[Federal Register Volume 60, Number 90 (Wednesday, May 10, 1995)]
[Notices]
[Pages 24904-24934]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-40510]



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UNITED STATES 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 April 17, 1995, through April 28, 1995. The 
last biweekly notice was published on April 26, 1995. [[Page 24905]] 

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 June 9, 1995, the licensee may file a request for a hearing with 
respect to issuance of the amendment to the subject facility operating 
license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public 
[[Page 24906]] 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.

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

    Date of amendment request: April 5, 1995
    Description of amendment request: The licensee proposes to revise 
Technical Specification (TS) 3/4.9, Refueling Operations, to be 
consistent with NUREG-1431, Standard Technical Specifications, 
Westinghouse Plants, and to relocate the applicable sections from the 
TS that do not meet the Commission's screening criteria for retention.
    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:
    This change does not involve a significant hazards consideration 
for the following reasons:
    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes will have no significant impact on the 
safety, reliability, or operation of fuel handling equipment or 
activities. These changes will simplify the Technical Specifications 
and implement the recommendations of the Commission's Final Policy 
Statement on Technical Specification Improvements based upon the 
assumptions and analyses contained in the bases of NUREG-1431. Those 
elements that involve relocations to plant procedures are 
administrative in nature and do not involve any modifications to 
plant equipment or operation. Therefore, there would be no increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not introduce any new equipment or 
require existing equipment to operate to perform a function 
different from that previously evaluated in the Final Safety 
Analysis Report or Technical Specifications. The changes are 
consistent with the new Standard Techical Specification and 
assumptions contained in NUREG-1431 and in the Commission's Final 
Policy Statement on Technical Specification Improvements. Therefore, 
the proposed changes would not increase the possibility of a new or 
different type of accident from any accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed changes do not affect any of the parameters which 
relate to the margin of safety as described in the [Bases] of the 
Technical Specifications or the Final Safety Analysis Report. 
Accordingly, NRC Acceptance Limits are not affected by these 
changes. For those specifications being relocated to other plant 
documents, these changes are purely administrative. Therefore, the 
proposed changes do not involve a significant reduction in the 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: David B. Matthews

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

    Date of application for amendment request: September 15, 1992, as 
supplemented April 21, 1995
    Description of amendment request: As a result of findings by a 
Diagnostic Evaluation Team inspection performed by the NRC staff at the 
Dresden Nuclear Power Station in 1987, Commonwealth Edison Company 
(ComEd, the licensee) made a decision that both the Dresden Nuclear 
Power Station and sister site Quad Cities Nuclear Power Station, needed 
attention focused on the existing custom Technical Specifications 
(TSs).
    The licensee made the decision to initiate a Technical 
Specification Upgrade Program (TSUP) for both Dresden and Quad Cities. 
The licensee evaluated the current TSs for both Dresden and Quad Cities 
against the Standard Technical Specifications (STSs) 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 TSs 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 TSs to the STSs contained in NUREG-
0123.
    The TSUP for Dresden and Quad Cities is not a complete adaption of 
the STSs. The TSUP focuses on (1) integrating additional information 
such as equipment operability requirements during shutdown conditions, 
(2) clarifying requirements such as limiting conditions for operations 
and action statements utilizing STS terminology, (3) deleting 
superseded requirements and modifications to the TSs based on the 
licensee's responses to Generic Letters (GLs), and (4) relocating 
specific items to more appropriate TS locations.
    The application dated September 15, 1992, as supplemented April 21, 
1995, proposed to upgrade only Sections 2.0 (Safety Limits and Limiting 
Safety System Settings), 3/4.11 (Power Distribution Limits), and 3/4.12 
(Special Test Exceptions) of the Dresden and Quad Cities TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the [[Page 24907]] issue of no significant 
hazards consideration, which is presented below:
    Section 2.0
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because:
    The proposed changes to Specifications 1/2.1 and 1/2.2 to delete 
the present Applicability and Objective sections represent 
administrative changes to format and presentation of material. The 
proposed changes provide the user with a format that will allow 
better access to needed information and provides concise Safety 
Limit, Limiting Safety System Settings, Applicability and Action 
requirements. The additions of Applicability and Action requirements 
represent clarification of intended requirements that do not 
presently state all required conditions of operability or provide 
clearly stated Action statements if the requirements are not met. 
The combining of the two sections and added requirements follow STS 
guidelines that are in use at many operating BWRs with similar 
design and operating configurations as Dresden and Quad Cities 
Stations. Operability requirements for Safety Limits have been 
chosen to reflect only those Operational Modes where the Safety 
Limits apply. Operability requirements for Limiting Safety System 
Settings are already stated in other sections of the Technical 
Specifications, thus reference to the appropriate operability 
requirement is made rather than repeating the requirement in the 
Limiting Safety System Setting Specification.
    Deletion of the Power Transient Safety Limit does not impact any 
safety analyses. The safety analyses assume the Reactor Protection 
System (RPS) operates as designed and the reactor scrams when the 
neutron flux exceeds the limiting safety system setting. The 
proposed Technical Specifications will continue to provide a highly 
reliable system to operate as assumed in the safety analyses. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The reactor water level low scram setpoint is changed (for Quad 
Cities) to be consistent with other reactor water level setpoints in 
the Technical Specifications and the STS. The setpoint is equivalent 
to the current requirement but is expressed as the reactor water 
level above the top of active fuel.
    The scram discharge volume scram level is converted for Dresden 
Unit 2 and Unit 3 to gallons to be consistent with the Quad Cities 
Units. The proposed setpoints are consistent with the current 
specifications. The change in the units does not represent a change 
in the physical setpoint.
    The proposed change to delete the APRM Downnscale Scram trip 
function for Quad Cities has been evaluated by Commonwealth Edison 
and General Electric and previously approved for Dresden Station. 
The events of concern with respect to the APRM/IRM companion trip 
are the Control Rod Drop Accident and the low power Rod Withdrawal 
Error. The FSAR and reload safety analyses do not credit this scram 
function in the termination of either of these events. Since this 
scram function is not credited in the termination of these events, 
the elimination of this scram function has no adverse effect on 
previously evaluated accidents.
    The change to the low condenser vacuum scram setpoint from 23 
inches Hg to 21 inches of Hg is consistent with an identical change 
made to Quad Cities Units 1 and 2. The low condenser vacuum scram is 
an anticipatory scram and is not credited in any transient analysis. 
Thus the reduction in the setpoint will not affect any transient 
analysis.
    The proposed changes do not alter the intent of existing 
setpoints or accident assumptions and follow existing requirements 
at other operating BWRs for operability and Action statements. 
Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any previously evaluated because:
    The proposed administrative changes to the format and 
arrangement of material do not affect technical requirements or 
assumptions of any potential accident and; therefore, cannot create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    The proposed addition of Applicability and Action requirements 
enhance the understanding and usability of the Technical 
Specifications and thus represent an improvement over present 
specifications. New requirements are modeled after those in use at 
operating BWRs and do not represent requirements that will adversely 
affect potential accident analyses or assumptions. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    Deletion of the Power Transient Safety Limit does not involve a 
change in the design or operation of any systems assumed to operate 
in the safety analyses. Therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The change in the units for the Reactor Water Level scram 
function do not change any physical plant setpoints. The setpoint 
will remain the same but will be expressed as the level above the 
top of active fuel. The change does not create the possibility of a 
new or different kind of accident.
    The conversion of the Scram Discharge Volume scram setpoint from 
inches to gallons does not alter any physical plant setpoints. The 
setpoint will remain the same but will be expressed in gallons 
rather than inches. The change will provide consistency between 
Dresden and Quad Cities.
    The deletion of the APRM Downscale Scram Trip Function does not 
introduce any new accident. The limiting accidents, Control Rod 
Drop, Rod Withdrawal Error, in the operating region of transition 
between the Startup and Run Operational Modes are well understood 
and are evaluated in FSAR and reload analyses. Other control rod 
initiated events which are less limiting in this region are subsets 
of the low power Rod Withdrawal Error event and are bounded by it 
and the design basis Control Rod Drop Accident. General Electric has 
indicated that, for reactivity insertion mechanisms at very low 
power, the only effect of the deletion of the APRM downscale scram 
would be that the initial power level could be a few percent lower 
which would not have a significant effect on the severity of the 
event. In addition, proper overlap between the IRMs and APRMs is not 
affected since the calibration requirements are not being changed.
    The change in the low condenser vacuum scram function will not 
create the possibility of a new or different kind of accident 
because the function is not recognized in any of the transient 
analysis. The low condenser vacuum scram function is an anticipatory 
scram.
    The proposed changes do not involve a significant reduction in 
the margin of safety because:
    The proposed administrative changes to format, arrangement of 
material, clarification of requirements and other non-technical 
changes do not affect any safety aspects of the plant and as such 
can not involve a significant reduction in the margin of safety.
    The proposed Applicability statements require availability of 
Safety Limits and Limiting Safety System Settings when required to 
perform their respective functions. Proposed Actions for Safety 
Limits allow only 2 hours to be in Hot Shutdown and then reference 
Specification 6.4 to ensure that proper reports are made and restart 
is prohibited until approved by the NRC. These provisions help 
ensure that present margins are not significantly reduced.
    Deletion of the Power Transient Safety Limit does not impact the 
margin assumed in the safety analyses. The safety analyses assume 
the RPS operates as designed and the reactor scrams when the neutron 
flux exceeds the limiting safety system setting. The margins assumed 
in the design of the RPS and in the safety and transient analyses 
calculations have not been revised. Therefore, this change does not 
involve a significant reduction in the margin of safety.
    The change in units to the Reactor Water Level scram setpoint 
and the Scram Discharge Volume scram setpoint do not involve a 
significant reduction in the margin of safety because the changes do 
not represent a change in the physical setpoints.
    The reduction in the Low Condenser Vacuum scram setpoint does 
not represent a reduction in the margin of safety because the scram 
is not credited in any transient analysis.
    The APRM Downscale Scram Trip Function is not credited in the 
termination of any FSAR or reload safety analysis event. As such, 
the elimination of this scram function has no effect on any margin 
of safety.
    Section 3/4.11
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because:
    In general, the proposed changes represent the conversion of 
current requirements to a [[Page 24908]] 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. These proposed 
changes are consistent with the current safety analyses and have 
been previously determined to represent sufficient requirements for 
the assurance of 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 Generic Changes to the technical specifications involve 
administrative changes to format and arrangement of the material. As 
such, these changes cannot involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The current specifications require the reactor to be placed in 
cold shutdown when a thermal limit was exceeded and not restored 
within the allotted 2 hours, but the proposed specifications require 
the reactor to be less than 25% of rated thermal power if this 
condition occurred. The change eliminates a shutdown and requires 
the power level to be reduced to the point that the limits are no 
longer applicable.
    Therefore, the change will not increase the probability or 
consequences of an accident.
    Create the possibility of a new or different kind of accident 
from any previously evaluated because:
    In general, the proposed changes represent the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis. Others 
represent minor curtailments of the current requirements which are 
based on generic guidance or previously approved provisions for 
other stations. These changes do not involve revisions to the design 
of the station. Some of the changes may involve revision in the 
operation of the stations; however, these changes provide additional 
restrictions which are in accordance with the current safety 
analyses, or are to provide for additional testing or surveillance 
which will not introduce new failure mechanisms beyond those already 
considered in the current safety analyses. Therefore, these changes 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Since the Generic Changes proposed to the technical 
specifications are administrative in nature, they cannot create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The requirement to reduce thermal power to less than 25% of 
rated thermal power rather than place the reactor in cold shutdown 
will not create a new or different kind of accident because the 
thermal limits are not required in operational mode 1 when thermal 
power is less than 25% of rated power.
    Involve a significant reduction in the margin of safety because:
    In general, the proposed changes represent the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis. Others 
represent minor curtailments of the current requirements which are 
based on generic guidance or previously approved provisions for 
other stations. Some of the 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 within their acceptance limits. These enhancements 
compensate for the individual minor reductions, such that taken 
together, the proposed changes will not significantly reduce the 
margin of safety.
    The Generic Changes proposed in this amendment request are 
administrative in nature and, as such, do not involve a reduction in 
the margin of safety.
    Section 3/4.12
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because:
    The proposed Specification 3/4.12 is a new section which will 
provide the user with a format that will allow better access to 
needed information and provide concise Applicability and Action 
requirements. The additions of Applicability and Action requirements 
represent classification of intended requirements that do not 
presently state all required conditions of operability or provide 
clearly stated Action statements if the requirements are not met. 
The combining of the two sections and the added requirements follow 
Standard Technical Specifications (STS) guidelines that are in use 
at many operating BWRs with similar design and operating 
configurations as Dresden and Quad Cities Stations.
    The proposed Section 3/4.12 involves the relocation of present 
requirements into one section identical to STS provisions. The 
changes also implement the Applicability and Action provisions of 
the STS and later operating BWR plants that have been evaluated and 
found acceptable for use at Dresden and Quad Cities. Present 
Surveillance Requirements are replaced, where applicable, with 
proven STS guidelines that are being used at plants with a system 
similar to that at Dresden and Quad Cities. The changes in the 
present Surveillance Requirements add testing requirements that are 
not presently in the Dresden and Quad Cities technical 
specifications. The proposed changes do not affect accident 
assumptions other than a minor increase in the initial power level 
(approximately 0.2% to 1%) and as such, do not involve a significant 
increase in the probability of an accident previously evaluated. The 
proposed specifications add additional requirements to 
specifications currently contained in the Technical Specifications. 
Since the proposed changes to the Technical Specifications implement 
requirements that have been demonstrated to provide acceptable 
operability provisions at other facilities with a design similar to 
that at Dresden and Quad Cities, the proposed changes do not 
significantly increase the consequences of an accident previously 
evaluated.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any previously evaluated because:
    The proposed administrative changes to the format and 
arrangement of material do not affect technical requirements or 
assumptions of any potential accident and; therefore, cannot create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    The proposed addition of Applicability and Action requirements 
enhance the understanding and usability of the Technical 
Specifications and thus represent an improvement over present 
specifications. New requirements are modeled after those in use at 
operating BWRs and do not represent requirements that will adversely 
affect potential accident analyses or assumptions. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    The proposed changes do not involve a significant reduction in 
the margin of safety because:
    The proposed administrative changes to format, arrangement of 
material, clarification of requirements and other non technical 
changes do not affect any safety aspects of the plant and as such 
can not involve a significant reduction in the margin of safety.
    In addition, the commission has provided guidance concerning the 
application of standards for determining whether significant hazards 
consideration exists by providing certain examples (51 FR 7751) of 
amendments that are considered not likely to involve significant 
hazards considerations. Commonwealth Edison has reviewed the 
proposed changes against these examples and believes that the 
proposed changes fall within the scope of example (ii) ``a change 
that constitutes an additional limitation, restriction, or control 
not presently included in the technical specifications''.
    The proposed amendment does not involve a significant relaxation 
of the criteria used to establish safety limits, a significant 
relaxation of the bases for the limiting safety system settings or a 
significant relaxation of the bases for the limiting conditions for 
operations. Therefore, based on the guidance provided in the Federal 
Register and the criteria established in 10 CFR 50.92(c), the 
proposed change does not constitute a significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this [[Page 24909]] review, it appears that the three standards of 10 
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

