did not 
affect the staff's initial proposed no significant hazards 
consideration findings. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated July 26, 1995.No 
significant hazards consideration comments received: No.
    Local Public Document Room Location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.

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

    Date of application for amendments: April 12, 1995
    Brief description of amendments: The amendments delete Technical 
Specification (TS) 3/4.3.4, ``Turbine Overspeed Protection,'' and its 
associated Bases. The deletion of TS 3/4.3.4 and its Bases provides 
Duke Power Company the flexibility to implement the manufacturer's 
recommendations for turbine steam valve surveillance test requirements. 
These test requirements will be contained in the Selected Licensee 
Commitment Manual.
    Date of issuance: July 21, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance
    Amendment Nos.: 131 and 125
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 21, 1995 (60 FR 
32361) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 21, 1995. No significant 
hazards consideration comments received: No
    Local Public Document Room Location:  York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.

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

    Date of application for amendments:  January 18, 1995.
    Brief description of amendments: The amendments relocate the 
requirements for the seismic instrumentation, meteorological 
instrumentation, and loose-part detection system, and the associated 
Bases and surveillance requirements, from the TS to the Selected 
Licensee Commitment Manual (Chapter 16 of the FSAR). This will allow 
future changes to these controls to be performed under the provisions 
of 10 CFR 50.59. No changes are being made to the technical content of 
the affected TS pages.
    Date of issuance: July 24, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 132 and 126
    Facility Operating License Nos. NPF-35 and NPF-52: 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 July 24, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room Location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.

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

    Date of application for amendments: April 12, 1995
    Brief description of amendments: The amendments delete Technical 
Specification (TS) 3/4.3.4, ``Turbine Overspeed Protection,'' and its 
associated Bases. The deletion of TS 3/4.3.4 and its associated Bases 
provides Duke Power Company the flexibility to implement the 
manufacturer's recommendations for turbine steam valve surveillance 
test requirements. These test requirements will be contained in the 
Selected Licensee Commitment (SLC) Manual. The SLC Manual is Chapter 16 
of the Updated Final Safety Analysis Report.
    Date of issuance: August 2, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance
    Amendment Nos.: 156 and 138
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 21, 1995 (60 FR 
32362) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 2, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room Location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223.

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

    Date of application for amendments: September 28, 1994, as 
supplemented 

[[Page 42617]]
by letters dated May 3 and June 14, 1995.
    Brief description of amendments: The amendments revise Technical 
Specification Tables 3.3-3, 3.3-4, 3.3-5, and 4.3-2 of the Engineered 
Safety Features Actuation System Instrumentation tables to update the 
``Loss of Power'' function.
    Date of issuance: August 2, 1995
    Effective date: As of the date of issuance to be implemented within 
60 days, or 60 days after the completion date of the Unit 2 
modification, whichever is later.
    Amendment Nos.: 157 and 139
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 65811) The May 3 and June 14, 1995, letters provided clarifying 
information that did not change the scope of the September 28, 1994, 
application and the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated August 2, 1995. No 
significant hazards consideration comments received: No.
    Local Public Document Room Location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223.

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

    Date of application for amendments: January 18, 1995
    Brief description of amendments: The amendments delete selected 
Technical Specification (TS) requirements related to instrumentation 
from the TS, and relocate them to the Selected Licensee Commitment 
(SLC) Manual, with their associated Bases and surveillance 
requirements. No changes are being made to the technical content of the 
affected TS pages. Future changes to the SLC Manual (Chapter 16 of the 
Final Safety Analysis Report) will be controlled by the provisions of 
10 CFR 50.59. The relocated requirements include the following:
    TS 3/4.3.3.3, Seismic Instrumentation
    TS 3/4.3.3.4, Meteorological Instrumentation
    TS 3/4.10, Loose-Part Detection System
    Date of issuance: August 2, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance
    Amendment Nos.: 158 and 140
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11132) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 2, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room Location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223.