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

    Date of application for amendment request: December 15, 1993, as 
supplemented by letter dated April 21, 1995
    Description of amendment request: As a result of findings by a 
Diagnostic Evaluation Team inspection performed by the NRC staff at the 
Dresden Nuclear Power Station in 1987, Commonwealth Edison Company 
(ComEd, the licensee) made a decision that both the Dresden Nuclear 
Power Station and sister site Quad Cities Nuclear Power Station, needed 
attention focused on the existing custom Technical Specifications (TSs) 
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 TSs for both Dresden and Quad Cities 
against the Standard Technical Specifications (STSs) 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 TSs 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 TSs to the STSs contained in NUREG-
0123.
    The TSUP for Dresden and Quad Cities is not a complete adaption of 
the STSs. The TSUP focuses on (1) integrating additional information 
such as equipment operability requirements during shutdown conditions, 
(2) clarifying requirements such as limiting conditions for operations 
and action statements utilizing STS terminology, (3) deleting 
superseded requirements and modifications to the TSs based on the 
licensee's responses to Generic Letters (GLs), and (4) relocating 
specific items to more appropriate TS locations.
    The December 15, 1993, and April 21, 1995, applications proposed to 
upgrade only Section 5.0 (Design Features) of the Dresden and Quad 
Cities TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because:
    In general, the proposed amendment represents the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis. 
Implementation of these changes will provide continued assurance 
that specified [parameters remain] within their acceptance limits, 
and as such, will not significantly increase the probability or 
consequences of a previously evaluated accident. Some of the 
proposed changes to the current Technical Specifications (CTS) 
represent minor curtailments of the current requirements which are 
based on generic guidance or previously approved provisions for 
other stations. The proposed amendment for current Dresden and Quad 
Cities Station's Technical Specifications Section 5.0 represent a 
minor relaxation of the current requirements, and is based on BWR-
STS (NUREG-0123) guidelines or later operating BWR plant's NRC 
accepted changes. The proposed changes are consistent with the 
current safety analyses and have been previously determined to 
represent sufficient requirements for the assurance and reliability 
of equipment assumed to operate in the safety analysis. Any 
deviations from CTS or STS requirements do not significantly 
increase the probability or consequences of any previously evaluated 
accidents for Dresden or Quad Cities Stations.
    Details describing the plant's design are presented in TSUP 
Section 5.0. There are no Limiting Conditions for Operation (LCO) or 
Surveillance Requirements (SR) encompassed within TSUP Section 5.0. 
This information is administrative in nature and consistent to the 
UFSAR; therefore, the probability of any accident previously 
evaluated is not increased by the proposed amendment.
    Create the possibility of a new or different kind of accident 
from any previously evaluated because:
    In general, the proposed amendment represents the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis. Others 
represent minor relaxations 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. The proposed changes are administrative in nature 
and do not involve a revision in the operation of the station. As 
such, there are no changes to the current safety analysis. 
Therefore, the proposed changes 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 Specifications Section 5.0 is based on BWR-STS guidelines 
or later operating BWR plants' NRC accepted changes. The proposed 
amendment has been reviewed for acceptability at the Dresden or Quad 
Cities Nuclear Power Stations considering similarity of system or 
component design versus the BWR-STS or later operating BWRs. Any 
deviations from CTS or BWR-STS requirements do not create the 
possibility of a new or different kind of accident previously 
evaluated for Dresden and Quad Cities Stations. No new modes of 
operation are introduced by the proposed changes. The proposed 
changes maintain at least the present level of operability, and in 
some cases are more conservative. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    Involve a significant reduction in the margin of safety because:
    In general, the proposed amendment represents the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis. Others 
represent minor curtailments of the current requirements which are 
based on generic guidance or previously approved provisions for 
other stations. The proposed amendment to Technical Specification 
Section 5.0 implements present requirements, or the intent of 
present requirements in accordance with the guidelines set forth in 
the STS. Any deviations from CTS or BWR-STS requirements do not 
significantly reduce the margin of safety for Dresden or Quad Cities 
Stations. These changes do not involve revisions to the design of 
the station. The proposed changes are administrative in nature and 
do not involve a revision in the operation of the station. As such, 
there are no changes to the current safety analysis. Therefore, the 
proposed changes will not introduce new failure mechanisms beyond 
those already considered in the current safety analyses. Therefore, 
because the proposed changes are administrative in nature, do not 
involve a revision in the operation of the station and maintains the 
current design requirements specified in the UFSAR, the proposed 
changes do not involve a significant reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards 
consideration. [[Page 24910]] 
    Local Public Document Room location: For Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: December 13, 1994
    Description of amendment request: The proposed amendment would 
revise the Palisades' technical specifications (TSs) to add a high 
thermal performance (HTP) departure from nucleate boiling correlation 
to Safety Limit 2.1. The HTP correlation is used for the high thermal 
performance fuel loaded during recent fuel cycles.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to the TS adds the HTP critical heat flux 
correlation to the Safety Limit - Reactor Core Section 2.1. The HTP 
correlation is an NRC approved methodology for a Departure from 
Nucleate Boiling (DNB) Correlation for high thermal performance 
(HTP) fuel as is used at Palisades. The HTP correlation is an 
extension of the currently approved ANFP correlation. There are no 
associated changes in plant operation. Palisades fuel loaded in 
cycle 9 and later meet the requirements of the HTP correlation. 
Therefore, operation of the facility in accordance with the proposed 
TS would not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The HTP correlation will allow for more accurate DNB predictions 
within the applicable operating conditions for fuels with the HTP 
design used at Palisades. There are no changes in plant operation. 
Therefore operation of the facility in accordance with the proposed 
TS would not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    As stated previously, the HTP correlation will allow for more 
accurate DNB predictions within the applicable operating conditions 
for fuel with the HTP design. There are no associated changes in 
plant operation. Therefore, operation of the facility in accordance 
with the proposed TS 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: Van Wylen Library, Hope 
College, Holland, Michigan 49423.
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
    NRC Project Director: Cynthia A. Carpenter, Acting

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: January 18, 1995
    Description of amendment request: The proposed amendments would 
relocate the requirements for the seismic instrumentation, 
meteorological instrumentation, and loose-part detection system from 
the Technical Specifications to the Selected Licensee Commitment (SCL) 
Manual. 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 Technical Specification 
pages.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1
    The requested amendments will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. Relocation of the affected TS sections to the SLC Manual 
will have no effect on the probability of any accident occurring. In 
addition, the consequences of an accident will not be impacted since 
the above instrumentation will continue to be utilized in the same 
manner as before. No impact on the plant response to accidents will 
be created.
    Criterion 2
    The requested amendments will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. No new accident causal mechanisms will be created as a 
result of relocating the affected TS requirements to the SLC Manual. 
Plant operation will not be affected by the proposed amendments and 
no new failure modes will be created.
    Criterion 3
    The requested amendments will not involve a significant 
reduction in a margin of safety. No impact upon any plant safety 
margins will be created. Relocation of the affected TS requirements 
to the SLC Manual is consistent with the content of the Westinghouse 
RSTS [Revised Standard Technical Specifications], as the NRC did not 
require technical specification controls for the affected 
instrumentation in the RSTS. The proposed amendments are consistent 
with the NRC philosophy of encouraging utilities to propose 
amendments that are consistent with the content of the RSTS.
    Based upon the preceding analyses, Duke Power Company concludes 
that the requested amendments do not involve a significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: April 3, 1995
    Description of amendment request: The amendments will incorporate 
line-item TS improvements to Specifications 3/4.8.1 ``Electrical Power 
Systems-A.C. Sources,'' and 4.8.1.2.2 ``Electrical Power Systems-
Shutdown.'' The proposed changes are consistent with recommendations 
for Emergency Diesel Generator (EDG) Surveillance Requirements in 
NUREG-1366, and regulatory guidance provided in Generic Letter (GL) 93-
05 and GL 94-01. This proposal also contains FPL's commitment to 
implement a maintenance program for monitoring and maintaining EDG 
performance for both St. Lucie Units consistent with 10 CFR 50.65 and 
the guidance of Regulatory Guide 1.160.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Operation of the facility in accordance with the proposed 
amendment would not [[Page 24911]] involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The license amendments proposed for St. Lucie Units 1 and 2 will 
incorporate line-item Technical Specification (TS) improvements for 
Emergency Diesel Generators (EDG) pursuant to guidance provided in 
Generic Letters (GL) 93-05 and 94-01. The EDGs are not accident 
initiators, the proposed TS changes do not involve any assumptions 
relative to accident initiators in the plant safety analyses, and 
therefore the proposed amendments will not impact the probability of 
occurrence for accidents previously analyzed.
    The EDG line-item TS improvements associated with GL 93-05 are 
based on recommendations designed to remove unwarranted requirements 
for testing during power operation and other factors that are 
counter-productive to safety in terms of equipment degradation and 
availability. These recommendations resulted from a comprehensive 
study of industry-wide EDG surveillance requirements and subsequent 
findings reported by the NRC in NUREG-1366. The proposed amendments 
are consistent with the GL 93-05 guidance for implementing such 
recommendations.
    Similarly, GL 94-01 provides guidance for a line-item TS 
improvement that will remove accelerated testing requirements from 
the TS provided that the licensee commits to a maintenance program 
for monitoring and maintaining EDG performance that includes the 
applicable provisions of the maintenance rule (10 CFR 50.65). Such a 
program will further assure EDG availability. Since the availability 
of EDGs is assumed in certain success paths for mitigating analyzed 
accidents, an improvement in EDG availability will enhance accident 
mitigation capabilities.
    Therefore, operation of the facility in accordance with the 
proposed amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendments incorporate line-item TS improvements to 
EDG surveillance testing requirements, and will not change the 
physical plant or the modes of plant operation defined in the 
Facility License. The changes do not involve the addition or 
modification of equipment, nor do they alter the design or methods 
of operation of plant systems. Plant configurations that are 
prohibited by TS will not be created by the amendments. Therefore, 
operation of the facility in accordance with the proposed amendment 
would not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed amendments are designed to improve EDG availability 
by eliminating unwarranted surveillance testing. The presently 
specified surveillance intervals are not changed. The proposed 
changes do not otherwise alter the basis for any technical 
specification that is related to the establishment of, or the 
maintenance of a nuclear safety margin. Therefore, operation of the 
facility in accordance with the proposed amendment would not involve 
a significant reduction in a margin of safety.
    Based on the above discussion and the supporting Evaluation of 
Technical Specification changes, FPL has determined that the 
proposed license amendment involves no significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036
    NRC Project Director: David B. Matthews, Director

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

    Date of amendment request: March 7, 1995
    Description of amendment request: The proposed amendment would add 
an Exception to Technical Specifications (TS) 3.6.A and 3.6.C. The 
Exception would permit reduced component cooling water flow for short 
periods of time, while component cooling water heat exchangers are 
shifted.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The staff's review is 
presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Plant experience shows that the component cooling water heat 
exchangers can be shifted in a few minutes; well within the time limit 
for Remedial Action under this TS 3.6.A or C, or TS 3.0.A. Thus, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change does not affect equipment reliability when such 
equipment is required to be operable. Existing TS 3.6 and its Remedial 
Action statement govern the plant circumstances under which cooling 
water subsystems are required, and specify the maximum time such 
subsystems may be unavailable. The proposed change does affects neither 
operating requirements nor the time limit on restoring system 
operability.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed change does not significantly alter the availability 
or condition of the cooling water subsystems and, therefore, does not 
alter the accident analysis or its associated conclusions. The proposed 
change would permit flow in one component cooling water train to be 
reduced below that required for operation of the emergency core cooling 
systems in the recirculation mode, for a short period of time. The 
amount of time that flow is reduced is small, and full flow operation 
can be easily restored within the time required for design heat load 
removal. Thus, there is no significant reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that this amendment request involves no significant hazards 
consideration.
    Local Public Document Room location:  Wiscasset Public Library, 
High Street, P.O. Box 367, Wiscasset, ME 04578
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, 329 Bath Road, Brunswick, ME 04011
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: April 18, 1995
    Description of amendment request: The proposed amendment would 
allow the use of the ANSI/ANS 5.1-1979 decay heat model for post-loss 
of coolant accident containment cooling analysis.
    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 
[[Page 24912]] consideration, which is presented below:
    NNECO has reviewed the proposed change in accordance with 
10CFR50.92 and concluded that the change does not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
compromised. The proposed change does not involve an SHC because the 
change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The change to the decay heat model used to determine post-
accident conditions cannot affect the probability of any accident. 
No changes to plant operation or design would occur due to the new 
analysis.
    The new model cannot directly affect the consequences of an 
accident, since it is the tool used to predict the temperature 
effects of the postulated accident. However, using the ANSI/ANS 5.1-
1979 model could change the anticipated actions necessary to respond 
to an event. Changing the response action could possibly affect the 
consequences of an accident. This model change will not have such an 
effect. Operator actions to throttle LPCI [low pressure coolant 
injection], CS [core spray], or ESW [emergency service water] pump 
flow are taken based upon observed conditions, not predetermined 
data points from the analysis.
    Operability of the emergency core cooling systems (ECCS) can be 
shown for temperatures that are higher than those predicted by the 
containment cooling analysis.
    Therefore, the utilization of the ANSI/ANS 5.1-1979 decay heat 
model does not involve a significant increase in the probability or 
consequences of a previously evaluated accident.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed license amendment only revises the predicted 
temperature that result from a postulated accident. There is no 
change to the design or operation of any system or component. Since 
this change only deals with the post-accident effects of currently 
analyzed accidents, there is no possibility of creating a new or 
different kind of accident.
    3. Involve a significant reduction in the margin of safety.
    The early design documentation stated that the ECCS components 
were designed for post-accident torus temperatures of 203 deg.F. As 
this issue evolved, NNECO performed operability determinations which 
showed that peak temperatures of 209 deg.F were acceptable. 
Utilizing a more accurate decay heat model which results in lower 
predicted peak temperatures demonstrates the acceptability of the 
plant design. Therefore, replacing the May-Witt decay heat model 
with the ANSI/ANS 5.1-1979 model does not result in a decrease in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: March 29, 1995
    Description of amendment request: The proposed amendment changes 
Technical Specifications to revise peaking factor penalties based on 
NRC approved methods.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed changes do not involve an SHC because the changes 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed changes to the action statements of Sections 
3.2.2.1 and 3.2.2.2 are purely administrative and therefore they do 
not adversely affect the probability or consequences of an accident 
previously analyzed. The proposed changes to Surveillance 
Requirements 4.2.2.1.2.e, 4.2.2.1.4.e, 4.2.2.2.2.e and 4.2.2.2.4.e 
and Section 6.9.1.6.b are based on the NRC approved methodology for 
calculating the penalty to be applied to FQM(Z). The 
margin for the FQRTP limit is still maintained by the 
proposed changes. In addition, the penalty is included in the COLR 
[Core Operating Limits Report] which will be maintained and 
controlled per the requirements of 10CFR50.59. Therefore, the 
proposed changes do not increase the probability or consequences of 
an accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed changes to the Action Statement of Sections 3.2.2.1 
and 3.2.2.2 are purely administrative and therefore, they do not 
create the possibility of a new or different kind of accident from 
any previously analyzed. The proposed changes to Surveillance 
Requirements 4.2.2.1.2.e, 4.2.2.1.4.e, 4.2.2.2.2.e, and 4.2.2.2.4.e 
and Section 6.9.1.6.b do not create a malfunction that is different 
from those previously evaluated. The changes do not involve 
positioning reactivity systems or plant components into any new 
configuration or sequence not previously analyzed. Therefore, the 
changes will not create the possibility of a new or different kind 
of accident from any other previously analyzed.
    3. Involve a significant reduction in the margin of safety.
    The proposed changes to the action statements of Sections 
3.2.2.1 and 3.2.2.2 are purely administrative and therefore they 
will not reduce the margin of safety. The proposed changes to 
Surveillance Requirements 4.2.2.1.2.e, 4.2.2.1.4.e, 4.2.2.2.2.e and 
4.2.2.2.4.e and Section 6.9.1.6.b do not reduce the margin to the 
FQRTP limit. The approved methods more distinctly evaluate 
the expected changes to FQM than previously existed. 
Therefore, there is no impact on the margin of safety as specified 
in the Technical Specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee

PECO Energy Company, Public Service Electric and Gas Company, 
Delmarva Power and Light Company, and Atlantic City Electric 
Company, Docket No. 50-277, Peach Bottom Atomic Power Station, Unit 
No. 2, York County, Pennsylvania

    Date of application for amendment: March 30, 1995
    Description of amendment request: The proposed change would revise 
Technical Specifications Section 4.7.D.1.b.(1) by adding a footnote to 
exempt the High Pressure Coolant Injection [HPCI] motor-operated valve 
MO-2-23-015 from quarterly stoke testing requirements until refueling 
outage 2RO11.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or [[Page 24913]] consequences of an accident 
previously evaluated.
    The proposed change does not serve as an initiator or 
contributor to any accidents previously evaluated. It does not 
decrease the effectiveness of equipment relied upon to mitigate 
previously evaluated accidents. A calculation was performed and it 
has been determined the leakage through the valve's packing will be 
within the allowable limits of containment leakage (La). While 
positioning the valve in the backseated position does increase its 
stroke time, it has been calculated and demonstrated that the valve 
will close within the TS time limit of 20 seconds.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change does not serve as an initiator or 
contributor to any of the accidents previously evaluated. The 
proposed change does not introduce any new modes of plant operation.
    Implementation of the proposed changes will not affect the 
design function or configuration of any component or introduce any 
new operating scenarios or failure modes or accident initiation. It 
does not impair or prevent safety systems from performing their 
safety function.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not serve as an initiator or 
contributor to any accidents evaluated in the [Safety Analysis 
Report] SAR. It has no impact on any safety analysis assumptions. 
Exempting the HPCI valve MO-2-23-015 from quarterly stroke testing 
until 2RO11 does not impact its reliability or affect its ability to 
perform its intended safety function. The change does not adversely 
affect the assumptions or sequence of events used in any accident 
analysis.
    Therefore, 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: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
Pennsylvania 19101
    NRC Project Director: John F. Stolz

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

    Date of application for amendments: March 16, 1995
    Description of amendment request: This amendment would change the 
existing requirements for the Source Range Monitors (SRM) while the 
plant is in the refueling condition to requirements based on the 
Improved Technical Specifications in NUREG-1433, ``Standard Technical 
Specification General Electric Plants, BWR/4.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    The proposed changes to the SRM requirements will not increase 
the probability or consequences of an accident previously evaluated. 
The SRMs are not assumed to function during any UFSAR [Updated Final 
Safety Analysis Report] design basis accident or transient analysis. 
This TS change will not alter any safety limits which ensure the 
integrity of fuel barriers, and will not result in any increase to 
onsite or offsite dose. Additionally, continued availability of the 
SRMs in the refuel mode is ensured through additional testing 
requirements being added by this TS change. The changes to the SRM 
requirements will not alter the operation of equipment assumed to be 
available for the mitigation of accidents or transients.
    The proposed changes are based on NUREG-1433, ``Standard 
Technical Specifications General Electric Plants, BWR/4,'' and are 
consistent with the PECO Energy submittal of September 29, 1994, 
requesting an overall conversion, based on NUREG-1433. The overall 
conversion to the ITS [Improved Technical Specifications] included 
both technically justified deviations from the NUREG, and 
technically justified changes from the PBAPS current TS.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes to the SRM requirements will not create the 
possibility of a new or different type of accident from any 
previously evaluated. The SRMs are not assumed to function during 
any analyzed UFSAR design basis accident or transient analysis. 
Additionally, the changes will not involve any changes to plant 
systems, structures or components (SCCs) which could act as new 
accident initiators. Implementation of the proposed changes will 
effect the manner in which these SCCs are tested; however, TS 
requirements that govern routine testing and verification of plant 
components and variables are not assumed to be initiators of any 
analyzed event.
    3. The proposed change does not result in a significant 
reduction in the margin of safety.
    No margins of safety are reduced as a result of the proposed TS 
changes. No safety limits will be changed as a result of this TS 
change. The proposed change does not involve a reduction in the 
margin of safety because SRMs are not credited in any safety 
analysis. At least one SRM will remain operable during rod 
withdrawal during core alterations and rod withdrawal will not occur 
if no SRMs are operable. Excessive reactivity additions will be 
quickly identified and mitigated by the Intermediate Range Monitors 
and associated rod blocks. The Average Power Range Monitor Flux 
scram, and not any SRM function, is credited for mitigating a rod 
withdrawal or reactivity addition accident.
    Use of a spiral offload or reload pattern will provide assurance 
that the SRM will be in the optimum position for monitoring changes 
in neutron flux levels during core alternations.
    The changes proposed in this TS change do not introduce any 
hardware changes, and will not alter the intended operation of plant 
structures, systems or components utilized in the mitigation of 
accidents or transients. Additionally, these changes will not 
introduce any new failure modes of plant equipment not previously 
evaluated.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
Pennsylvania 19101
    NRC Project Director: John F. Stolz

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

    Date of application for amendments: March 22, 1995 [[Page 24914]] 
    Description of amendment request: The amendment would revise Note 
(1) for Technical Specifications Tables 3.7.2 through 3.7.4 by reducing 
the Local Leak Rate Test (LLRT) hold time duration from one hour to 20 
minutes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not serve as an initiator or 
contributor to any accidents previously evaluated. It does not 
decrease the effectiveness of equipment relied upon to mitigate 
previously evaluated accidents. The change does not involve any 
physical changes to any plant systems, structures, or components.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changed does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change does not serve as an initiator or 
contributor to any of the accidents previously evaluated. The 
proposed change does not introduce any new modes of plant operation.
    Implementation of the proposed changes will not affect the 
design function or configuration of any component or introduce any 
new operating scenarios or failure modes or accident initiation. It 
does not impair or prevent safety systems from performing their 
safety function.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed change does not serve as an initiator or 
contributor to any accidents evaluated in the SAR [Safety Analysis 
Report]. It has no impact on any safety analysis assumptions. 
Changing the LLRT duration hold time from one hour to 20 minutes 
does not impact equipment reliability. The change does not adversely 
affect the assumptions or sequence of events used in any accident 
analysis. Therefore, the propose 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: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
Pennsylvania 19101
    NRC Project Director: John F. Stolz

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

    Date of amendment request: November 21, 1994, as supplemented by 
letter dated April 6, 1995
    Description of amendment request: The proposed amendment would make 
changes affecting the Administrative Controls Section of the Technical 
Specifications (TSs). The areas proposed to be changed are: 1) NEEDS 
[Nuclear Effectiveness and efficiency Design Study] Organization Title 
Changes, 2) Minimum Shift Crew Composition, 3) Delete Independent 
Techincal Review Section from TS, 4) Delete NRB [Nuclear Review Board] 
Review Section from TS, and 5) Delete NRB Audit Section from TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed Technical Specifications changes do not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed TS changes to revise the organization position 
titles, PORC [Plant Operations Review Committee] composition 
description, and eliminate the Assistant Superintendent - Operations 
position do not involve any physical modifications to plant 
structures, systems, or components (SSC), or the manner in which 
these SSC are operated, maintained, modified, tested, or inspected. 
The proposed changes to position titles will not change the 
requirements for the qualifications and training of personnel in any 
management or supervisory position. Personnel will continue to meet 
the guidance specified in ANSI/ANS 3.1-1978 as required by Technical 
Specification 6.3.1. The probability of occurrence of an accident is 
based in part on: the training and qualifications of the personnel 
filling key plant management and supervisory positions; clear lines 
of authority, responsibility and communication; and, adequate 
management and corporate oversight of plant performance and 
activities. The proposed TS changes do not change any of these 
management and organizational elements.
    Allowing the Plant Manager to designate appropriately qualified, 
trained and experienced members of the LGS [Limerick Generating 
Station] staff as members of the PORC, as proposed, will not degrade 
the effectiveness of the PORC. The qualifications, training and 
experience level of the PORC will meet the requirements listed in 
ANSI/ANS 3.1-1978, and the required PORC quorum (including the use 
of alternates) will not be affected.
    Elimination of the position of Assistant Superintendent - 
Operations eliminates a level of supervision between the Plant 
Manager and the Shift Managers. The Shift Managers, who hold SRO 
licenses, will report directly to the Senior Manager - Operations. 
Other organizational changes within the Operations group (i.e., 
establishment of the positions of Manager - Operations Services and 
Manager - Operations Support) will ensure that the Senior Manager - 
Operations has sufficient time to properly supervise and monitor on-
shift performance. The Senior Manager -Operations and/or an 
Operations Manager will be required to hold a Senior Reactor 
Operator (SRO) license. Individuals filling these positions will 
satisfy the applicable training, qualifications, and experience 
requirements of ANSI/ANS 3.1-1978.
    The consequences of an accident could be affected by the 
qualifications and training of plant management and supervisory 
personnel. However, the proposed changes do not change the 
qualifications and training of personnel in any management or 
supervisory position. Personnel will continue to meet the criteria 
specified in ANSI/ANS 3.1-1978 as required by TS 6.3.1.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed TS changes to increase the minimum shift crew 
composition do not involve any physical changes to plant SSC.
    The probability of the occurrence of an accident is based in 
part on the operating crew and their ability to safely operate the 
plant. The increase in the minimum on-shift crew composition and the 
associated changes improves the capability of the on-shift crew to 
safely operate the plant and SSC, thereby reducing the probability 
of a situation that could result in an accident. The increase in the 
minimum on-shift crew composition will improve the manner in which 
the SSC are operated, maintained, tested, and inspected.
    The consequences of an accident could be affected by an 
operating error. However, the proposed TS changes increase the 
number of licensed operators required to be on-shift, and therefore, 
increase the capability of the on-shift crew to properly operate the 
facility and to implement the appropriate emergency procedures to 
reduce the consequences of an accident.
    The proposed changes will also delete redundant and/or relocate 
existing independent technical review and, Nuclear Review Board 
review and audit requirements from TS that are and/or will be 
contained in the LGS UFSAR [Updated Final Safety Analysis Report]. 
Removal of redundant/relocation of existing requirements does not 
affect any equipment important to safety, or involve any physical 
modifications to plant SSC, therefore, is not associated with an 
accident initiator or accident mitigator and [[Page 24915]] can not 
affect the probability of occurrence of an accident or increase the 
consequences of an accident. The licensee controlled UFSAR 
containing the requirements will be maintained using the provisions 
of 10 CFR 50.59, or 10 CFR 50.54(a), as appropriate, and are subject 
to the change control process in the Administrative Controls Section 
(6.0) of the Technical Specifications. Since future changes to 
related licensee-controlled documents will be evaluated per 10 CFR 
50.59 or 10 CFR 50.54(a), no increase (significant or insignificant) 
in the probability or consequences of an accident previously 
evaluated will be allowed.
    Therefore, these proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS changes to revise the organization position 
titles, PORC composition description, and eliminate the Assistant 
Superintendent - Operations position do not involve any physical 
modifications to plant structures, systems, or components (SSC), or 
the manner in which these SSC are operated, maintained, modified, 
tested, or inspected. The proposed changes to position titles will 
not change the requirements for the qualifications and training of 
personnel in any management or supervisory position. Personnel will 
continue to meet the guidance specified in ANSI/ANS 3.1-1978 as 
required by Technical Specification 6.3.1. Therefore, these proposed 
TS changes do not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed changes to the on-shift crew composition can not 
create the possibility of a new or different type of accident than 
previously evaluated in the SAR since implementation of the changes 
will not involve any physical changes to the plant SSC. The increase 
in the minimum on-shift crew composition increases the ability of 
the operating crew to ensure that the SSC are properly operated, 
maintained, tested and inspected. An increase in the required number 
of licensed operators on each shift improves the ability of the crew 
to adequately operate the facility, to respond to accident 
conditions, and to implement applicable plant procedures. Therefore, 
these proposed TS changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes will also delete redundant and/or relocate 
existing independent technical review and, Nuclear Review Board 
review and audit requirements from TS that are and/or will be 
contained in the UFSAR. The changes will not alter the plant 
configuration (no new or different type of equipment will be 
installed) or create changes in methods governing normal plant 
operation that will introduce new failure modes. These changes will 
not impose different requirements and proper control of information 
will be maintained. These changes will not alter assumptions made in 
the safety analysis and licensing basis. Therefore, these changes 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed TS changes to revise the organization position 
titles, PORC composition description, and eliminate the Assistant 
Superintendent - Operations position, do not reduce the margin of 
safety because positions with equivalent authority and 
responsibility are established and the new positions have equivalent 
requirements for education, experience and training. Allowing the 
Plant Manager to designate appropriately qualified, trained and 
experienced members of the LGS staff as members of the PORC will not 
degrade the effectiveness of the PORC because the qualifications, 
training and experience level of the PORC will meet the requirements 
listed in ANSI/ANS 3.1-1978 and the required PORC quorum (including 
the use of alternates) will not be affected. Elimination of the 
position of Assistant Superintendent - Operations eliminates a level 
of supervision between the Plant Manager and the Shift Managers. If 
the Senior Manager - Operations does not hold an SRO license, then 
an Operations Manager must hold an SRO license. This individual will 
1) be qualified to fill the Senior Manager - Operations position, 2) 
have the same management authority over the licensed operators as 
the Senior Manager - Operations, and 3) by being designated by 
Administrative procedures assures that there is always an individual 
holding a current SRO license in one of the Operations management 
positions. Other organizational changes (i.e., establishment of the 
positions of Manager - Operations Services and Manager - Operations 
Support), will ensure that the Senior Manager -Operations has 
sufficient time to properly supervise and monitor on-shift 
performance. Therefore, these changes do not involve a significant 
reduction in a margin of safety.
    The proposed changes to the on-shift crew composition increases 
the number of licensed SROs per shift to be one (1) above the 
minimum number required by the regulations. Additionally, the title 
changes are consistent with the organization and reporting 
relationships discussed in the regulation and the LGS Updated Final 
Safety Analysis Report (UFSAR). The Shift Manager holds a SRO 
license for both units and is assigned responsibility for overall 
plant operation at all times when there is fuel in any unit. The 
other SROs on the shift report to the Shift Manager and at least one 
(1) of the SRO licensed individuals is in the Main Control Room when 
either unit is in an operating mode other than cold shutdown or 
refuel. The increase in the minimum on-shift crew composition and 
the associated changes improves the capability of the on-shift crew 
to safely operate the plant and SSC. Therefore, these changes do not 
involve a significant reduction in a margin of safety.
    The proposed changes will also delete redundant and/or relocate 
existing independent technical review and, Nuclear Review Board 
review and audit requirements from TS that are and/or will be 
contained in the LGS UFSAR. The changes will not reduce the margin 
of safety since they have no impact on any safety analysis 
assumptions. In addition, any future changes to the UFSAR will be 
evaluated per the requirements of 10 CFR 50.59 or 10 CFR 50.54(a), 
as appropriate. Therefore, these changes will not involve a 
significant reduction in a margin of safety.
    The existing requirement for NRC review and approval of 
revisions, in accordance with 10 CFR 50.90, to these TS details and 
requirements proposed for relocation, does not have a specific 
margin of safety upon which to evaluate. However, since the proposed 
changes to delete redundant and/or relocate requirements are 
consistent with the BWR Standard Technical Specifications (NUREG-
1433) and the four criteria set forth in the NRC ``Final Policy 
Statement on Technical Specifications Improvements for Nuclear Power 
Reactors,'' and since the change controls for proposed relocated 
details and requirements provide an equivalent level of regulatory 
authority, revising the TS to reflect the approved level of detail 
and requirements ensures no 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: 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. 50-333, James 
A. FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: February 22, 1995
    Description of amendment request: The proposed changes to the James 
A. Fitzpatrick Technical Specifications establish operability and 
surveillance requirements for the Reactor Vessel Overfill Protection 
Instrumentation that initiates feedwater pump turbine trips, and a main 
turbine trip, on high reactor vessel water level.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated 
because: [[Page 24916]] 
    The proposed changes involve the addition of new operability and 
surveillance requirements to the Technical Specification regarding 
the current high reactor water level trip feature for the feedwater 
pump turbines and main turbine. The changes do not introduce any new 
modes of plant operation, make any physical changes, or alter any 
operational setpoints associated with the plants instrumentation and 
controls. Further, the Fitzpatrick UFSAR [Updated Final Safety 
Analysis Report], Section 14.5.9, for the Feedwater Controller 
Failure operational transient does not take credit for the automatic 
high reactor vessel water level trip of the feedwater pump turbines. 
The Fitzpatrick UFSAR analysis (Section 14.5.9), for the Feedwater 
Controller Failure operational transient assumes an automatic high 
reactor vessel water level trip of the main turbine. Incorporating 
these requirements into the Technical Specifications provides 
additional assurance that a trip feature described in the UFSAR 
remains functional. For these reasons the changes do not increase 
the probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from those previously evaluated because:
    The proposed changes do not introduce any new accident 
initiators or failure mechanisms since the changes do not introduce 
any new modes of plant operation, make any physical changes, or 
alter any operational setpoints. Accordingly, the changes do not 
create the possibility of a new or different kind of accident from 
those previously evaluated.
    3. Involve a significant reduction in the margin of safety 
because:
    The proposed changes establish operability and surveillance 
requirements for the design feature that trips the feedwater pump 
turbines and main turbine on high reactor vessel water level. The 
requirements will assure the continued operability of a trip 
function that is designed to initiate protective measures in the 
event of excessive feedwater flow. Tripping the feedwater pump 
turbines and main turbine on high reactor vessel water level, 
precludes potential adverse safety implications associated with a 
reactor overfill condition. Accordingly, the proposed changes will 
enhance the plant safety margin.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Ledyard B. Marsh