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:  February 4, 1994, as 
supplemented June 29, 1995.
    Brief description of amendments: These amendments modify the 
Technical Specifications (TSs) related to containment air locks (TSs 
1.8, 3/4.6.1.1 and 3/4.6.1.3) and associated Bases to make them as 
close to the NRC's Improved Standard Technical Specifications (NUREG-
1431) as the plant-specific design will permit. The changes in TS 3/
4.6.1.1 and 3/4.6.1.3 modify surveillance requirements and limiting 
conditions for operation and effect numerous administrative and format 
changes.
    Date of issuance: July 26, 1995
    Effective date: Units 1 and 2, as of the date of issuance and shall 
be implemented within 60 days.
    Amendment Nos.: 190 and 72
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Units 1 and 2 Technical Specifications, and the Unit 2 
License.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37070) The June 29, 1995 letter did not change the original no 
significant hazards consideration determination or expand the scope of 
the July 20, 1994 Federal Register notice.The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
July 26, 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 
ElectricStation, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: May 12, 1995
    Brief description of amendment: The amendment removed the specific 
scheduling requirements for Type A containment leakage rate tests from 
the Technical Specifications for Waterford 3 and replaced these 
requirements with a requirement to perform Type A, testing in 
accordance with Appendix J to 10 CFR Part 50. The proposed changes 
adopt the wording for primary containment integrated leak rate testing 
that is consistent with the requirements of the Combustion Engineering 
Improved Standard Technical Specifications (NUREG 1432). The proposed 
changes also include several administrative changes.
    Date of issuance: August 3, 1995
    Effective date: August 3, 1995, to be implemented within 60 days of 
issuance.
    Amendment No.: 110
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29876) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 3, 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.

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, 
Appling County, Georgia

    Date of application for amendments: October 13, 1994, as 
supplemented by letters dated January 13 and May 4, 1995.
    Brief description of amendments: The amendments revise the 
Technical Specifications to lower the anticipated transient without 
scram-recirculation pump trip (ATWS-RPT) setpoint by approximately 2 
feet 2 inches to minimize the potential for RPTs following reactor 
scram, and allow restarting the recirculation pump following an RPT 
when the temperature differential between the coolant at the reactor 
bottom head and the reactor steam dome cannot be obtained, provided 
certain conditions are met.
    Date of issuance: July 21, 1995
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: 196 and 136
    Facility Operating License Nos. DPR-57 and NPF-5. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 

[[Page 42618]]
65813). The January 13 and May 4, 1995, letters provided clarifying 
information that did not change the scope of the October 13, 1994, 
application and initial proposed no significant hazards consideration 
determination.The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 21, 1995. No significant 
hazards consideration comments received: No
    Local Public Document Room Location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513.

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

    Date of application for amendment:  June 1, 1995
    Brief description of amendment: The amendment revises the TMI-1 
Technical Specifications to allow the use of two zirconium-based 
advanced fuel rod cladding materials manufactured by the Babcock & 
Wilcox Fuel Company.

    Date of issuance:  July 24, 1995
    Effective date:  July 24, 1995
    Amendment No.:  194
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 21, 1995 (60 FR 
32366) The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated July 24, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room Location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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 13, 1993 as supplemented by letter 
dated January 31, 1995
    Brief description of amendment: The amendment revises Attachment 3 
of the license conditions to remove several license conditions 
pertaining to the Division I and II Transamerica Delaval, Inc. 
emergency diesel generators. The conditions pertain to engine overhaul 
frequency, maintenance and surveillance program, and inspection of 
crankshafts, cylinder heads, engine block, and turbochargers.
    Date of issuance: July 25, 1995
    Effective date: July 25, 1995
    Amendment No.: 82
    Facility Operating License No. NPF-47. The amendment revised the 
operating license.
    Date of initial notice in Federal Register: August 4, 1993 (58 FR 
41505) The additional information contained in the supplemental letter 
dated January 31, 1995, was clarifying in nature and thus, within the 
scope of the initial notice and did not affect the staff's proposed no 
significant hazards consideration determination.The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated July 25, 1995.No significant hazards consideration comments 
received. No.
    Local Public Document Room Location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803.

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: April 27, 1995, as supplemented by 
letters dated May 4 and 25, 1995.
    Brief description of amendments: The amendments revised the tables 
associated with Technical Specifications (TSs) 3/4.3.3.5, Remote 
Shutdown System, to eliminate the requirement for core exit 
thermocouples (CETs). The amendments also revised the tables associated 
with TS 3/4.3.3.6, Accident Monitoring Instrumentation, to require two 
operable channels of CETs, where each channel is required to have at 
least two operable CETs per core quadrant. Each channel is also 
required to have at least four operable CETs in at least one quadrant 
to support the operability of the subcooling margin monitors.
    Date of issuance: July 24, 1995
    Effective date: July 24, 1995
    Amendment Nos.: Unit 1 - Amendment No. 77; Unit 2 - Amendment No. 
66
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 21, 1995 (60 FR 
32366) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 24, 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.