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

    Date of amendment request: March 2, 1995
    Description of amendment request: The proposed changes to the James 
A. Fitzpatrick Technical Specifications extend the surveillance test 
intervals for the snubber systems to support 24 month operating cycles.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes increase the interval between snubber 
functional tests. These changes are consistent with the guidance 
provided in Generic Letter 91-04. These changes do not involve any 
physical changes to the plant, nor do they alter the way snubbers 
function. The type of testing and the actions taken if a snubber 
fails a functional test remain the same. The review of the snubber 
installation and maintenance records will continue to ensure that 
the snubbers service life is not exceeded prior to the next 
scheduled review. The proposed changes to bases 4.0 and 4.6 clarify 
that the snubber functional testing interval is consistent with the 
length of the operating cycle. Therefore, the proposed changes do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes increase the interval between snubber 
functional tests. These changes are consistent with the guidance 
provided in Generic Letter 91-04. The proposed changes do not change 
the ability of the snubbers to provide dynamic load support during a 
design basis accident. Past operating experience indicates that the 
snubber program at the FitzPatrick plant adequately identifies 
snubber failures. No changes are proposed to the type of testing 
performed only to the surveillance interval length. The proposed 
changes do not modify the design or operation of plant equipment, 
therefore, no new or different failure modes are introduced. The 
Technical Specification for snubber testing is self-corrective. If 
any snubber fails a functional test, Technical Specifications 
require additional testing of a 10% sample of that type of snubber 
until no more failures are found. The functional test criteria 
remains unchanged and ensures a 95% confidence level that at least 
90% of the snubbers are operable. The proposed changes to bases 4.0 
and 4.6 clarify that the snubber functional testing interval is 
consistent with the length of the operating cycle. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes increase the interval between snubber 
functional tests. These changes are consistent with the guidance 
provided in Generic Letter 91-04. The proposed changes do not alter 
the configuration of the snubbers nor change the manner in which the 
snubbers function. Operation of the facility remains unchanged by 
the proposed changes. An evaluation of past equipment performance 
indicates that snubber operability is not time dependent. The 
proposed changes to bases 4.0 and 4.6 clarify that the snubber 
functional testing interval is consistent with the length of the 
operating cycle. Therefore, a longer surveillance test interval will 
not degrade snubber performance and will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Ledyard B. Marsh

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

    Date of amendment request: April 12, 1995
    Description of amendment request: The proposed changes to the James 
A. FitzPatrick Technical Specifications extend the surveillance test 
intervals for the nuclear steam supply system to support 24 month 
operator cycles.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes extend the surveillance test intervals for 
nuclear steam supply system components. These changes are consistent 
with the guidance provided in Generic Letter 91-04. The proposed 
changes do not involve any modification to the plant, nor do they 
alter equipment functions. On-line testing will provide a redundant 
and early means of demonstrating system [[Page 24917]] operability. 
Based on past results, SRV [safety/relief valve] mechanical 
performance has been good. No SRV setpoint changes are involved in 
this application. The proposed change to bases section 4.6 clarifies 
that the nuclear steam supply system surveillance testing interval 
is consistent with the length of the operating cycle. Therefore, the 
proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes extend the surveillance test intervals for 
nuclear steam supply system components. These changes are consistent 
with the guidance provided in Generic Letter 91-04. The proposed 
changes do not affect the way in which the nuclear steam supply 
system operates nor alter the type of surveillance testing 
performed. SRV drift analyses indicate that SRV drift with a 3% 
tolerance would be acceptable for (i.e., bounded by) a 24 to 30 
month interval. Leaking or partially open SRVs are detected by the 
acoustic monitoring system. Since the proposed changes do not modify 
the design or equipment of the plant, no new failure modes are 
introduced. The proposed change to bases section 4.6 clarifies that 
the nuclear steam supply system surveillance testing interval is 
consistent with the length of the operating cycle. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes extend the surveillance test intervals for 
nuclear steam supply system components. These changes are consistent 
with the guidance provided in Generic Letter 91-04. The proposed 
changes do not alter the configuration of the nuclear steam supply 
system nor change the manner in which the system functions. 
Operation of the facility remains unchanged by the proposed changes. 
An evaluation of past equipment performance indicates that SRV 
mechanical performance has been good. In addition, SRV drift has 
been analyzed to be within the allowable tolerance for the extended 
surveillance interval. The proposed change to bases section 4.6 
clarifies that the nuclear steam supply system surveillance testing 
interval is consistent with the length of the operating cycle. 
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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Ledyard B. Marsh

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

    Date of amendment request: March 3, 1995, as supplemented April 12, 
1995
    Description of amendment request: The licensee commenced operating 
on a 24-month fuel cycle, instead of the previous 18-month fuel cycle, 
with cycle 9. Fuel cycle 9 started in August 1992; however, the 
licensee shut down the facility in February 1993 for a performance 
improvement outage. Although a firm restart date has not yet been 
established, restart is expected in the spring of 1995. In order to 
accommodate operation on a 24-month cycle after the facility restarts, 
the licensee requested an amendment to the Technical Specifications 
(TSs) to incorporate the indicating instrument calibration frequency 
changes listed below:
    (1) The licensee proposed changing the calibration frequency for 
the containment water level monitor instrumentation (specified in TS 
Table 4.1-1) to accommodate operation on a 24-month cycle.
    (2) The licensee proposed changing the calibration frequency for 
the auxiliary feedwater (AFW) flow rate instrumentation (specified in 
TS Table 4.1-1) to accommodate operation on a 24-month cycle.
    (3) The licensee proposed changing the calibration frequency for 
the containment building ambient temperature sensors (specified in TS 
Table 4.1-1) to accommodate operation on a 24-month cycle.
    (4) The licensee proposed changing the calibration frequency for 
the seismic monitoring instrumentation (specified in TS Table 4.10-2) 
to accommodate operation on a 24-month cycle.
    In addition, the licensee proposed adding a new surveillance 
requirement to TS Table 4.1-1 for testing the core exit thermocouples.
    These proposed changes follow the guidance provided in Generic 
Letter 91-04, ``Changes in Technical Specification Surveillance 
Intervals to Accommodate a 24-Month Fuel Cycle,'' as applicable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Consistent with the criteria of 10 CFR 50.92, the enclosed 
application is judged to involve no significant hazards based on the 
following information:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of any accident 
previously evaluated?
    Response:
    The proposed changes do not involve a significant increase in 
the probability or consequences of any accident previously 
evaluated. The proposed changes extend the calibration frequency (to 
24 months) for the:
     containment temperature channels,
     containment water level monitoring system channels,
     seismic instrumentation channels, and
     auxiliary feedwater flow rate channels.
    These changes are being made to accommodate a 24 month operating 
cycle. The proposed changes in the calibration frequencies do not 
involve any plant hardware changes, nor do they change the way the 
systems function.
    Extension of the calibration and surveillance test intervals in 
question were evaluated and the results documented in [New York 
Power Authority (NYPA) Report No. IP3-RPT-MULT-00424, ``Indicating 
Instruments Surveillance Test Extensions,'' May 1993]. An Instrument 
Drift Analysis for the indicating instruments [NYPA Report No. IP3-
RPT-MULT-00407, ``Instrument Drift Analysis for Indicating Loops,'' 
April 1993] was performed to evaluate past and future instrument 
drift. The results of these evaluations and analyses indicate that 
the calibrations in question can safely be extended to accommodate 
the 24 month operating cycle.
    For containment temperature, auxiliary feedwater flow and 
seismic instrumentation, past instrument drift has generally been 
within acceptable limits. Some drift exceeding the calibration 
tolerance did occur for the triaxial time-history accelographs, but 
on-line testing should ensure that instrument drift over the longer 
cycle does not degrade system performance. For containment water 
level systems (except containment building level), new electronic 
transmitters were recently installed. Due to the lack of data, an 
instrument drift analysis was not performed. However, the new 
containment water level transmitters improved the overall channel 
accuracy.
    Future instrument drift was predicted and used to update 
existing loop accuracy calculations, with the following results. (1) 
For the containment temperature channels, the loop accuracy 
calculations were revised to incorporate the larger channel 
uncertainties. Postulated drift over 30 months should have a 
negligible effect on the EOPs [Emergency Operating Procedures] and 
plant shutdown. (2) For the containment system sump water levels, 
future drift is not a concern because the containment building water 
level is used post accident. The larger uncertainties can safely be 
accommodated by changing the EOP setpoint for transfer to cold leg 
recirculation. (3) For the seismic instrumentation, past drift was 
negligible, and future drift is not expected to be cycle length 
dependent. (4) For the auxiliary [[Page 24918]] feedwater flow rate 
channels, the larger uncertainties can be safely accommodated by 
changing the EOP setting for the minimum AFW flow required for heat 
removal.
    For the containment temperature and seismic instrumentation, on-
line testing provides added assurance that the instrumentation is 
functioning as required.
    [For the core exit thermocouples, adding a requirement to 
conduct testing every 18 months will serve to ensure system 
operability. This new testing requirement does not change the way 
the plant operates or involve hardware modifications.]
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any previously 
evaluated?
    Response:
    The proposed changes do not create the possibility of a new or 
different kind of accident from any previously evaluated. The 
proposed changes extend the calibration frequency (to 24 months) for 
the:
     containment temperature channels,
     containment water level monitoring system channels,
     seismic instrumentation channels, and
     auxiliary feedwater flow rate channels.
    These changes are being made to accommodate a 24 month operating 
cycle. The proposed changes in the calibration frequencies do not 
involve any plant hardware changes, nor do they change the way the 
systems function.
    Extension of the calibration and surveillance test intervals in 
question were evaluated and the results documented in [same as 
Question (1)]. An Instrument Drift Analysis for the indicating 
instruments [same as Question (1)] was performed to evaluate past 
and future instrument drift. The results of these evaluations and 
analyses indicate that the calibrations in question can safely be 
extended to accommodate the 24 month operating cycle. For the 
containment temperature and seismic instrumentation, on-line testing 
provides added assurance that the instrumentation is functioning as 
required.
    [For the core exit thermocouples, adding a requirement to 
conduct testing every 18 months will serve to ensure system 
operability. This new testing requirement does not change the way 
the plant operates or involve hardware modifications.]
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response:
    The proposed changes do not involve a significant reduction in a 
margin of safety. The proposed changes extend the calibration 
frequency (to 24 months) for the:
     containment temperature channels,
     containment water level monitoring system channels,
     seismic instrumentation channels, and
     auxiliary feedwater flow rate channels.
    These changes are being made to accommodate a 24 month operating 
cycle. The proposed changes in the calibration frequencies do not 
involve any plant hardware changes, nor do they change the way the 
systems function.
    For containment temperature, auxiliary feedwater flow and 
seismic instrumentation, past instrument drift has generally been 
within acceptable limits. Some drift exceeding the calibration 
tolerance did occur for the triaxial time-history accelographs, but 
on-line testing should ensure that instrument drift over the longer 
cycle does not degrade system performance. For containment water 
level systems (except containment building level), new electronic 
transmitters were recently installed. Due to the lack of data, an 
instrument drift analysis was not performed. However, the new 
containment water level transmitters improved the overall channel 
accuracy.
    [For the core exit thermocouples, adding a requirement to 
conduct testing every 18 months will serve to ensure system 
operability. This new testing requirement does not change the way 
the plant operates or involve hardware modifications.]
    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 Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
County, New Jersey

    Date of amendment request: March 30, 1995
    Description of amendment request: The proposed change to the 
Technical Specifications eliminates the defined term CONTROLLED 
LEAKAGE, removes Controlled Leakage flow from the Reactor Coolant 
System Operational Leakage Limiting Condition for Operation (LCO), and 
establishes a new Seal Injection Flow LCO.
    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. Do not involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    Changing the Technical Specification to limit seal injection 
flow instead of seal leakoff flow does not affect the probability of 
any accident previously evaluated. Maintaining adequate Emergency 
Core Cooling System (ECCS) flow during Loss of Coolant Accident 
(LOCA) ensures that the consequences of these accidents are 
unaffected. The existing Technical Specification allows seal 
injection throttle valve positioning that could result in seal 
injection flow path resistance values below those used in the Salem 
ECCS hydraulic flow analyses. Reduced line resistances could result 
in inadequate ECCS flow to the reactor core. Revising the Technical 
Specification to limit RCP seal injection flow ensures that the 
accident analysis assumptions are maintained, and the previously 
evaluated accident consequences remain unchanged.
    Therefore, it may be concluded that the proposed changes do not 
increase the probability or consequences of an accident previously 
evaluated.
    2. Do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed changes do not involve any hardware modifications 
or result in any functional changes to system operation. RCP seal 
injection flow is used as a limiting parameter in-place of RCP seal 
leakoff flow.
    Since design requirements continue to be met and the RCS 
pressure boundary is not challenged, no new failure mode is created. 
Thus, an accident different from any already evaluated is not 
created by this change.
    Therefore, it may be concluded that the proposed changes do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Do not involve a significant reduction in a margin of safety.
    The proposed changes do not alter the manner in which Safety 
Limits or Limiting Safety System Setpoints are determined. 
Controlled Leakage (RCP seal leakoff)is removed from the Reactor 
Coolant System Leakage Limiting Condition for Operation (LCO), and a 
new seal injection LCO is established. The new LCO continues to 
limit seal injection flow during accident conditions. The limiting 
parameter is changed from RCP seal leakoff flow to RCP seal 
injection flow. These changes ensure that the accident analysis 
assumptions and existing margins of safety are maintained. The seal 
injection flow specification limit is not applicable in Mode 4 and 
lower, because high seal injection flow is less critical due to 
lower Reactor Coolant System (RCS) pressure and decay heat removal 
requirements in these modes. Reactor coolant pump seal injection 
flow must be limited in Modes 1, 2, and 3 to ensure adequate 
Emergency Core Cooling System Flow.
    Therefore, it may be concluded that the proposed changes do not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, New Jersey 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
[[Page 24919]] Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

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

    Date of amendment request: April 6, 1995 (TS 95-05)
    Description of amendment request: The proposed change would (1) 
replace the reference to Table 3.6-2 from Definition 1.7.a.2 for 
Containment Integrity with a phrase that will allow the valves to be 
opened under administrative control; (2) replace the reference to Table 
3.6-2 from Surveillance Requirement 4.6.1.1 with a phrase that will 
allow the valves to be opened under administrative control; (3) delete 
the reference to Table 3.6-1 from Technical Specification 3.6.1.2; (4) 
delete Table 3.6-1, ``Bypass Leakage Paths to the Auxiliary Building -- 
Secondary Containment Bypass Leakage Paths;'' (5) revise Specification 
3.6.3 to delete the reference to Table 3.6-2, add a footnote that 
discusses the opening of penetrations intermittently, add the phrase to 
take exception to the containment vacuum isolation valves, and add an 
action statement to indicate that Specification 3.0.4 does not apply to 
the specification; (6) delete Surveillance Requirement 4.6.3.1; (7) 
delete references to Table 3.6-2 in Specifications 4.6.3.2 and 4.6.3.3 
and additional wording added to indicate that the specifications apply 
to automatic containment isolation valves; (8) delete Table 3.6-2, 
``Containment Isolation Valves'' and add a note to the page indicated 
that the information has been intentionally deleted; (9) revise 
Specification 3.8.3.1 to specify that the Limiting Condition for 
Operation applies to primary and backup containment penetration 
conductor overcurrent protective devices associated with each 
containment electrical penetration shall be operable, add a phrase to 
indicate that the scope of these protective devices excludes those 
circuits for which credible fault currents would not exceed the 
electrical penetration design rating, and delete the phrase that 
references appropriate plant instructions in the action statement; (10) 
delete the phrase that references appropriate plant procedures from 
Specification 4.8.3.1; (11) delete the phrase from SR 4.8.3.1.a.3 that 
indicates that a complete listing of all fuses to be verified in 
accordance with the requirement will be maintained in appropriate plant 
instructions; (12) replace the phrase ``appropriate plant instructions 
based on'' with ``procedures prepared in conjunction with'' in SR 
4.8.3.1.b; (13) replace the reference to Table 3.8-2 in Specification 
3.8.3.2 with a phrase that indicates that the Requirement is applicable 
to valves used in safety systems; (14) delete Table 3.8-2, ``Motor 
Operated Valves Thermal Overload Protection,'' and replace it with a 
note that indicates that the pages are intentionally blank; and (15) 
incorporate appropriate changes to the Bases to reflect these changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The removal of the component listings from the SQN TSs will not 
create an increase in the probability or consequences of any 
accident previously evaluated. Although no longer in the TSs, the 
components listed in Tables 3.6-1, 3.6-2, and 3.8-2 will be 
contained in administratively controlled documents. This equipment 
must be tested at the required intervals and each unit's action 
statements must still be adhered to. These procedures are revised 
and approved in accordance with requirements of TS Section 6.5.1A. 
This review process also requires an evaluation based on 10 CFR 
50.59 requirements. As indicated in GL 91-08, this is adequate 
control for changes to these components lists.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The removal of the component lists from the TSs does not modify 
safety-related equipment or systems, nor does it change any safety-
related setpoints used to prevent or mitigate previously analyzed 
accidents. The component lists are presently located in separate 
documents that are subject to the requirements of 10 CFR 50.59. 
Also, the limiting condition of operation requirements remain in 
effect and appropriate actions will be taken if any limits are 
exceeded. Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The margin of safety is not affected by the removal of the 
previously discussed component lists from the TS. Appropriate 
measures presently exist to control the setpoint of the components 
listed. Any changes to these setpoints are controlled by the SQN 
design change process that is subject to the requirements of 10 CFR 
50.59 in which the reduction of the present margin of safety is 
addressed. The proposed amendment continues to require operation 
within the set values for these components, and appropriate actions 
to be taken when or if the limits are exceeded. Based on these 
controls, this amendment will not involve a reduction in a margin of 
safety.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: March 24, 1995
    Description of amendment request: The licensee has requested a one-
time extension of the performance intervals for certain Technical 
Specification Surveillance Requirements (SR). Affected SRs include 
penetration leak rate testing, valve operability testing, instrument 
calibration, response time testing, and logic system functional tests. 
The proposed changes are requested to support refueling outage 5 
scheduled to begin no later than February 15, 1996.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed TS change requests a one-time extension of the 
surveillance intervals related to: a) RPS Instrumentation 
calibration, LSFTs, and response time testing; b) Isolation 
Actuation System Instrumentation calibration, LSFTs, and response 
time testing; c) ECCS Actuation Instrumentation calibration, LSFTs, 
and response time testing; d) Control Rod Block Instrumentation 
calibration and LSFTs; e) Remote Shutdown Instrumentation and 
Controls calibration and operability testing; f) 
[[Page 24920]] Accident Monitoring Instrumentation calibration; g) 
Plant Systems Instrumentation calibration and LSFTs; h) Primary 
Containment automatic valve actuation; i) Reactor Coolant System 
Pressure Isolation Valve (PIV) testing; j) system automatic 
initiation testing; and, k) Emergency Diesel Generator inspection 
and testing.
    Also proposed is the re-establishment of the baseline for the 
``N times 18 months'' cumulative surveillance interval for response 
time testing.
    The discussion in the License Amendment Request demonstrates the 
following:
    i) Rosemount transmitter calibration period extension is 
acceptable based on Rosemount D8900126, Revision A which supported 
extension of the calibration interval from 18 months to 30 months 
based on the reduction in the drift allowance;
    ii) Extrapolation of plant specific calibration data is 
acceptable in supporting the extension of other calibration 
surveillance intervals to RFO-5;
    iii) LSFT interval extension 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;
    iv) Response time testing interval extension for Isolation 
Actuation and ECCS Actuation instrumentation channels is acceptable 
based on the BWR Owners Group (BWROG) Licensing Topical Report NEDO-
32291 (January 1994) which provided the necessary justification for 
elimination of response time testing and, therefore, provides a 
suitable argument for extending the interval for a short period of 
time. The NRC approved the use of NEDO-32291 as a basis for License 
Amendment Requests, with additional conditions specified, in a 
letter to the BWROG in December 1994.
    v) Response time testing interval extension for RPS 
Instrumentation channels is acceptable because: i) there are 
redundant sensors that can initiate the scram function; ii) one-out-
of-two redundancy exists in every individual instrument channel 
within each trip function; iii) several redundant and diverse 
instrument channels are provided which can detect and generate a 
scram signal; iv) the failure probability is a small fraction of the 
total control rod insertion (scram) failure probability; v) failure 
of instrumentation in the sluggish mode is a small fraction of its 
overall failure modes; and iv) NRC Safety Evalution Report dated 
August 2, 1993 (Peach Bottom Atomic Power Station, Units 2 and 3 
docket) has previously provided approval for extension of the RPS 
response time testing surveillance interval from 18 to 24 months.
    vi) Response time testing interval extension for the Main Steam 
Line isolation is acceptable because i) redundancy and diversity 
exist in individual instrument channels within a trip function; ii) 
instrumentation response time is a small fraction of the overall 
response time of the actuating device; iii) instrumentation failure 
probability is a very small portion of the total MSIV failure 
probability; and, iv) failure of instrumentation in the sluggish 
responding mode is a small fraction of its overall failure modes.
    vii) Containment Isolation Valve leakage determination and 
actuation interval extension is acceptable based on: i) redundancy 
provided in the design of the penetrations; ii) the periodic testing 
of the valves during power operation; and, iii) the short period of 
time the interval is being extended.
    viii) Reactor Coolant System PIVs have exhibited low as-found 
leak rates as measured during the last refueling outage; there is 
substantial margin available for the PIVs from the as-left leakage 
to the allowed TS leakage; the requested extension of the 
surveillance interval is small; and the conclusion of NUREG-1463, 
``Regulatory Analysis for the Resolution of Generic Safety Issue 
105: Interfacing System Loss-of-Coolant Accident in Light Water 
Reactors'' (July 1993), and the confirmation of the PNPP Individual 
Plant Examination that the ISLOCA (for which PIVs are provided to 
prevent) is not a risk concern to BWRs or PNPP.
    ix) System initiation and actuation testing interval is 
acceptable based on the periodic testing of components during power 
operation and the short period of time the interval is being 
extended.
    x) Emergency Diesel Generator testing interval extension is 
acceptable based on: i) the past testing results which support 
extension for the short period of time; ii) the testing that is done 
during power operation; and, iii) the short period of time the 
interval is being extended.
    xi) The re-establishment of the baseline for the ``N times 18 
months'' cumulative surveillance interval for response time testing 
is acceptable in that the extension of the cumulative interval would 
not be for more than the individual extension requested and 
justified herein.
    Therefore, from the above it is shown that the proposed change 
will not significantly increase the probability 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 a one-time extension of the 
surveillance intervals for instrument calibration, instrument 
channel LSFT and response time testing, containment isolation valve 
leakage determination and actuation, PIV leak rate determination, 
system actuation testing, and diesel generator inspection and 
testing. The proposed changes do not necessitate a physical 
alteration to the plant (no new or different type of equipment will 
be installed). The requested extension durations are small as 
compared to the overall interval allowed by TS; drift data supports 
extension of the calibration intervals; NRC and industry evaluations 
support extension of LSFT; industry evaluations and redundancy in 
system design support extension of response time testing; past 
testing and periodic testing provides confidence of no effect on 
equipment availability by extending the confidence of no effect on 
equipment availability by extending the surveillance interval. 
Therefore, the change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    In addition, the requested re-establishment of the baseline at 
RFO-5 for the ``N time 18 months'' cumulative surveillance interval 
for response time testing is acceptable in that the cumulative 
surveillance interval will not be extended by more than that which 
is proposed for individual response time tests during RFO-5. The 
individual response time test surveillance interval extensions have 
been justified herein. The justification for individual response 
time test surveillance interval extensions applies to the cumulative 
surveillance interval extension which is requested and will be 
granted by allowing the re-establishment of the baseline of the ``N 
times 18 months'' surveillance interval to the response time testing 
dates for those response time tests to be performed during RFO-5. 
The proposed changes do not necessitate a physical alteration to the 
plant (no new or different type of equipment will be installed). 
Therefore, the change does not create the possibility of a new or 
different kind of accident.
    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 instrument calibration, instrument 
channel LSFT, and response time testing, containment isolation valve 
leakage determination and actuation, PIV leak rate determination, 
system actuation testing, and diesel generator inspection and 
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, drift data 
supports extension of the calibration intervals, NRC and industry 
evaluations support extension of LSFT, industry evaluations and 
redundancy in system design support extension of response time 
testing, past testing and periodic testing provides 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.
    In addition, the requested re-establishment of the baseline at 
RFO-5 for the ``N times 18 months'' cumulative surveillance interval 
for response time testing is acceptable in that the cumulative 
surveillance interval will not be extended by more than that which 
is proposed for individual response time tests during RFO-5. The 
individual response time test surveillance interval extensions have 
been justified herein. The justification for individual response 
time test surveillance interval extensions applies to the cumulative 
surveillance interval extension which is requested and will be 
granted by allowing the re-establishment of the baseline of the ``N 
times 18 months'' surveillance interval to the response time testing 
dates for those response [[Page 24921]] time tests to be performed 
during RFO-5. The proposed changes do not necessitate a physical 
alteration to the plant (no new or different type of equipment will 
be installed). Therefore, 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

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: April 3, 1995
    Description of amendment request: The proposed amendment would add 
new programmatic requirements governing radiological effluent into the 
Administrative Controls section of the Technical Specifications in 
accordance with Generic Letter 89-01, ``Implementation of Programmatic 
Controls for Radiological Effluent Technical Specifications in the 
Administrative Controls Section of Technical Specifications and the 
Relocation of Procedural Details of RETS to the Offsite Dose 
Calculation Manual or to the Process Control Program.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes are administrative in nature and alter only 
the format and location of programmatic controls and procedural 
details relative to radioactive effluent, radiological environmental 
monitoring, solid radioactive wastes, and associated reporting 
requirements. Compliance with applicable regulatory requirements 
will continue to be maintained. In addition, the proposed changes do 
not alter the conditions or assumptions in any of the Updated Safety 
Analysis Report (USAR) accident analyses. Since the USAR accident 
analyses remain bounding, the radiological consequences previously 
evaluated are not adversely affected by the proposed changes. 
Therefore, it can be concluded that the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes do not involve any changes to the 
configuration or method of operation of any plant equipment. 
Accordingly, no new failure modes have been defined for any plant 
system or component important to safety nor has any new limiting 
single failure been identified as a result of the proposed changes. 
Also, there will be no change in types or increase in the amounts of 
any radioactive effluent released offsite. Therefore, it can be 
concluded that the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed changes do not involve any actual change in the 
methodology used in the control of radioactive effluents, solid 
radioactive wastes, or radiological environmental monitoring. These 
changes are considered administrative in nature, provide for the 
relocation of procedural details outside the Technical 
Specifications, and add appropriate administrative controls in the 
Technical Specifications to provide continued assurance of 
compliance with applicable regulatory requirements. These proposed 
changes also comply with the guidance contained in Generic Letter 
89-01. Therefore, it can be concluded that the proposed changes do 
not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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