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 2, 1995
    Brief description of amendments: The amendments revised Technical 
Specifications 3.4.2.2. and 3.7.1.1 (Table 3.7-2) by relaxing the lift 
setting tolerances of the pressurizer safety valves from plus or minus 
1% to plus or minus 2% and the main steam safety valves from plus or 
minus 1% to plus or minus 3%, respectively. In addition, a footnote was 
added to require that the pressurizer safety valves and main steam 
safety valves setpoint tolerances be restored to within plus or minus 
1% whenever a lift setting is determined to be outside plus or minus 1% 
following valve testing.
    Date of issuance:  July 25, 1995
    Effective date:  July 25, 1995, to be implemented within 30 days of 
issuance.
    Amendment Nos.: Unit 1 - Amendment No. 78; Unit 2 - Amendment No. 
67
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29877) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 25, 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.

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

    Date of application for amendment: March 7, 1995, as supplemented 
on June 7, 1995.
    Brief description of amendment: The amendment adds an Exception to 
Technical Specifications 3.6.A and 3.6.C. The Exception permits reduced 
component cooling water flow for short periods of time, while component 
cooling water heat exchangers are shifted.
    Date of issuance: July 24, 1995

[[Page 42619]]

    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 151
    Facility Operating License No. DPR-36: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24911) The June 7, 1995, submittal provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated July 24, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room Location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578.

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

    Date of application for amendment: May 24, 1995
    Brief description of amendment: The amendment permits an individual 
who does not have a current senior reactor operator (SRO) license for 
Millstone Unit 1 to hold the Operations Manager position. In this case, 
the Operations Manager position would require the individual to have 
previously held an SRO license at a boiling water reactor and the 
individual serving in the capacity of the Assistant Operations Manager 
to hold a current SRO license for Millstone Unit 1. In addition, the 
amendment renumbers the applicable sections.
    Date of issuance: July 24, 1995
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 83
    Facility Operating License No. DPR-21. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 21, 1995 (60 FR 
32370) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 24, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room Location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

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

    Date of application for amendment: April 18, 1995
    Brief description of amendment: The amendment allows the use of the 
ANSI/ANS 5.1-1979 decay heat model for the post-loss of coolant 
accident containment cooling analysis.
    Date of issuance: July 24, 1995
    Effective date: As of the date of issuance to be implemented 
immediately.
    Amendment No.: 84
    Facility Operating License No. DPR-21. Amendment revised the 
license.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24911). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 24, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room Location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

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

    Date of application for amendment: April 28, 1995
    Brief description of amendment: The amendment revises the diesel 
generator fuel oil testing that is performed on new fuel prior to the 
addition of new fuel to the storage tank.
    Date of issuance: July 26, 1995
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 118
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29881) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 26, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room Location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

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

    Date of application for amendments: November 14, 1994 as 
supplemented by letter dated April 10, 1995.
    Brief description of amendments: These amendments relocate Nuclear 
Review Board (NRB) review requirements, Independent Safety Engineering 
Group (ISEG) requirements, and certain review and audit requirements 
from the TS to the Peach Bottom Quality Assurance Program.
    Date of issuance: July 25, 1995
    Effective date: July 25, 1995
    Amendments Nos.: 208 and 212
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 65822) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 25, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room Location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

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

    Date of application for amendment: July 27, 1994, as supplemented 
May 26, July 10, and July 25, 1995
    Brief description of amendment: This amendment revises the Allowed 
Out-of-Service Times (AOTs) for Inoperable Station Service Water System 
(SSWS) pumps, inoperable safety Auxiliaries Cooling System (SACS) 
pumps, and inoperable Emergency Diesel Generators (EDGs). In addition, 
this amendment also allows on-line maintenance of the EDGs.
    Date of issuance: August 1, 1995
    Effective date: August 1, 1995
    Amendment No.: 75
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 
45033) The supplemental letters did not change the original no 
significant hazards consideration determination nor the original 
Federal Register notice. The 