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

    Date of amendment request: February 24, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Surveillance Requirement 4.6.1.7.4 
and its associated Bases to delete the quarterly verification of the 
measured leakage rate for containment mini-purge supply and exhaust 
isolation valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed revision does not involve a significant hazards 
consideration because operation of Callaway Plant with this change 
would not:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed revision to the T/S will not adversely impact plant 
safety since the requirement to perform the quarterly surveillance 
will still be implemented to verify valve leakage and seal 
degradation. The mini-purge valves will still perform their intended 
safety function to close within 5 seconds after receipt of an 
isolation signal.
    2) Create the possibility of a new or different kind of accident 
from any previously evaluated.
    There are no design changes being made that would create a new 
type of accident or malfunction and the method and manner of plant 
operation remain unchanged. Deletion of the individual leakage rate 
for these valves does not affect the severity of any accident 
previously evaluated. The consequences of a valve failure or 
malfunction are not increased by the removal of the acceptance 
criteria, leakage rate will still be measured on a quarterly basis 
as is currently done to determine if the seals are degrading.
    3) Involve a significant reduction in a margin of safety.
    There are no changes being made to the safety limits or safety 
system settings that would adversely impact plant safety. The valves 
will still be surveilled on a quarterly basis to verify leakage and 
seal degradation to assure gross failure will not occur and that 
containment integrity is maintained.
    Based on the above discussions, it has been determined that the 
requested Technical Specification change does not involve a 
significant increase in the probability or consequences of an 
accident or create the possibility of a new or different kind of 
accident or condition over previous evaluations; or involve a 
significant reduction in a margin of safety. Therefore, the 
requested license amendment does not involve a significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251. [[Page 24922]] 
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: April 17, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Table 2.2-1 and associated Bases to 
reduce repeated alarms and partial reactor trips related to the C-4 
control system interlock and the Overpower Delta-T (OP[delta]T) reactor 
trip setpoint.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed revision does not involve a significant hazards 
consideration because operation of Callaway Plant with this change 
would not:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Overall protection system performance will remain within the 
bounds of the accident analyses documented in Final Safety Analyses 
Report (FSAR) Chapter 15, WCAP-10961-P for Category 1 plants such as 
Callaway, and WCAP-11883 since no hardware changes are proposed.
    The OP[delta]T reactor trip function provides protection against 
excessive power (fuel rod integrity protection within the fuel 
temperature design basis). No credit is taken for the OP[delta]T 
trip in the Chapter 15 licensing basis accident analyses. The 
[delta]T trip function is credited in non-licensing basis analyses 
of various steamline breaks.
    The OP[delta]T trip will continue to function in a manner 
consistent with the plant design basis. There will be no change to 
the OP[delta]T safety analysis limit listed in FSAR Table 15.0-4. 
Therefore, there will be no degradation in the performance of or an 
increase in the number of challenges to equipment assumed to 
function during an accident situation.
    The reactor trip system response time, as defined in the 
Technical Specifications, will be unaffected.
    These Technical Specification revisions do not involve any 
hardware changes nor do they affect the probability of any event 
initiators. There will be no change to normal plant operating 
parameters or accident mitigation capabilities. Therefore, these 
changes will not increase the probability or consequences of an 
accident or malfunction.
    2) Create the possibility of a new or different kind of accident 
from any previously evaluated.
    As discussed above, there are no hardware changes associated 
with these Technical Specification revisions nor are there any 
changes in the method by which any safety-related plant system 
performs its safety function. Revisions to the OP[delta]T values for 
K4 and K6 will require scaling changes for summing 
amplifier cards (NSA cards) in the 7300 Process Protection System. 
These scaling changes are straightforward and similar in nature to 
those performed to implement OL Amendments 72 and 84 associated with 
the implementation of relaxed axial offset control (RAOC) and a 
revised OT[delta]T f1([delta]I) penalty function. These scaling 
changes will not affect the normal manner of plant operation. There 
will be a reduction in the incidence of C-4 alarms and partial 
reactor trips. There will be less of a need to reduce power during 
on-line surveillance testing.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of these changes. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of these changes. 
Therefore, the possibility of a new or different kind of accident is 
not created.
    3) Involve a significant reduction in a margin of safety.
    There will be no change to the Overpower [delta]T safety 
analysis limit listed in FSAR Table 15.0-4. Available setpoint 
calculation margin will be used to increase the K4 value, 
reflected as a new bias on a summing amplifier card in each of the 
four protection loops. This will also require corresponding 
decreases in the OP[delta]T Total Allowance and Allowable Value in 
Technical Specification Table 2.2-1. Available margin in the 
OP[delta]T trip protection function will be used to decrease the 
K6 value, reflected as a new gain on a summing amplifier card 
in each of the four protection loops.
    As discussed above, the response time of the OP[delta]T reactor 
trip function will remain unchanged.
    It has been confirmed that the Z and S terms currently listed in 
Table 2.2-1 for the OP[delta]T trip function will remain 
conservative. The change in K4 will result in a decrease in the 
Total Allowance and Allowable Value for OP[delta]T; however, this 
does not affect any margin of safety since the safety analysis 
limit, which preserves the overpower safety margin, is unchanged.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. There will be no impact on the overpower 
limit, DNBR limits, FQ, F[delta]H, LOCA PCT, peak local power 
density, or any other margin of safety.
    Based upon the preceding information, it has been determined 
that the proposed changes to the Technical Specifications do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated, create the possibility of a new or 
different kind of accident from any accident previously evaluated, 
or involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: October 28, 1994
    Description of amendment request: The proposed amendment would 
remove the Neutron Monitoring System (NMS) and Control Rod Position 
instrumentation from the Vermont Yankee Technical Specifications for 
post-accident monitoring. Administrative changes are also proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change to remove the NMS and Control Rod 
Position instrumentation from the Technical Specifications for post-
accident monitoring is consistent with NRC requirements concerning 
this instrumentation.
    Wide Range Neutron Flux (NMS instrumentation) is presently 
included in the [boiling water reactor] BWR Standard Technical 
Specifications, but the NRC has recently determined [letter, USNRC 
to VYNPC, dated April 29, 1993] that this instrumentation need not 
meet R.G. 1.97 Category 1 criteria and that licensees may request 
the removal of this instrumentation from their post-accident 
monitoring Technical Specifications. Control Rod Position 
instrumentation is considered R.G. 1.97 Category 3 which is required 
to meet the least stringent design and qualification criteria as 
specified in this regulatory guide.
    Testing, calibration and maintenance of this instrumentation 
will continue to assure operability of instrumentation. The portions 
of the NMS and the Control Rod Position instrumentation systems to 
be removed from the post-accident monitoring Technical 
Specifications do not perform any automatic control or trip 
function. In addition, this instrumentation does not provide 
information that is required to permit the control room operator to 
take manual actions that are required for safety systems to 
accomplish their safety functions for design basis accident 
events. [[Page 24923]] 
    At a BWR, when all control rods are inserted, these control rods 
cannot be withdrawn without deliberate operator action. The proposed 
change does not result in any system hardware modification or new 
plant configuration. The requested change to post-accident 
monitoring instrumentation does not impact any [Final Safety 
Analysis Report] FSAR safety analysis involving the NMS or Control 
Rod Position System. These monitoring functions are not contributors 
to the initiation of accidents.
    The administrative changes to correct a typographical error and 
instrument ranges will have no effect on plant hardware, plant 
design, safety limit setting or plant system operation and 
therefore, do not modify or add any initiating parameters that would 
significantly increase the probability or consequences of any 
previously analyzed accident.
    Therefore, it is concluded that there is not a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The function of the instrumentation to be removed from the 
Technical Specifications is for monitoring only. These indications 
are not necessary for operators to accomplish any safety functions.
    The proposed change does not involve any change in hardware, 
Technical Specification setpoints, plant operation, redundancy, 
protective function or design basis of the plant. There is no impact 
on any existing safety analysis or safety design limits. NMS and 
Control Rod Position monitoring functions do not initiate nuclear 
system parameter variations which are considered potential 
initiating causes of threats to the fuel and the nuclear system 
process barrier.
    As discussed above, the proposed administrative change only 
corrects a typographical error concerning equipment identification 
numbers and listed instrument ranges. This change does not affect 
any equipment and they do not involve any potential initiating 
events that would create any new or different kind of accident.
    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 to remove the NMS and Control Rod 
Position instrumentation from the Technical Specifications for post-
accident monitoring does not affect any existing safety margins. The 
original NMS design basis for BWRs never required a post-accident 
neutron monitoring function since there are no design basis 
accidents that rely on operator action to control reactor power. 
This is also true for Control Rod Position monitoring.
    Existing Technical Specifications requirements for automatic 
trip functions are unaffected. Failure of the indication of reactor 
power from the NMS or the Control Rod Position System does not 
preclude the ability of the reactor operator to determine reactor 
power levels. Alternate indications are available to ascertain 
reactor power. These include reactor coolant boron concentrations, 
flux levels from the Traversing Incore Probe (TIP) System and the 
status of plant parameters which are linked to reactor power. In 
addition, alternate means of determining reactor power have been 
incorporated into the Emergency Operating Procedures (EOPs).
    Operation, testing and maintenance of this instrumentation will 
remain the same. System functions are the same. Post-accident 
functional design criteria as described in [BWR Owners Group Topical 
Report NEDO-31558-A, dated March 29, 1993], and approved by the NRC 
are satisfied by present equipment installed at VY. NMS 
instrumentation is still included in the Technical Specifications 
for the [Reactor Protection System] RPS. Control Rod Position 
instrumentation does not perform any safety function.
    As discussed above, the proposed administrative changes do not 
affect any equipment involved in potential initiating events or 
safety limits.
    Based upon the above, it is concluded that the proposed change 
does not involve a significant reduction in a margin of safety.
    Based upon the above, we conclude that the proposed change does 
not constitute a significant hazards consideration as defined in 
10CFR50.92(c).
    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: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301
    Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray, 
One International Place, Boston, MA 02110-2624
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: March 30, 1995
    Description of amendment request: The licensee is requesting 
temporary changes to Technical Specifications (TS) 3.7.3.1, ``Component 
Cooling Water Subsystem - Operating,'' and 3.7.4.1, ``Service Water 
System - Operating,'' for NA-1&2. The proposed TS changes will allow 
one of the two service water loops to be isolated from the component 
cooling water heat exchangers during power operation in order to 
refurbish the isolated service water headers.
    NA-1&2 is currently pursuing refurbishment of the 18-inch, 20-inch 
and 24-inch diameter service water supply and return lines to/from the 
NA-1 and NA-2 component cooling heat exchangers (CCHXs). Refurbishment 
of this piping presents a challenge in that it is not possible to 
isolate and plug or blank the section to be worked in a 7-day time 
period. The purpose of the proposed change is to request temporary 
changes to the existing servicewater (SW) and component cooling water 
(CC) TS to permit orderly and efficient conduct of the pipe 
refurbishment project during two-unit power operation. Specifically, 
the licensee is proposing to temporarily change TS 3.7.4.1 ``Service 
Water System - Operating'' to allow operation of the SW system with one 
independent source of SW to/from the NA-1 and NA-2 CCHXs for two 
periods of up to 49 days each. This proposed change also allows the 
automatic closure feature of the SW valves to/from the CCHXs to be 
defeated during the 49-day periods. In addition, the licensee proposes 
to temporarily change TS 3.7.3.1 ``Component Cooling Water Subsystem - 
Operating'' with a footnote which considers the CC subsystems OPERABLE 
with only one independent source of SW provided to/from the CCHXs 
during these 49-day periods. Further, the proposed change would allow 
that during operation with only one SW header available to/from the 
CCHXs, the provisions of Specification 3.0.4 would not be applicable 
provided two SW loops are capable of providing cooling for the other 
operable plant components.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Specifically, operation of North Anna Power Station in accordance 
with the proposed Technical Specifications changes will not:

    Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The piping refurbishment project and the proposed temporary 
changes to the SW and CC Technical Specifications have been 
evaluated to assess their impact on the normal operation of the SW 
and CC systems and to ensure that the design basis safety functions 
of each system are preserved. The SW system is required to function 
during all normal and emergency operating conditions. During normal 
plant operation, the SW system provides cooling water to the CCHXs, 
charging pump coolers, instrument air compressor coolers, and 
control room chiller condensors of both units. During the two 49-day 
periods, one header will [operate] with its 24-inch piping to/from 
the CCHXs temporarily blanked. To avoid operation of the SW pump at 
abnormal conditions (low flow) on this ``partially deadlocked'' 
header, a temporary cross-connect will be installed to by-pass the 
CCHXs. [[Page 24924]] 
    SW system operation with the cross-connect installed was 
evaluated for design basis accident (DBA) conditions. The DBA 
condition for the SW system is a loss-of-coolant accident on one 
unit with simultaneous loss-of-offsite-power to both units. A SW 
system hydraulic analysis has been performed to verify that adequate 
flow is provided to the containment recirculation spray heat 
exchangers (RSHXs) with the temporary cross-connect installed and 
throttled open assuming the occurrence of the most limiting single 
failure. Therefore, there is no increase in probability or 
consequences of the DBA condition.
    Utilizing only one SW header to supply flow to the CCHXs has the 
potential to affect the reliability of the CC system and all of the 
equipment cooled by CC. The activities to be performed during the 
refurbishment project and the various system alignments required 
have been evaluated using the Individual Plant Examination (IPE) 
Probabilistic Safety Assessment (PSA) model for North Anna Power 
Station. This model is used in a manner that is generally consistent 
with the Nuclear Energy Institute (NEI)/Electric Power Research 
Institute (EPRI) draft PSA Applications Guide (Revision H). The 
effect on the PSA model is a slight increase in the frequency of 
reactor trips and an increase in the probability of RHR failure.
    The increased frequency of reactor trips is due to the decreased 
reliability of the CC system to supply cooling to the reactor 
coolant pump (RCP) motors. When only one SW header is available to 
the CCHXs, the increased frequency of losing this single header can 
be conservatively estimated by combining the failure probability of 
both SW pumps (approximately 1.5E-4 based on IPE PSA data). Also 
considered was the frequency of pipe rupture anywhere in the single 
available header. When the single SW header fails to supply cooling 
to the CCHXs, the CC system will heatup causing inadequate cooling 
for sustained operation of the RCPs. Tripping these pumps results in 
a reactor trip. The second SW header can be expected to supply other 
equipment with cooling. A sensitivity analysis shows the increase in 
CDF as a result of the increased reactor trip frequency to be less 
than 1E-8 per year.
    The CC system is also included in the PSA model as a support 
system for RHR cooling. The RHR system is used to reduce reactor 
coolant system temperatures from 350 deg.F (hot shutdown) to 
140 deg.F (cold shutdown). The only accident initiator that requires 
the unit to be cooled down and placed on RHR cooling are sequences 
which are initiated with a steam generator tube rupture. (Note that, 
for the North Anna plant design, RHR is separate from the safety 
injection system and the low head safety injection pumps.) The 
increased probability for the loss of RHR when only one SW header is 
available to the CCHXs is estimated using fault tree analysis and is 
dominated by the failure of both SW pumps. The probability for the 
loss of both SW pumps aligned to the CCHXs is estimated to be 1.5E-
4. The effect of this increase in RHR failure probability was 
determined by adding this probability to the top single event in the 
RHR function and recalculating the new CDF. The resulting increase 
in CDF as a result of RHR system failure following a steam generator 
tube rupture is less than 1E-8 per year.
    The CC system is further included in the PSA model as part of 
the loss of RCP seal cooling as an initiating event and as a loss of 
function during other initiating event scenarios. The effect on the 
probability for a loss of RCP seal cooling due to losing CC cooling 
to the RCP thermal barriers is negligible due to the high 
reliability of the charging system to provide seal injection.
    The total effect of this pipe refurbishment project was 
estimated by a sensitivity analysis combining both the change in the 
reactor trip initiating event frequency and the increased failure 
probability of RHR resulting in less than a 1E-6 per year increase 
in CDF. Since this project will not affect the containment systems, 
there would not be any significant change in off-site dose, except 
that resulting directly from the increase in CDF. These minor 
increases in CDF and off-site dose are less than what is defined as 
risk significant in the NEI/EPRI draft PSA Applications Guide.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed temporary Technical Specifications changes do not 
affect the basic method of operation of the SW or CC systems. The 
purpose of the proposed changes is to permit extended operation of 
the CC system with one independent source of SW cooling. During the 
project, there will be a significant time period when all the CCHXs 
are aligned to one SW loop, the possibility of an interruption of SW 
supply to the heat exchangers during a DBA is eliminated by 
defeating the closure of the 24-inch SW isolation MOVs to the CCHXs 
on a SI/CDA signal. Both SW headers will be available for equipment 
required for safe shutdown of the units (i.e., RSHXs, charging 
pumps, and CR/ESGR chillers). The SW pipe repair activities and the 
installation/removal of the SW cross-connect piping do not create 
the possibility for a malfunction of equipment different than 
previously evaluated. Therefore, implementation of the restoration 
project and approval of the proposed Technical Specifications 
changes will not introduce any new accident initiators nor affect 
the performance of accident mitigation systems.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes to the schedule only provide operational 
flexibility to perform the required SW pipe refurbishment. The 
Technical Specifications continue to require the SW and CC systems 
to remain functional during the period with a single SW supply to 
the CCHXs. As stated in item (1) above, the SW system is fully 
capable of performing its DBA function during the course of the pipe 
refurbishment project with the proposed Technical Specification 
changes in place. The effect of this pipe refurbishment project on 
CC system reliability was estimated by a sensitivity analysis 
combining both the change in the reactor trip initiating event 
frequency and the increased failure probability of RHR resulting in 
less than a 1E-6 per year increase in CDF. Since this project will 
not affect the containment systems, there would not be any 
significant change in off-site dose, except that resulting directly 
from the increase in CDF. These minor increases in CDF and off-site 
dose are less than what is defined as risk significant in the NEI/
EPRI draft PSA Applications Guide. Therefore, there is not a 
significant reduction in margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: David B. Matthews

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.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick 
County, North Carolina

    Date of application for amendments: March 31, 1995
    Brief description of amendments: The proposed amendments would 
provide an exception to Technical Specification (TS) 3.0.4. TS 3.0.4 
allows entry of a unit into another operational condition only if the 
conditions of the Limiting Conditions for Operation (LCOs) are met 
without reliance on TS action statements. The exception requested by 
[[Page 24925]] the licensee would allow a change in a unit's 
operational condition in a specific situation in which the unit's LCO 
concerning the minimum number of operable offsite power circuits is not 
fully satisfied. Specifically, the exception would allow an operational 
mode change of a unit if the second unit is in Operational Condition 4 
or 5 (i.e., cold shutdown or refueling) and one of the second unit's 
offsite power circuits is inoperable.
    Date of publication of individual notice in Federal Register: April 
13, 1995 (60 FR 18860)
    Expiration date of individual notice: May 15, 1995
    Local Public Document Room location: The University of North 
Carolina at Wilmington, William Madison Randall Library, 601 S. College 
Road, Wilmington, North Carolina 28403-3297

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

    Date of amendment request: April 10, 1995, as supplemented April 
12, 1995
    Brief description of amendment request: The proposed amendment 
would revise Technical Specification (TS) 4.6.2.2.d to delete the 
reference to the specific test acceptance criteria for the Containment 
Recirculation Spray Pumps and replace the specific test acceptance 
criteria with reference to the requirements of the Inservice Testing 
(IST) Program. In addition, the 18-month test frequency would be 
replaced with the test frequency requirements specified in the IST 
Program. The current footnote (1) pertaining to the performance of 
recirculation spray pump 2RSS*P21A would be deleted.
    Date of publication of individual notice in Federal Register: April 
18, 1995 (60 FR 19417)
    Expiration date of individual notice: May 18, 1995
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

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

    Date of amendment request: April 14, 1995
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to allow the use of the 
Westinghouse Electric Corporation sleeving process for repairing steam 
generator tubes.
    Date of publication of individual notice in Federal Register: April 
21, 1995 (60 FR 19969)
    Expiration date of individual notice: May 22, 1995
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578.

Notice Of Issuance Of Amendments To Facility Operating Licenses

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

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick 
County, North Carolina

    Date of application for amendments: July 22, 1994, as supplemented 
on March 6, 1995
    Brief description of amendments: The amendments change the 
Technical Specifications to implement a performance based assessment 
program, including corresponding organizational and functional changes. 
Specifically, the changes affect the independent review function, the 
independent assessment of plant activity and the Independent Safety 
Engineering Group. These functions will be performed by the Nuclear 
Assessment Section (NAS). The NAS's fundamental role will be to: (1) 
assist plant management in the early identification of issues that may 
prevent the plant from achieving quality, and (2) ensure effective 
correction of deficiencies.
    Date of issuance: April 18, 1995
    Effective date: April 18, 1995
    Amendment Nos.: 177 and 208
    Facility Operating License Nos. DPR-71 and DPR-62. Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 
45017) The March 6, 1995, submittal added Radiation Protection to the 
list of assessments in TS 6.5.5.2 and reworded Section 6.5.4.4, but did 
not change the no significant hazards consideration determination as 
published in the Federal Register. The Commission's related evaluation 
of the amendments is contained in a Safety Evaluation dated April 18, 
1995.No significant hazards consideration comments received: No
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

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

    Date of application for amendment: June 18, 1992, as supplemented 
December 8, 1992 and February 3, 1995
    Brief description of amendment: The amendment adds limiting 
conditions of operation and surveillance requirements for the 
pressurizer power-operated relief valves and their associated block 
valves whenever average temperature is above 350 degrees F or the 
reactor is critical. Specifications are also added for low-temperature 
overpressure protection [[Page 24926]] whenever average temperature is 
less than 350 degrees F and the reactor coolant system is not vented to 
the containment.
    Date of issuance: April 14, 1995
    Effective date: April 14, 1995
    Amendment No.: 162
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 2, 1992 (57 
FR 40208). Renoticed on March 1, 1995 (60 FR 11127) The December 8, 
1992, letter corrected a typographical error and did not affect the no 
significant hazards consideration. The licensee's letter dated February 
3, 1995, proposed a revision to the TS regarding block valve testing in 
accordance with Generic Letter 90-06 recommendations. The proposed 
change was noticed on March 1, 1995 (60 FR 11127). The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated April 14, 1995.No significant hazards consideration comments 
received: No
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.

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

    Date of application for amendment: November 4, 1994, as 
supplemented April 6, 1995.
    Brief description of amendment: The amendment changes the testing 
frequency of the turbine overspeed protection valves from monthly to 
quarterly to implement an enhancement recommended by Generic Letter 93-
05, ``Line-Item Technical Specification Improvements to Reduce 
Surveillance Requirements for Testing During Power Operation.'' The 
April 6, 1995 submittal provided clarifying information only, and did 
not change the proposed no significant hazards determination.
    Date of issuance: April 27, 1995
    Effective date: April 27, 1995
    Amendment No.: 164
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 7, 1994 (59 FR 
63115) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 27, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.

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: January 19, 1995, as 
supplemented March 20, 1995
    Brief description of amendment: The amendment revises Technical 
Specification 4.0.3 and its associated Bases to provide for a delay 
period in which to perform a surveillance that was not performed within 
its specified frequency.
    Date of issuance: April 17, 1995
    Effective date: April 17, 1995
    Amendment No.: 56
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8742) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 17, 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: July 22, 1994, as supplemented 
March 6, 1995.
    Brief description of amendment: The amendment implements a 
performance- based assessment program, including corresponding 
organizational and functional changes. Specifically, the changes affect 
the Independent Review (IR) function, the independent assessment of 
plant activity and the Independent Safety Engineering Group. These 
functions will be performed by the proposed Nuclear Assessment Section 
(NAS). The NAS will perform internal evaluations and assessment 
activities and serve as plant management's staff for the objective 
oversight of plant performance relating to nuclear safety, reliability, 
and quality. The NAS's fundamental role will be to: (1) assist plant 
management in the early identification of issues which may prevent the 
plant from achieving quality performance on a sustained basis; and (2) 
ensure effective correction of deficiencies.
    Date of issuance: April 21, 1995
    Effective date: April 21, 1995
    Amendment No.: 57
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 
45019) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 21, 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 No. 50-374, LaSalle County 
Station, Unit 2, LaSalle County, Illinois

    Date of application for amendment: March 31, 1995
    Brief description of amendment: The amendment revises the safety/
relief valve (SRV) safety function lift setting allowable tolerance 
band from -3/+1% to plus or minus 3% and includes a requirement for the 
lift settings to be within plus or minus 1% of the technical 
specification limit following testing.
    Date of issuance: April 25, 1995
    Effective date: Immediately, to be implemented prior to restart 
from the sixth refueling outage.
    Amendment No.: 89
    Facility Operating License No. NPF-18: The amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes (60 FR 17590 dated April 6, 
1995). That notice provided an opportunity to submit comments on the 
Commission's proposed no significant hazards consideration 
determination. No comments have been received. This notice also 
provided for an opportunity to request a hearing by May 8, 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 
amendment, finding of exigent circumstances, and final determination of 
no significant hazards consideration is contained in a Safety 
Evaluation dated April 25, 1995.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 
61348. [[Page 24927]] 

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment:  June 1, 1994, as supplemented 
on January 25, 1995, April 7, April 19, and April 26, 1995.
    Brief description of amendment: The amendment revises Technical 
Specification Section 3.10 to allow extended Rod Position Indication 
(RPI) deviation limits and on-line calibration of the RPI channels for 
cycle 13 only.
    Date of issuance: April 28, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 182
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37069). The January 25, April 7, April 19, and April 26, 1995, 
submittals provided clarifying information that did not affect the 
initial no significant hazards determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
April 28, 1995.No significant hazards consideration comments received: 
No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: February 10, 1995, as 
supplemented March 27 and 30, 1995
    Brief description of amendment: This amendment revises the 
Technical Specifications to allow a one-time deferral of several 18-
month interval surveillance tests until the upcoming scheduled 
refueling outage to avoid the necessity of imposing a plant shutdown 
solely for the sake of their performance. In the March 30, 1995, letter 
the license also withdrew its request for deferral of several 
surveillance tests.
    Date of issuance: April 20, 1995
    Effective date: April 20, 1995
    Amendment No.: 164
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11131) The March 27 and 30, 1995, letters provided clarifying 
information which was within the scope of the initial notice and did 
not affect the staff's original proposed no significant hazards 
consideration determination.The Commission's related evaluation of the 
amendment and of the withdrawalof certain surveillance test deferrals 
is contained in a Safety Evaluation dated April 20, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.

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

    Date of application for amendments: February 23, 1995, as 
supplemented by letter dated March 21, 1995.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3.8.2.1 and TS 3.8.3.1 to allow installation of 
replacement equipment in response to an Electrical Distribution Systems 
Functional Inspection, conducted by the NRC in July 1991. The existing 
breaker arrangement could result in a trip of both the battery and main 
breakers if a fault occurs on one of the 125-V dc panelboards. The 
licensee committed to have these breakers replaced in 1995 with a 
better coordinated design to eliminate the concern.
    Date of issuance: April 14, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 155 and 137
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 8, 1995 (60 FR 
12791) The March 21, 1995, letter provided clarifying information that 
did not change the scope of the February 23, 1995, 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 April 14, 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: September 2, 1992
    Brief description of amendments: These amendments revise the 
Appendix A Technical Specifications relating to the required 
surveillance frequency for comparing the incore and excore axial 
imbalance. The revision requires comparison of the incore to excore 
axial imbalance at least once every 31 Effective Full Power Days above 
15 percent of rated thermal power rather than once every 31 days above 
15 percent of rated thermal power as was previously required.
    Date of issuance: April 26, 1995
    Effective date: April 26, 1995
    Amendment Nos.: 186 and 67
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 14, 1992 (57 FR 
47128) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 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 Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Unit Nos. 1 and 2, Pope County, Arkansas

    Date of amendment request: June 20, 1994
    Brief description of amendments: The amendments relocated the 
requirements of the quality assurance program and the security and 
emergency plans from the administrative controls section of the 
technical specifications to the respective licensee-controlled 
documents.
    Date of issuance: April 25, 1995
    Effective date: 90 days from date of issuance
    Amendment Nos.: 179 and 160
    Facility Operating License Nos. DPR-51 and NPF-6. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 17, 1994 (59 FR 
42340) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 25, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

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

    Date of amendment request: August 5, 1993
    Brief description of amendment: The amendment removed the 
requirements associated with loose-part detection [[Page 24928]] system 
from the Technical Specifications for Waterford Steam Electric Station, 
Unit 3. These requirements will be incorporated into the Waterford 3 
Updated Final Safety Analysis Report and maintained under the 
provisions of 10 CFR 50.59.
    Date of issuance: April 20, 1995
    Effective date: April 20, 1995
    Amendment No.: 104
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 15, 1993 (58 
FR 48382) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 20, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

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

    Date of amendment request: April 4, 1995, as supplemented by letter 
dated April 5, 1995
    Brief description of amendment: The amendment changed the Appendix 
A Technical Specifications (TSs) by revising the TSs for moderator 
temperature coefficient. The amendment approves a one time deviation by 
excluding the two-thirds end-of-cycle moderator temperature coefficient 
test requirement for Cycle 7.
    Date of issuance: April 27, 1995
    Effective date: April 27, 1995
    Amendment No.: 105
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration: Yes (60 FR 18431, dated April 11, 
1995). The 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 May 11, 1995, but stated 
that 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 determination of no significant 
hazards consideration is contained in a Safety Evaluation dated April 
27, 1995.No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Mississippi Power & 
Light Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit 
1, Claiborne County, Mississippi

    Date of application for amendment: October 12, 1994
    Brief description of amendment: The amendment removed License 
Condition 2.C.(26) related to Turbine Disk Integrity.
    Date of issuance: April 17, 1995
    Effective date: April 17, 1995
    Amendment No: 121
    Facility Operating License No. NPF-29. Amendment revises the 
license.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55868) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 17, 1995. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120.

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

    Date of application for amendments: May 23, 1994
    Brief description of amendments: These amendments will relocate the 
seismic monitoring instrumentation Limiting Conditions of Operation, 
Surveillance Requirements and the associated tables contained in 
Technical Specifications 3.3.3.3, 4.3.3.3.1 and 4.3.3.3.2 to the 
Updated Final Analysis Report.
    Date of issuance: April 25, 1995
    Effective date: April 25, 1995
    Amendment Nos.: 135 and 74
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 6, 1994 (59 FR 
34664) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 25, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

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

    Date of application for amendments: January 20, 1995
    Brief description of amendments: The amendments revise the 
administrative requirements of Technical Specification (TS) 6.4.1.2 
related to the areas of technical expertise that must be represented on 
the Plant Review Board (PRB). The licensee proposed this change in 
order to maintain an appropriate level of PRB expertise after the 
implementation of a planned reorganization that includes combining 
certain departments that are listed separately in the current TS 
6.4.2.1 requirements.
    Date of issuance: April 27, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 84 and 62
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 6, 1995 (60 FR 
7077) The April 4, 1995, letter provided additional and clarifying 
information that did not change the scope of the January 20, 1995, 
application or the initial proposed no significant hazards 
consideration determination.The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated April 27, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830.