[[Page 42620]]
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated August 1, 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. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of application for amendment: April 25, 1994, as supplemented 
July 24, 1995
    Brief description of amendment: This amendment eliminates the 
requirement from the Hope Creek Technical Specifications to perform 
Type C leak rate tests, in accordance with 10 CFR Part 50, Appendix J, 
of identified containment isolation valves that penetrate the primary 
containment and terminate below the minimum water level in the 
suppression chamber (torus). The valves are still subject to testing in 
accordance with the American Society of Mechanical Engineers Boiler and 
Pressure Vessel Code.
    Date of issuance: August 1, 1995
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 76
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29632) The supplemental letter did not change the original no 
significant hazards consideration determination nor the original 
Federal Register notice.The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated August 1, 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 Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
County, New Jersey

    Date of application for amendments: April 18, 1995
    Brief description of amendments: The amendments delete the 
quarterly leak rate test for the containment pressure-vacuum relief 
valves that is currently required because of the valves' resilient seat 
material. The changes are being made to accommodate replacement of the 
resilient valve seat material with a hard seat (metal-to-metal) design. 
The valves would remain in the 10 CFR Part 50, Appendix J, Type C leak 
rate test program.
    Date of issuance: August 1, 1995
    Effective date: Unit 1, As of the date of issuance, to be 
implemented prior to restart following the twelfth refueling outage; 
Unit 2, As of the date of issuance, to be implemented prior to restart 
following the current refueling outage.
    Amendment Nos.: 172 and 153
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 23, 1995 (60 FR 
27342) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 1, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room Location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.

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

    Date of application for amendments: August 26, 1994
    Brief description of amendments: These amendments revise Technical 
Specification 3/4.7.5, ``Control Room Emergency Air Cleanup System,'' 
to provide an exception to Limiting Condition for Operation 3.0.4 for 
Modes 5 and 6 and for a defueled configuration. These amendments also 
add the applicability statement ``or during movement of irradiated fuel 
assemblies.''
    Date of issuance: July 26, 1995
    Effective date: July 26, 1995
    Amendment Nos.: Unit 2 - Amendment No. 123; Unit 3 - Amendment No. 
112
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55891) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 26, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room Location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.

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

    Date of application for amendments:  December 16, 1994; 
supplemented July 19, 1995 (TS 94-06)
    Brief description of amendments: The amendments replace the present 
Auxiliary Feedwater system Specification 3/4.7.1.2 with new 
specifications that are modeled after the Westinghouse Standard 
Technical Specifications.
    Date of issuance: August 2, 1995
    Effective date: August 2, 1995
    Amendment Nos.: 206 and 196
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: February 1, 1995 (60 FR 
6309) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 2, 1995.No significant hazards 
consideration comments received: None
    Local Public Document Room Location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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

    Date of application for amendments: November 29, 1994
    Brief description of amendments: These amendments allow the use of 
ZIRLO, a new zirconium-based alloy, as a fuel cladding material.
    Date of issuance: July 27, 1995
    Effective date: July 27, 1995
    Amendment Nos.: 202 and 202
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 4, 1995 (60 FR 
508) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 27, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room Location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.

Notice Of Issuance Of Amendments to facility Operating Licenses And 
Final Determination Of No Significant Hazards Consideration And 
Opportunity For A Hearing (Exigent Public Announcement Or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the 

[[Page 42621]]
Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By September 15, 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 

[[Page 42622]]
a supplement which satisfies these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

    Date of application for amendment: July 28, 1995
    Brief description of amendment: This amendment deletes the portion 
of License Condition 2.C.(1) that references Attachment 1. Attachment 1 
requires the pump in the keepwarm system on the emergency diesel 
generator to satisfy the requirements of the American Society of 
Mechanical Engineers Code, Section III, Class 3.
    Date of issuance: August 3, 1995I11Effective date: August 3, 1995
    Amendment No.: 88
    Facility Operating License No. NPF-42: The amendment revised the 
operating license.Public comments requested as to proposed no 
significant hazards consideration: No.The Commission's related 
evaluation of the amendment, finding of emergency circumstances, and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated August 3, 1995.
    Local Public Document Room Location: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman
    Dated at Rockville, Maryland, this 16th day of August 1995.
    For The Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects - III/IV Office of Nuclear 
Reactor Regulation
[Doc. 95-20122 Filed 8-15-95; 8:45 am]
BILLING CODE 7590-01-F