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

    Date of amendment request: November 8, 1994, as supplemented by 
letter dated March 14, 1995.
    Brief description of amendments: The amendments require that only 
one of the two battery chargers associated with each Class 1E 125-VDC 
Channel I and Channel IV is operable.
    Date of issuance: April 17, 1995
    Effective date: April 17, 1995, to be implemented within 31 days.
    Amendment Nos.: Unit 1 - Amendment No. 73; Unit 2 - Amendment No. 
62 [[Page 24929]] 
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 7, 1994 (59 FR 
63123) The March 14, 1995, supplement withdrew that portion of the 
proposed amendments where the required wording was already incorporated 
into the Technical Specifications by amendments issued on February 14, 
1995, in response to another amendment request. The March 14, 1995, 
letter also provided clarifying information and did not change the 
original no significant hazards consideration determination. The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated April 17, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

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

    Date of application for amendment: November 10, 1994, as 
supplemented March 1, 1995
    Brief description of amendment: The amendment revises the Duane 
Arnold Energy Center Technical Specification Section 3.2.A to refer to 
the Offsite Dose Assessment Manual for the setpoint of the Offgas Stack 
Radiation Monitor and makes the ``Applicable Operating Mode'' and the 
``Action'' statements for these instruments consistent with the 
required function. The Action statement for the other instruments which 
initiate Secondary Containment isolation is also revised to be 
consistent with the current practice and with the function of those 
instruments. The Basis is also revised to add further description of 
the function and requirements.
    Date of issuance: April 25, 1995
    Effective date: April 25, 1995
    Amendment No.: 209
    Facility Operating License No. DPR-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 65815) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 25, 1995.The March 1, 
1995, submittal provided supplemental information that did not change 
the initial proposed no significant hazards consideration 
determination.No significant hazards consideration comments received: 
No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401.

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

    Date of application for amendments: April 6, 1994
    Brief description of amendments: The amendments delete part of 
License Condition 2.C.(4) to Operating License No. DPR-58 and part of 
License Condition 2.C.(3)(o) to Operating License No. DPR-74 on fire 
protection. The related fire protection safety evaluation also changes 
three of the modifications listed in Table 1 of the Safety Evaluation 
Report of July 31, 1979, that supported amendments nos. 31 and 12 to 
Operating Licenses No. DPR-58 and No. DPR-74, respectively.
    Date of issuance: April 19, 1995
    Effective date: April 19, 1995
    Amendment Nos.: 194 and 180
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: September 28, 1994 (59 
FR 49429) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 19, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

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

    Date of application for amendment: March 9, 1995
    Brief description of amendment: The amendment revises Technical 
Specification Section 4.6.1.2.a, Primary Containment/Containment 
Leakage. This change allows the second Type A containment leak rate 
test to be performed at refueling outage 5 instead of refueling outage 
4, consistent with an exemption to 10 CFR Part 50, Appendix J which has 
been granted.
    Date of issuance: April 24, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 65
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 23, 1995 (60 FR 
15310) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 24, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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: December 2, 1994
    Brief description of amendment: The amendment changes the Millstone 
3 Technical Specification Table 4.3-1 by adding a note for certain 
Functional Units which would allow an entry into Mode 2 or Mode 1 
before performing calibration for the power range detectors.
    Date of issuance:  April 26, 1995
    Effective date:  As of the date of issuance to be implemented 
within 30 days.
    Amendment No.: 109
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 1, 1995 (60 FR 
6304) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 26, 1995. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
LondonTurnpike, Norwich, CT 06360.

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

    Date of application for amendments: February 23 and March 3, 1995
    Brief description of amendments: The amendments revise the Prairie 
Island Technical Specifications section 4.4.A.5 to add the phrase ``and 
all approved exemptions.'' after the reference to 10 CFR Part 50, 
Appendix J. This revision will allow implementation of approved 
exemptions from the testing schedule requirements of 10 CFR Part 50, 
Appendix J, Section III.D.1.(a).
    Date of issuance: April 18, 1995
    Effective date: April 18, 1995, with full implementation within 30 
days.
    Amendment Nos.: 117 and 110
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications. [[Page 24930]] 
    Date of initial notice in Federal Register: March 15, 1995 (60 FR 
14025). The March 3, 1995, letter provided clarifying information 
within the scope of the original submittal and did not change the 
staff's initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 18, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location:  Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

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

    Date of application for amendments: August 17, 1994 (Reference LAR 
94-06)
    Brief description of amendments: The proposed amendments increase 
the allowed outage time of the refueling water storage tank (RWST) for 
adjustment of boron concentration from one to eight hours as contained 
in Technical Specifications Section 3.5.5.
    Date of issuance: April 14, 1995
    Effective date: April 14, 1995, to be implemented within 30 days of 
issuance
    Amendment Nos.: Unit 1 - Amendment No. 101; Unit 2 - Amendment No. 
100
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 12, 1994 (59 FR 
51621) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 14, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

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: October 25, 1994 as 
supplemented February 13, 1995.
    Brief description of amendments: The amendment clarifies the 
technical specification surveillance requirements and bases for high 
pressure coolant injection system testing at low reactor pressure.
    Date of issuance: April 18, 1995
    Effective date: April 18, 1995Amendments Nos.: 200 and 202
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications. Public comments requested as to 
proposed no significant hazards consideration: Yes (59 FR 55498 dated 
November 7, 1994). That notice provided an opportunity to submit 
comments on the Commission's proposed no significant hazards 
consideration determination, and also provided an opportunity to 
request a hearing by December 7, 1994. No comments or requests for 
hearings have been received. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated April 18, 1995.
    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.

Philadelphia Electric Company, Public Service Electric and Gas 
Company, Delmarva Power and Light Company, and Atlantic City 
Electric Company, Docket No. 50-278, Peach Bottom Atomic Power 
Station, Unit No. 3, York County, Pennsylvania

    Date of application for amendment: January 13, 1995 as supplemented 
by letters dated March 14, 1995 and April 12, 1995.
    Brief description of amendment: The requested changes would modify 
Tables 3.7.1 and 3.7.4 of the Technical Specifications (TS) to reflect 
a change in the number of primary containment penetrations and 
isolation valves associated with the traversing in-core probe (TIP) 
system. In order to prevent confusion with the staff's review of PECO's 
September 29, 1994 application to implement improved TS at Peach 
Bottom, the staff is issuing the license amendment regarding the TIP 
system for Unit 3 only.
    Date of issuance: April 24, 1995
    Effective date: April 24, 1995
    Amendment No.: 203
    Facility Operating License No. DPR-56: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11139) The March 14, 1995 and April 12, 1995, letters 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 April 24, 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 Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
County, New Jersey

    Date of application for amendments: September 20, 1994
    Brief description of amendments: The amendments modify the 
Technical Specifications for auxiliary feedwater to reduce the 
secondary side steam pressure required for testing the turbine driven 
auxiliary feedwater pump and to allow 24 hours to perform the test 
after reaching the minimum test pressure.
    Date of issuance: April 17, 1995
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: 165 and 146
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55889) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 17, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.

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

    Date of application for amendment: February 3, 1994, as 
supplemented September 19, 1994, and November 23, 1994
    Brief description of amendment: The amendment changes the Technical 
Specifications to reflect a reduction in Reactor Coolant System flow.
    Date of issuance: April 17, 1995
    Effective date: April 17, 1995
    Amendment No.: 147
    Facility Operating License No. DPR-75: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 15, 1995 (60 FR 
14028) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 17, 1995.No 
[[Page 24931]] significant hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.

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

    Date of application for amendment: March 13, 1995
    Brief description of amendment: This amendment revises Technical 
Specification 4.4.2.4.a to replace specific leakage rate testing 
frequencies for containment isolation valves that require Type C 
testing for the 1995 refueling outage to be completed prior to exiting 
Cold Shutdown tentatively scheduled for April 27, 1995.
    Date of issuance: April 26, 1995
    Effective date: April 26, 1995
    Amendment No.: 59
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 22, 1995 (60 FR 
15167) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 26, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.

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

    Date of amendment request: February 23, 1994 (LAR 94-005, TXX-
94034)
    Brief description of amendments: These amendments changed Technical 
Specification (TS) 3/4.5.1, ``Emergency Core Cooling Systems, 
Accumulators, Cold Leg Injection,'' to: 1) allow a one hour allowed 
outage time following discovery of a closed cold leg injection 
accumulator discharge isolation valve in Modes 1, 2, or 3; 2) eliminate 
the redundant requirement to reverify accumulator boron concentration 
following fill from the refueling water storage tank RWST; 3) remove 
the accumulator water level and pressure channel analog channel 
operational test and channel calibration from the TSs; and 4) change 
the accumulator limits to analysis values rather than indicated values. 
Also these amendments modified TS 3/4.5.2, ``ECCS Subsystems - Tavg 
 350 deg.F'' to reduce the visual inspection frequency 
following containment entries.
    Date of issuance: April 27, 1995
    Effective date: April 27, 1995, to be implemented within 30 days.
    Amendment Nos.: Unit 1 - Amendment No. 40; Unit 2 - Amendment No. 
26
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39597) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 27, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019.

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

    Date of amendment request: August 9, 1994, (LAR 94-013, TXX-94211)
    Brief description of amendments: These amendments eliminated ``High 
Negative Neutron Flux Rate'' reactor trip function based on analyses 
which demonstrate that the protection provided by the reactor trip 
function is not required. The affected Technical Specifications were: 
2.2.1, ``Reactor Trip System Instrumentation Setpoints,'' and 3/4.3.1, 
``Reactor Trip System Instrumentation.'' Also affected was Bases 
Section 2.2.1.
    Date of issuance: April 17, 1995
    Effective date: April 17, 1995, to be implemented within 30 days.
    Amendment Nos.:  Unit 1 - Amendment No. 39; Unit 2 - Amendment No. 
25
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 28, 1994 (59 
FR 49438) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 17, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019.

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

    Date of application for amendment: September 9, 1994, as 
supplemented on December 22, 1994
    Brief description of amendment: The amendment revises the Technical 
Specification (TS) 3/4.8.2.1, 3/4.8.2.2, 3/4.8.3.1, and 3/4.8.3.2. The 
changes address the 125-volt DC buses and adds provisions for swing 
battery chargers, and removes provisions for the 4160-volt and 480-volt 
AC emergency buses.
    Date of issuance: April 18, 1995
    Effective date: April 18, 1995
    Amendment No.: 99
    Facility Operating License No. NPF-30. Amendment revises the 
Technical Specification Bases and FSAR.
    Date of initial notice in Federal Register: January 4, 1995 (60 FR 
506) The December 22, 1994, letter provided supplemental information 
that did not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated April 18, 1995. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

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

    Date of application for amendments: February 14, 1995
    Brief description of amendments: These amendments modify the 
Technical Specifications (TS) to revise Section 4.4.D of the TS to 
permit approved exemptions to the containment integrated leak rate test 
frequency requirements.
    Date of issuance: April 18, 1995
    Effective date: April 18, 1995
    Amendment Nos.: 196 and 196
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 15, 1995 (60 FR 
14029) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 18, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.

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

    Date of application for amendments: September 6, 1994, as 
supplemented March 7, 1995
    Brief description of amendments: These amendments modify the 
Technical Specifications to revise the [[Page 24932]] review 
responsibilities of the Station Nuclear Safety and Operating Committee 
and the Management Safety Review Committee.
    Date of issuance: April 21, 1995
    Effective date: April 21, 1995
    Amendment Nos.: 197 and 197
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 12, 1994 (59 FR 
51631) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 21, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.

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

    Date of application for amendment: April 1, 1993
    Brief description of amendment: This amendment revises TS 3.8.1, 
``A.C. Sources'' by increasing the minimum required level of diesel 
generator fuel storage capacity. This change is based on testing and 
revised calculations that demonstrated that the existing levels of DG 
fuel storage were inadequate to meet the post-loss of coolant accident 
fuel consumption requirements for seven days of operation.
    Date of issuance: April 25, 1995
    Effective date: April 25, 1995, to be implemented within 30 days of 
issuance
    Amendment No.: 136
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1993 (58 FR 
28065) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 25, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352.

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

    Date of application for amendment: August 24, 1994 as supplemented 
on January 23, 1995.
    Brief description of amendment: The amendment revises Kewaunee 
Nuclear Power Plant (KNPP) Technical Specification (TS) 3.1.b.1 and 
Figure TS 3.1-4 regarding Low Temperature Overpressure (LTOP) 
protection for the reactor coolant pressure boundary. The change 
extends the LTOP requirements through the end of operating cycle 21 or 
18.40 effective full power years. The Basis Section has also been 
modified to reflect these changes.
    Date of issuance: April 26, 1995
    Effective date: April 26, 1995
    Amendment No.: 120
    Facility Operating License No. DPR-43. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 12, 1994 (59 FR 
51632). The January 23, 1995, submittal, provided additional reference 
material which did not change the initial no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated April 26, 1995.No 
significant hazards consideration comments received: None.
    Local Public Document Room location:  University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.

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

    Date of application for amendment: November 8, 1994, as 
supplemented on January 9, February 14, March 8, and April 3, 1995.
    Brief description of amendment: The amendment revises Kewaunee 
Nuclear Power Plant (KNPP) Technical Specification (TS) 3.1.d, 
``Leakage of Reactor Coolant,'' TS 4.2.b, ``Steam Generator Tubes,'' 
and TS 3.4.a, ``Steam Generators,'' to allow application of a voltage-
based repair limit for the steam generator (SG) tube support plate 
(TSP) intersections experiencing outside diameter stress corrosion 
cracking (ODSCC). The amendment also reduces the allowed primary-to-
secondary operational leakage from any one SG from 500 gallons per day 
(gpd) to 150 gpd. These changes to the tube repair criteria are 
applicable for the 1995 to 1996 operating cycle (Cycle 21) only.
    Date of issuance: April 17, 1995
    Effective date: April 17, 1995
    Amendment No.: 118
    Facility Operating License No. DPR-43. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 7, 1994 (59 FR 
63127). The January 9, February 14, and March 8, and April 3, 1995, 
submittals provided clarifying information which did not change the 
initial no significant hazards consideration determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated April 17, 1995. No significant hazards 
consideration comments received: None.
    Local Public Document Room location:  University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.

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

    Date of application for amendment: September 7, 1994
    Brief description of amendment: The amendment revises Kewaunee 
Nuclear Power Plant (KNPP) Technical Specifications (TS) by adding two 
new sections, TS Section 3.0 and TS Section 4.0, with associated bases. 
TS Section 3.0 establishes the general requirements applicable to each 
of the Limiting Conditions for Operation (LCOs) within Section 3 of the 
KNPP TS. TS Section 4.0 establishes the general requirements applicable 
to Surveillance Requirements. The new requirements of TS 4.0.b also 
affect TS Sections 4.5, 4.6, 4.7, and Tables TS 4.1-2 and 4.1-3.
    Date of issuance: April 18, 1995
    Effective date: April 18, 1995
    Amendment No.: 119
    Facility Operating License No. DPR-43. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 12, 1994 (59 FR 
51632)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 18, 1995.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required [[Page 24933]] 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 June 9, 1995, the licensee 
may file a request for a hearing with respect to issuance of the 
amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine [[Page 24934]] 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).

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

    Date of application for amendment: April 13, 1995, as supplemented 
April 18, 1995.
    Brief description of amendment: Amendment revises TS Section 
4.4.3.f, g, and h to allow the post accident heat removal system 
surveillance test interval to be changed from a 12-month interval to a 
refueling outage interval.
    Date of issuance: April 19, 1995
    Effective date: April 19, 1995
    Amendment No.: 163
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.The Commission's final determination of 
significant hazards consideration and related evaluation of the 
amendment is contained in a Safety Evaluation dated April 19, 1995.
    Local Public Document Room location:  Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.
    Dated at Rockville, Maryland, this 3rd day of May, 1995.
    For The Nuclear Regulatory Commission
Elinor G. Adensam,
Acting Director, Division of Reactor Projects - III/IV, Office of 
Nuclear Reactor Regulation

[Doc. 95-11367 Filed 5-9-95; 8:45 am]

BILLING CODE 7590-01-F